View source for DCL-15-069, License Amendment Request 15-03 Application of Alternative Source Term - Updated Final Safety Analysis Report Markup. Part 6 of 8

{{Adams
| number = ML15176A537
| issue date = 06/17/2015
| title = License Amendment Request 15-03 Application of Alternative Source Term - Updated Final Safety Analysis Report Markup. Part 6 of 8
| author name = 
| author affiliation = Pacific Gas &amp; Electric Co
| addressee name = 
| addressee affiliation = NRC/NRR
| docket = 05000275, 05000323
| license number = DPR-080, DPR-082
| contact person = 
| case reference number = DCL-15-069
| document type = License-Application for Facility Operating License (Amend/Renewal) DKT 50, Updated Final Safety Analysis Report (UFSAR)
| page count = 45
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:DCPP UNITS 1 & 2 FSAR UPDATE dose limits of 10 CFR Part 20.
15.5.17.2.4 4       Post LOCA-aGssde.".t 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, Or its equivalent to any part of the boAdy, for the duration of the most seVere design basis accident. This basis is consistent With GDC 10, 17-1.
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,-97-1-1999 and 10 CFR 50.67.
The control room ventilation system is described in Section 9.4.47 It is designed to limit the concentration of post-accident activity in the control room air to levels consistent with requirements of GDC 19,-1-971-1999 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,-19711999 and 10 CFR 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 limits. ...p.suesh.e.
also been calculated for the expected case LB1LOCA to obtain a more Fealistic estimate of exposur~e 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)-Direct gamma radiation to the control, room. from act;i*ty in the cOntainmen. structure
('3)-
(4)(2) Dineret gamma radiation to the cOntr room from. ac.tivity in the contaiRment leakage 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                     Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE Timing 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. 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=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 radioactivity release points / control room receptors applicable to a LOCA at either unit are provided in Table 15.5-23B. The X/Q values presented in Table 15.5-23B take into consideration the various release points-receptors applicable to the LOCA to identify the bounding I/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 and summarized in the notes of Table 2.3-147 and Table 2.3-148 for Unit 1 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 (MeV/sec) and integrated gamma energy release (MeV-hr/sec) 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 bendline, 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                    Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR 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 control room Yeentilation system is designed to mninimizc infiltration of post accident airborne WtiVity into the control roero                        *ornplex. Mode 4 operation of the ventilation system provides zone isolation with filtered positive pressuization and filtere reiclation. Mode 4 operation of the ventilation system is initiated automatically and the least contaminated positive. prsuizain finlet is selected manually as described in Chapter 0.1.. Both the pressur;izattion and partial Fecirculation air flow pass through high efficiency particulate air (HEPA) and charcoal filte.
In addition to pesitive pressuriation, there are vetibules on ontrol,                      room    doors that will inize *Ta1bleinfiltration.                  1.5.5 31 identifies infiltration pathIays and fewl*ates that have been used in the calculatio of pos accident cOntro room radiological exposures.
Airborne radiation doses inside the control room were evaluated for aODBA LOCA.
Regulator; Guide 1.4, Revision I was used to determi~ne activity levels in the containment. Activity releases are based on a containment leakage of 0.1 percent'day for the 46rt day and 0.05 percent/day thereafter.
The containment leakage was assumed to be released unfiltered fromn the containment build-ing to the atmosphere. Recirculati-on loop leakages, assumed to be from an RHR pump seal, will pass through charcoal filters and be released to the atmosphere through the main vent at the top of the containment.
I d I-aaioactlVlW ;From Me a!tmocpnero WGUIa ener Mne cOnRoIF room: nRF9u49 MvO painways:
il v
                          ..- ;----X:---- *:-:    ......& I *L. ..... L. L.
Via~ incl           11A7-WA-ir MtHMPrbe inruu uarfcoat nHiterc Purct-aRSzio 15.5-84              Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE via infiltratiGn of air inleakage The flow rate of pressur.ization airito*the control room is 2100 Cfm. The fioW rate of recirculated contrIo
                  "    l room    through the charcoal filtehrs ir 2100 cfm. Previous analyses had not taken credit for re*ircUlation Of controli rom air. This was an unnc-cssa*       y conservatism inthat a passive failure had already been assumed to occur (RHR pump seal leak) and a second failure is not required.
A   10 GIV nleakage Faeper StnadReiwPan, Sertio64, was GGnser~atively assumed On the analysis due to the possible pathway through the single doors fromn the epent codensing unit areas to the 14VAG equipment room. Additionally, an asued 10 second delay in closure of the CRVS outside air isolation dampers Fesults in 2110 cfM Of Gontrol roomF infiltration for the first 10 seconds followi.ng the design basis LOGA.
Table 15.5-32 presents a summar, of the parameterS used in the analysis.
The GGnRtrol rom shielding is desigRned to minimize direct gammF      ' a adiatiOR (containm-nt shine).. CtrGI    roo. exposures resulting   from containment     shine were estimated using CSOSHnIItr. The control room reptr point is 27 feet fAcro            the contaiment  structure and protected by an additional 2.5 foet thirckeoncete shield. A fuiher ctributionea to control roomdifret gamma radiation results folm the atmospheric activity cloud external toethlecnrol room. COntrol rm expesures resulting from plume shine were estimated using !SOSHLO II. The shine exposure model assumes a parallelepiped radiation toure loeated diregtly above      th roomo.
control                The control roomdueptor point is protected by a 1.5 foot thick conrGete shield.
Control Room Operator Dose during 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 / 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 controlroom 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                       Revision 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 and (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 of post accident contrl room exposures and egress-ingress exposures developed in support of DCPP original licensing basis are listed in Table 15.5-33 and summarized below. The sum of the DBA case exposures are Within the specified criteria, and the cxpccted case exposures, demonOstrate the conser~atiGm of the IDBA case exposures.
: 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                  Revision 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 Nacuum relief release - this release occurs at accident initiation (before t=24hr), so there is no dose contribution to the control operator during routine ingress /egress 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 effected 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                        Revision 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:
o   Dose consequences due to containment leakage based on a TID-14844 based scenario will bound the dose consequences based on an AST scenario.
o  Thyroid 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
                      /egress 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 rem. 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 lI.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                      Revision 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 rem 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 TSC 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                    Revision 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 X/Q values presented take into consideration the various release points-receptors applicable to the LOCA to identify the bounding X/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                    Revision 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.
15.5.17.2.44-6          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 (1-)(3) ESF System Leakage (24(4) RHR Feiiuatien-pump seal Failureleepleakaee (5)      Controlled post accident containment"..ntingRWST 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 regu.ations 10 CFR Part 100, in Regulatory Guide 1.18310 CFR Part 50, and hthe-other regulatory guidance identified safety *nd,,egulatery guidesabove. 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 Pa,"t 10050.67 and Regulatory Guide 1.183-,
10 CFR Part 50, and 10 CFR Part 20.
15.5-91                  Revision 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.
Final!',, the analysiG demonstratees that the acceptance criteria are met as follows:
(1) The radiologiGa* conSequences of a major rupture of prim.ary coolant Pipes shall take into consideration fi6sion product releases due to leakage fromtc containment, post LOCA reirc'FUlation loop leakage in the Auxiliary Building (iRGlusive of a RHR pump seal failure resulting in a 50 gpmn leak for 30 Minutes starting at T-24 hre post LOCA), and containment shine a6 shown in SecGtin 15.5.17.2.11.
(2) The radiological consequences of a major ruPture of primary co.lant pipes shall not exceed the dose limits of 10 CFR 100.11 as outlined bclow-:
15.5-92                     Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE
: i. An ind,,dua!l located at any point on the boundary of the exclusion area for the tWo hours immediately following the onset of the postulated fission product rehalease sshall not receive a total radiation do-e to the whole body in excess of 25 rem OF a total Fadiation dose in excess of 3 r-m. toA the thyroid from iodine exposure as shown by the EAB whole body dose reported for containment shine in Sec-tio n 15.5. 17.2.6, and the remaining doses presonted in Tae ii. AR individual locsated at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting ffrom the postulated fission product release (during the entire period of its passage), shall not Fr~eeve a total radiation dose to the whole body in ex esof 25 rem, or a total radiation do inexes of 300 rem to the thyroid from iodine exposure as shown by the EAB whole body dose reported for cOntainment shine in Section 15.5.17-.2.6 (conscrwative when applied to the LPZ=), and thereainn doses presented in Table 15.5 75.
(3) Inaccordance w'ith the requirements of GDC 19, 1971, the dose to the coentro room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 Frem thyroid and beta skin, Reference 51) for the duration of the accident as shown in Table 15.53.
(4) In the event conRtrolled venting of the containment is imlmnedpst LO.A.
Using the containment hydrogen purge system (se~ves as a back up rapability for hydro~gen conrol1 to the hydrogen recOmbiners), an individual located at any point on the boundary of the exclusion area, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall no~t receive a total radiation dose to the whole body inR excGess of the annual dose limit of 10 CFR Part 20 as shown in Table 15.5 29.
15.5.18 RADIOLOGICAL CONSEQUENCES OF A MAJOR STEAM PIPE RUPTURE 15.5.18.1-      Acceptance Criteria The radiological consequences of a MVSLB 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 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.
15.5-93                    Revision 19 May 2010

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 0-1-)-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.T-he radiological consequences of a major steam p         rupture shall not-exceed the dose limits of 10 CF=R 100. 11 as outlined below:m (3)An individual located at any point on the boundar,' of the exclusion area for the two ho*ur immediately following the onset of the postulated fission product release shall not receive a total radiation dose in excess of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the pre existing idine spoike c~ase and 10 percent of the 10 CFR 100. 11 dose limits for the whole body and the thyroid for the accident initiated iodine spike case.
(4-)-
(5) An individual located at any point on the outer boundar,' of the low population-zone, who is exposed to theFadioactive loeud resulting from the postulated fission product release (during the entire peFi*d Of its passage), shall nRot eceive a total radiation dose in excess of the 10 CFR 100.11 dose iFim;ts for the whole body and the thyroid for the pre existing iodine spike case and 10 percent of the-10 CFR 100.11 dose limit, for the whole body and the thyroid for the accident initiiated iodine spike case.
(6y-(7) in accordance with the requirements of GDC 19, 197-1, the dose to the control room operator unde-r acident conditions shall not be in excess of 5 rem whole body Or its equivalent to any pa-t of the body (i.e., 30 eAm thyroid anrd beta skin, Reference 51) for the duration of the accident for both the pre accident and the accident initiated iodine spike cases.
15.5.18.2-         Identification of Causes and Accident Description 15.5.18.2.1         Activity Release Pathways As reported in Section 15.4.2, a major steam line rupture is not expected to cause cladding damage, and thus no release of fission products to the coolant is expected following this accident. If significant radioactivity exists in the secondary system prior to the accident, however, some of this activity will be released to the environment with the 15.5-94                    Revision 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 15.5.18.2.1 dis.....e.. the main steam" 1ine break (MSLB3) dose analysis of Freod Which is based on the OSGs. The OSG MVSLB1 dose analysi is bounding for the RSGs as diScussed in the following section;. (See-Table 6.4.2-1 o-f Reference 49 for a summar,' of OSG and RSG MSLBO staFeleases.)
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.
15.5.18.2.1 Rad~iological Assessment for AccGident Induced Leakage Bec~ause tubes in the faulted steam generator encounter a higher differential pressure during steam line rupture conditions than normal operating conditions, there is a-potential for pFFnarFy to secondaFy leakage in degFaded tubing to     increase to a rate that is higher than that during normnal operation. This leakage is referred to as accidn induc~ed leakage. This secstion provides the updated licensing basis description and radiological consequence analysis for a mnajor steam line rupture analysis using aR aCcidert induRed leak rate of 10.5 gpm (at room temperature conditions), which is higher than the operational leakage limit in the Technical Specifications. The NRC app15..1ed this analysis in aleater to PG&E dated FMebluar 20, 2003, Amendment: RE: Revision to Technical Specification 1.1, 'Definitions, Dose Equivalent 1131,' and Revised Steam Generator Tube Rupture and Main Steam Line BrFea Analysers." Application Of this. accident-induced leak rate is governed by SG Programn accrident induced leakage performance criteria documented in the Technical-Computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a MVSLB.
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 gCi/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 ltCi/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 giCi/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 (Ci) 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                      Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE Standard Temperature and Pressure (STP) conditions), is conservatively assumed to occur in the faulted SG. 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 212 OF, 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 continue for 10.73 hours, at which time shutdown cooling is initiated via operation of the RHR system and environmental releases are terminated.
15.5.18.2.3    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 X/Q 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                   Revision 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.
Primaryto Secondary SG 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/min)(38.8 s) / 60 s/minm 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 AtmosDheric Dispersion Factors 15.5-98                     Revision 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/Q values applicable to a MSLB 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 2.3-148.
The bounding control room dose following a MSLB at either unit is presented in Table 15.5-34. The method,.og''                            for perfoRm;ig the radiological assessm.ent follGow
                                               ,elected NRC SRP 15.1.5, "Steam System Piping Failures inside and Outside of Containment (P'./R),'" Revision 2, 1981. Using an accident induced leak rate of 10.5 gpmn (at room temperature conditions) in the faulted SG, caIculations using the LOCADOSE comnputer program dem.o.nstte that the off.ite doses are within 10 percent of 10 CFR 10.1 4
!;,,,;4*.-r,      ,-A .,+,,,
                        #    !l,,',rr,',r A*       * ;4-k;. f_rQn 10 41C74 I; ,;f*
The resultant doses from the MSLB event using an accident induced leak rate of 10.5 gpm are listed below. The limniting case is-the accnid-ent initiated iodine spike as the thyroid dose at the EAB is at the 30 rem limit.
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 case 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                   Revision 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 rem) TEDE as shown in Table 15.5-34.
(1) An individual located at any point on the boundar; of the exclusion area forth tw:o hours immediately follwing the onset of the postulated fission product 15.5.!8.2.!.
deliisfrthe whole b               andodythe thy              the por existing f roid                re               spike iodine aend1 zone","a  who '*                 to the* F-a`4;oa,-ti.,
1v,,*                       nl..tu,`  Fsnlting-. fr'.m   the p-,'ostul-*ate thf hyod      theoa         ccident initiated i         odine spike            as caseshown in Se                 ctionn release hall nt r                   adiation dose in                  ofexess the 10                  C!R 100.11 (2) in       individualoated caeand1t        th pctf e 10  aith  aypio 100.11 theCFR doseouteree~t       bondr 9171 limit     ofthe th e for wh       blee ole .... o populatin ody.. nd-roomepeatotal ,nn`4aGd`nt radiaittn                          offsdholl in exessnet be                       5 li        w0FR*01 for the accident initiated iodine spike casesas shownin Section 15.5.8.2.1.
tew hOle b6ed"qu5v5          th andt thoi  a     fo the prdy     existin iie           spik ca           and     10 ercint oftAhe          CfR 1e00.
510               d* ose*nlmis   fo       the awhle         bodtyhandthe tyid-(3     Iotedon  -Accrtance Crith      the reqie                            dose                       toie the contr acid        ninducedu     lak rate of 10.5 gpm. The limiting case is a thyroid dose at the EAB which cor                 sto the dose lit of 30 rem for an accident initiated iodine spike.
Thesbe hereiroom          opferdator bound dose estimates,           uinder the doses with the braccidntFondiion               RSGs shall         whic-h notbei       coanno~t ceilent excedtedseiis           of 10 CRewhl Repair                 (ARC)tequvadnos Criteria orimets 50.67,body                  for the steam anyeptancriei       tubes of        as the OSGsdothyride and83 generatorFthRodiegl30orem                                        beta200 15.5.19      an thelacdnontitdidnwpiecssa.soni RADIOLOGICAL ulned                         CONSEQUENCES OF A MAJOR RUPTURE                              eton1.821 OF A MAIN FEEDWATER PIPE EAs    anotd inZ  Setone 1551821rteaoitoeesiatsrricaheO                                                        ada 15.5.19.1-           -Acceptance Criteria The radiological consequences of a major rupture of a main feedwater pipe (referred to herein as a feedwater line break (FWLB)) 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 15.5-100                                 Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE 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 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 outlined below:
An individual located at any point on the boundar,' of the eXc!Usion area for the Me hourS immediately following the onset of the postulated fission product release shall not Sre ra total radiation dose to the whele body in excess of 25 rem Or a total radiation dose in exgess Of 300 rem tothe thyroid from iodine exptoree.
An indigvidual located at any point on the outer bouydax     ' of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entir    per oed ofits passage), shall not receive a total radiation doee to the whole body in exaeS of 25 rem, Or a total radiationd in e               sf 300 rem to the thyroid from iodine exposure 15.5.19.2-       Identification of Causes and Accident Description As reported in Section 15.4.2, a major feedwater line rupture is not expected to cause cladding damage, and thus no release of fission products to the coolant is expected following this accident. If significant radioactivity exists in the secondary system prior to the accident, however, some of this activity will be released to the environment with the feedwater 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. As discussed in Section 15.5.18, about 1.47cE+06 Ibmn of secondar; coolant is the limiting Condition IV.event release expected for a full cooldown without any condenser availability.
The radiological consequences of about 1.47E+96 Ibm of secondar; coolant releasge have been discussed in Section 15.5.18.
Per Standard Review Plan 15.2.8, Section III, Item 6 (Reference 86), the evaluation of the radiological consequences of a design basis FWLB may be based on a qualitative comparison to the results of the design basis MSLB.
The dose consequences following a FWLB will be bounded by a MSLB since the 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 & 2 FSAR 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.B1ased on the results discussed, it cGan be coc'luded that potential eXposures from major feedwater line ruptures will be well below the guideline levels specified in 10 CFR 100.11, and that the occurrence of such ruptures wou.not result in undue risk to the public.
Additionally, the analysis demonstrates that the acceptance criteria are met as follows:
The radiation dose to the whole body and to the thyoideof an individualloated at any point on the boundar; of the exclusion area for the two hours immediately followiRg the onset of the postulated fission poduct rele                gnificant as shown in Table 15.5 The      radiation dose to the whole body and to the thyerid of 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 as shonin fis pasge),period of its passage),oinsignifiant Table 155 9.
15.5.20 RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE (SGT-R) 15.5.20.1          Acceptance Criteria The radiological consequences of a SGTR 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 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.
(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                      Revision 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.
(1) The radiologica! consequences_ of a steam generator tube rupture shall not exceed the dose guidelines of SRP, Section 15.6.3, Revision 2, as outlined below
: i. An individual located at any point on the boundar,' of the exclusion area for the two hours mmediately following the onset of the postulated fissio product release shall not r*ceioe    a total radiation dose in excess of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the pre-existing iodine spike case, and 10 percent of the 10 C FR 100.11 dose limits forF the whole body and the thyroid for the accident initiated iodine spike i.An individual located at an" point on the outer boundar; 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 100.11 dose limits for the whole body and the thyroid for the pre existing iodine spike case, and 10 percent of the 10 CFR 100.11 dose limits for the whole body and the thyFrod for the accident initiated iodine spike case.
(2) In accr.Gdance with the requirements of GDC 19, 1971, the dose to the control roomR operator under accident conditions shall not be in excess of 5 rem w ho'ebd o t equivalent to any pa~t of the body (i'e., 30 rem thyroid and beta skin, Reference 51) for the duration of.the acciodent for both the pre accident and the accident initiated iodine spike cases.
15.5.20.2-      Identification of Causes and Accident Description 15.5.20.2.1      Activity Release Pathways This event is caused by the instantaneous rupture of a SG tube with a resultant release of primary coolant into the lower pressure secondary system. No melt or clad breach is postulated for the SGTR event. The calculation assumes a stuck-open PORV of the ruptured steam generator for 30 minutes. 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 15.5-103                     Revision 19 Mav
                                                                                        . *.-.j 201C:)

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 1.183, 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 pCi/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 p.Ci/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 335 times the equilibrium appearance rate corresponding to the 1 jiCi/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 piCi/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                    Revision 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                    Revision 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 pre-accident and accident initiated iodine spike cases.
Regardless of the starting point of the "Worst 2-hr Window," the 0-2 hrs X/Q is utilized.
15.5-106                      Revision 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 SGTR aGGident is FeaRalyzed                                                       tS fGF RSGs and is disGussed in SeGfien 15 3 and the tt       K                         I     ,rL         %.
A     LJ.I.
II II;il              4IneIJ.Te.ui t     r1rw dFau                                                                                       ti 1                                            Fey s teassf                                        he
     ..........                                                                                                 M-vm
                                                                                                                            -,M-15.5 20.2 1 Offkito Exunanure 1 ha a~,nI,             ,.,Gan        ,..'~ *ha      r.,Aialannn                     I  nnne.an,               ,annar'              af      -                     *n*              ar,,~nr *..ha            r.InGe urn Th                                        f th            di l i l                                                                      f t                                                                            t event assumer, that the FeaGtGF has beeR epeFatiAg at the maximum allewabla-T-e&#xfd;&#xfd; SpeGifiGatien (RefeFeRGe 22) 'iFAitG fGF PF;FnaFy Geelant aGt;yity and 1 9PM pFiFRaFy tO ser-,endaFy leakage forsuffir.ieRt time to e6tablish equilibFiurn GORGeRtFatiORG Gf-Fadienuelides ki the FeaGteF GGGIant and in the seGGndaFy GGelant. RadienuGlides ftem the       p~imrIIy            G~uuanti            eteF the steami genieIatFia                                                          the Fuptled tube ai~u p ivajI4I secondar; leakage, and are released to the atmosphere through the steam generator PORVs (and saft* valves) and via the condenser air ejector exhaust and/or the va*cuum pump exhaust (if in operatirn).
The quantity O*radioati*Vity released to th ent,                                                                                                           due t an SGTR, depends 9R          F;Fnar and 6eGGndaF GOGIant                                                       aGtiVit             iediRe s p ikin g                        effeGtS ,                pFimar N.')    SI I
.J~%flJI      ItflAI    3   IJU LI.Afl     II**      S flALAI III        I~     SI tA't5'I          5%.', LA Lt1-----II                                      I~     L.          I S~
bFeak flow flashi                                                                                                                                        i                                  h d    -      4k.-.
neebNeen                                                                                         hases              theI e"      5~
iem%5A.%l  I11f.        eA                                                             v..A...          F-1           .MOVII'54 .          .            -.tJA LI      1 ondenser hot well,.
(1)
Dosian- B6asisQ. Analyal Assumetbns Thne IIVjIO asumpionsLU1~                                          andU           roiiteteie Used ini the analysis~ Liiu LII in Table 15.5 64.
(2)              Sour-e Term calclain-s 1ne raalenuciiae concentrations in tne plr~amar ana sec)npjary system,
* I            I I
PrioF to ana toaiewin~i the SGR~e                                                a                 aeteFrminea as foiiows:
1ne :oalnc               la) concentratins in me reactor                                                        GoiaRlWll               wivD                asee upop
                                                                                                                                                   .......               . r    ......
()GGi eR Rlii d S ik ~~a5 G.-ere
                                                                                                                              " !h i ii l                                  i               -4 l      ie GOAr.eAtFatien is 1 ftGWgm of Dese E                                                                              i                             1,1
                                                                ..............             ....I    '  .. ... l  -.  ..   .     ...      ..   . .. L ...r    -- .   ..        ..            .  . ---..
Fel                                           !                                                                                i                      i                 it     h 15.5-107                                                                        Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE whic~h inrGeases the io~dine release rate ffrom the fuel to the coolant to a value 335 times greater than the release rat corresponding to the initial primnar; system iodine concentration. The initial appearance rate can be written as (1 5.5 15*
wheFe+
Pj-Equilibriumn appearance rate for iodine nuclide i A,- equilibrium RCS inventor,' of iodine nuclide GOrrespanding to 1 ltiG~gm of DE 1131 k,- removal coefficaient for iodine nuclide Pe F)   accident Spike A reactor transient has occurred prior to the SGTR and has raised the primary coolant iodine concentration from ito 60 j:Ei;igram of DE 1 131.
(b)     Thc initial secondar,' coolant iodine concentration is 0.1 ~I~iG'gram-(G)     The chemnical foFrm of iodine in the primar,' and secondary coolanti assumed to be elemental.7 (d) The initial no-ble gas concentrations inthe reactor coolant are based upon 651 pt~ilg of Xe-- 133 IDEC for the noble gasses Kr 85m, Kr 8:7, Kr 88, Xe 133, Xc 133m, Xe 135m, Xc 135, and Xe 138, using noble gas whole body dose conversion factors docGumented in FGR 12 (Reference 42) Table W'.1, associated with 1 percent fuel defects. The calculatio~n of Xe 13 DEC ignores the contribution fromn Kr 85 :and Xe 121 r due to low conentration and sinaI! dose convesionfactor.
(3) RadieaactMit' TrFanspo.d Analysis The iodine transport analysis, considers break flow flashing, steaming, and partitioning. The analysis assumes that a fraction of the ioedine carried by the break flow becomnes airborne immediately due to flashing and ato-mizeationR. The analysis consew.atively took no9 credit for crGubbing of-iodine contained in the atomized coolant droplets. The fraction of primary coo-lant i-d-ine which is not assumed to becomne airborne immediately mixes with the secondary water and is assumed to becomne airbore at a rate proportional to the steaming rate and the iodine partition coefficient This analysis Gensewatively assumes an iodine partition coefficient of 100 15.5-108                      Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE betweeR the steam generator liquid and steam phases. Droplet remova by the drier isco                tratively assumed to be negligible.
The folGWiRn assRumptiens and parameters were used t Gcal*;*ate the act~itiy released to the atmos6phere and the offsite doses folloWing a (a)     The massr- Of reac4tor coolant diGcharged into the 6econdar,' syste through the rupture an~d the mass of steam released from th ruptured and intact steam generators to the atmosphere are-presented On Table 15.4             1*.
(b)     The mass Of break flow that flashes to steam and is immediately released to the enVir*ornment is contained in Table 15.4 1                nd i presented inFigur -5.4.311. The break flow flashing fraction was GOnsevatively calculated asS Um~ng that 100 percent of the break flow i6fromR the hot leg side of the steam generator, whereas the break flow actually conS*st6 of flow from both the hot leg and cold leg sides of the 6team generator.
(c      hNo iodine scrubbing is credited for the break flow that flashes in the analysi6 and the iodine scrubbing efficiency is assumed to be 0 percent. Thus the location of the tube rupture is not significant for the radiological consequences. However, as discussed in Section 15.4.3.3, in the thermnal and hydraulic. analysis the tube rupture break flow is caIcUlated conser.atively assUminRg that the break is at the top of the tube sheet.
(d)      The      ruptur  /~
(OF. leakage)a    6ite is assumedr  to be      ~dwt la6Gg se...d..        w.ater b..ased. en Re-fe... rene 33, which concluded the effect of tube ulnovery     ';      essentially negligible for the rFadiologial nnOsequenRce      for the limiting SGT*R trans6int.
(e)      The total pFrimay to secondar,' leak           rate for the 3 intact steam generator-s isassumed to be 1.0 gpmn. The leakage to the intact 6team generator is arssumed to persist for the duration of the aGoident.
() The iodine partition coefficsient between the liquid and steam of th Fuptured steam generatoF             is assumed to be 100 foFrno      flashed flow and 1 for flashed flow. The iodine partition Goefficgcnt behveen the liquid and steam of the intact steam generator i6 assumed to be 44007 15.5-109                      Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE (E       The noble gases in the break flow and prmary to                         ,econdar leakage are assumed to be tranSferred instantly out of the stcam generator to the atmosphere. The whole body gamma doses are calculated combining the dose from the releae nolegses with the dose from the iodine releases..
(h)      For the accident initiated ioc            spike case, an iodin~e spikin~g faco of 335, obtained from Regul l*t* y Guide 1.195, May 2003 (Refe*re* e 44) is assumed.
(4) Offsite Dose Calcubation inequations, 15.5 17 and 15.5 18, no credit istaken fora cloud depletio by ground deposition or by radioactive decay during transpeot to the eXclusion area boundar; or to the outer bounda~' of the low population-zone. Off-site thyroid doses Wre calculated using the equation integrated activity of iodine       RUclide i relr ARPA  dwing he time o twiffl i   temV l j amm Zp
                   -bre-athing        rate duFrig timentera         in mnete 3Pe~epd-(T       1e1.*a 5        r. ,6!5.5)
                       -atmospheric       dispersion facGtGoduring time interalin
                        %Geodsimetee, (Table-I-5.568)
(DC;F=)     -     thyroid dose conversion factor via inhalation for iodine nuclide i in rem/Ci*(Table 15.5 69) w   -     thyroid dose via inhalation inrem Offcitee whole body gamma doses are calculated using,the equatien:
15.5-110                        Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE 0.25     i~,;rv
                                                          .x'IAD   NJ(15.5                             18) where:.
intefrated
                                                ;4AR4. ,      activity of noble gas,nulide  i rcleased during time inter.al j in Q; (Y 10)   -      atmospheric disperion fac.tor duing time intal         j seGOndrm0
______________               -average         gamma energy forF noble gas nuclide "in MeV~dis (Table 15.5 70
___________               11T              whole body gamma dose due to immeio               inRe (5)             ffitke Dose Re&UltS T-hyroid and whole body gamma doses at the- Excuin-h10- Area Boundary-and the outer boundary of the Low Population Zone are presented in Table 15.5 71. All of these RSG doses are within the allowable guidelines as specified by.the SRP, Revision 2 (SectionF 15.6.3).
The SGT-R dose analysis,Of record is based on the RSGs and al! doses arc within 10 CFR 100.11 limits. The limiting dose for the SGTR analysis-accepted by the NRC based on the OSGs, is the EAB zero- to- ihvo ho-ur-thyroid dose Of 30.5 rem for the accident initiated iodine spike analysis case. This, doeex-epeds; the- SRP 15.6.3- allo'.vable guideline value of 3 rem by 0.5 rem. However, the NRC found the 30.5 rem value acceptable-in a le~er to PG&E=, dated F~ebruary 20,2003, Isuneof Amendment RE:- Rev;isio to Technical Specification 1.1, 'Definitions, Dose E~quivalent-1 131,' and Revised Steam Generator Tube Rupture and Main Steam Line WreakAnalyses."
15.5.20.2.2 Control Room Exposures Aditenl nayrswere pefFndt eefln the airborne doese to the control room+
operators from an SGT-R. These calculations used the atmospheric releases of-radoativtydetermnined in the analysis discussed in Section 15.5.20.2.1 and-Reference 46. The control room is modeled as a discrete volume. The atmospheri dispersion factors calculated for the transfer of activity to the contro! r~oom intake contained On Table 15.5-6-8 -areused- to-determ~ine the activity availablep ;at the control 15.5-111                       Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE intake. The inflow (filtered and unfiltered) to the contro' room and the control room filtered r-ecirsUlation flow are used to calculate the concentration Of activity inthe control room. Control room parameterS used in the analysis are peresented in Table 15.5 72.
The controlI roo occupancy factors assume.d were taken from Table 15.5 32.
Thyroid, whole body gamma, an;d- beta skin doses                             &#xfd;are   cluae for 30 days, in the control room. A.lt.hough all release,.                 are terminated when the R+HR system i.. put in seric, the calculation iScontinued to account for additional doses due to continued GGGUpaRGl.
The total primary to.econdary                    leak rate is, assumed to be 1.0 gpm. The leakage to the 4ntact steam generator is assumed to persist for the duration of the accident.
The calculations determine the thyroid doses based on a                                   cient      lr iodine spike and based on an acident initiated iodine spike with a spiking far of 335. Both spike assumptions consider 0.1 1ACi/gm D.E. 1 131 scdaryati'~it'. The whole bo~dyj doses are calculated GGombining the dose fro~m the released noble gascs with the dose fromn the iodine releases.
 "nntroi roomrnmron-c aoozeq are a;;ilcU~ated using the following equation:
I 14=r" i:
E
:1
                                                                                                                                    "",\ I IU I  hi                             V                                                             \ .-..-...
wheFe4 DT,-      thyroid dose Yia inhalation; (Rem) 1(F -     thyroid dose conyerion factor via inhalation; for isotope i (Rem/Ci)
(Table 15.5 Q GAq-      concentration in th-e control roomA of isotope i, duFrig timeinea!,
G--iG-mlJU~la dependeemmu       IIenI lp   inlle-ka-    e,,  IllU    I-I-I-I.I-Jilml,              mt IIGFGIItonand
                       ;Fewth..
                         .... ,, g         d n tseo/m                                                      6
                 -ID\ brah ing;.. rat du"*.ring time ;*        .intera  j ....... ) (Table.1.5.8 15.5-112                                    Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE Control room whole body doses are calculated using the following-equation:
             *'w="'J B              GF)     V_ F i                                                  '.(15.5.   .
wheFe.
D B- whole body dose via cloud immeFsion (Rem)
GEI      - geomnetry factor, calculated based on Reference 17, using the equation s
CIF - 1173 where
                                  -f V.' is the coto room ,o..,     in f.*e 7v .338 VJ       I (Table 15.5 70)                                                              l w
             -    concentration in the contri room ot isotope 1,during time intea*i J,               I caicUiatea aepenacnt. upon iniear~ace, Iitered recircUIation ana Tlietcre i
                   ~G~-sem~
Control room skin doses are calculated using the following equation (15.521)
               /3                   -3     ii1 wheFe Do   -    whole body dose via cloud immersion (Rem)
         ---      average beta d s:ntegration energy for isotope i (Mevddis) (Table 15.5 70)
         &#xfd;-  concentration in the cont.O. room of isotope i, du.rig time         .........       j, calc~ulated dpen~dent upon inleakage, filtered recircGulation and filtercd infg    Nw-(secimn Table 15.5 74 presents the resulting airborne doses to the control room operators. The resultant doses are well below the guidelines of G.. 19, 1971, and are below the correspni             post LOCA control room, expo.sures presented in Table 15.5 33 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                      Revision 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/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 SGTR at either unit are provided in Table 15.5-64B. The X/Q values presented in Table 15.5-64B take into consideration the various release points-receptors applicable to the SGTR to identify the bounding x/Q values applicable to a SGTR 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 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                   Revision 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 rem) TEDE as shown in Table 15.5-64.
(1) An ind-i'idual loc-ated at an~y point on the boundar,' of the exclusion area forth two hu
* immeditely following the onset of the postulated fis+ion pr*du* t release shal;l
                       , no-t reei*iige a total radiation dose in e         dose guidelines of 9cessof SRP, Section 15.6.3, Rovision 2 (i.e., the 10 CFR 100.11 dose limits for the wAhole body and the thyroid for the pre existing iodine spike case and 10 penrent of the 10 CFR 100.11 dose lifrits forthe whole body and the thyrFid for the accident initiated iodine spike case) as shown in Table 15.5 71.
(2) An individua! located at any point on the outer boundar,' of the low population-zonc, who is exposed to the radioactive cloud resultin~g from the postulated fiss6ion pro-duc9t- release (during the entire period of its passage), shall not receive a total radiation dose in exce6s of the dose guidelines of SRP, Section 15.6.3, Rev.'ision 2 (i.e., 10 CFR 100.11 dose limits for the whole body and the thyroid for the pre existing iodine spike case and 10 percent of the 10 CFR 100. 11 dose limits for the whoe body and the thyroid for the accident initiated iodine spike case) as shown in Table 15.5 71.
(3) in accordanc~e with the requiremenits of GDC 19, 1971, the dose to the conro room operator under accident conditions shall not be      In   exGess of 5 rem whole body or its equiv:alent to an~y pa~t of the body (i.e., 30 rem thyroid and bheta skin, Reference 51) for the duration of the accident for both the pre existing and the accident intae         oiespike cases as shown in Table 15.5 7-4, As noted in Section 15.5.20.2, the above dose estimates reflect the RSGs and are within 10 CFR 100.11 limits. The SGT-R an~alysis accepted by the NRC based on 08G6 is the EAB zero) to two hour thyroid dose Of 30.5 rem for the accident initiated iodine spike aralysisgcase. This dose exeeds the SRP 15.6.3 allowable guideline value of 30 rem by 0.5 rem. However, the NRCsfoud the 30.5           rem value acceptable on a letter to PG&E=, dated F~ebruary 20,2003, "Issuance onf Amenpndmennt: RE:-= Revisiqogn ton Technical Specific~ation 1.1, 'Definitions, IDose Equivalenit 1 131,' and Revised Steamn Generator Tube Rupture and Main Steam Linie IBreak Analyses."
15.5.21 RADIOLOGICAL CONSEQUENCES OF A LOCKED ROTOR ACCIDENT 15.5.21.1       Acceptance Criteria The radiological consequences of a LRA 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 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.
The radiological consequences of a locaked rotor accident shal! not exceed the dose limits of 10 CFR 100.11 as outlined below:-
An individual located at any point on the boundar,' of the eXclusion area for the tw.o hoUriMmeit           following the onset of the postulated fission product release shall not (dureceia total radiation dose to the whole body in excress of 25    rem or a total radiation dose in eXcess of 3000 rem t the thyoid from iodine exposure.
An individua! !ocated at any point on the outer boundasy of the low population zone, who i5exposed to the radioactive cloud resulting from the postulated fisaIon product release (during the entire period of its passage), shall not receive a total radiation dose to the dWhule body in excess of 25fem, or a total radiation dose 6in   eced    ouf 300 areu to the thyroid from iodine exposure.
hi accredance with the     requirements of GDC 19,e1971, the dose to the     contry!rom operater under accident Gonditiens  lhall    nst eien pdexcs Of 5rem whole body Or its equivalent to any part of the body (i.e., 30 rem thyroid and beta skin, Reference 51) for the duration of the accident.
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 primary 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                    Revision 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      -ircUmstanc-s,            a locked ro9tr aGridelnt could cause sm-all a*mnG,,tr Of fu-el cla-;dding failure in the core. if this occurFs, some fission products will enter the coolant and will mostly remain in the 9oolant Uil cleaned up by the p*imar, coolant demineralizerns, or in the case of noble gases, untoI stripped from the coolant. Follonwing such an incident, there are several possible moedes Of release of some of this actiVity to the enviomnment.
in the short termn, if the accident occurs at a time when significant prmnar; to secondar,'
secondar,' system. The Roble gases will be discharged to the atmosphere via the air ejectors or by way of atmospheri steam dump. The iedines Will remain moestly in the liquid form and be picked up by the blowdown treatment system. Some fraction of the-iodfines, howe..er, wil! be released via the air ejectors Or by way of atmospheric steam dump. Inaddition, if an atmospheric steam dump is necessary, some of the activity contained in the secondar,' system prior to the accident will be released.
The amounts of stea~m released dependAon the- time relief valves rmiopnand the availability Of condenser bypass- coGolig capacaity. The amounts of rdocieiodine release deen n the amounts of steamn released, the amoeunt Of activity coentained in the secondar,' system prior to the accident, and the amount contained in the pia, coolant which leaks into the sec..d..                st..em. As d.iscussed in Section 15.5.10, the amount of 6teamA released following the locked rotor accident, if no condenser cooling is available, would not exceed approximately I .7E+06 Ibm.. in the analysis of both the design basis case and the expected case, this amount of steam was assumed to be Regulatory requirements provided for the LRA in pertinent sections of Regulatory Guide 1.183 including Appendix G is used to develop the dose consequence model.
For the design; basis-case, it was assumed that the plant had been operating continuously with 1 percent fuel cladding defects and 1 gpm prImaF to SecOndar,'
leakage. For the expected case calcul-ation, operation at 0-.2 pecet defectsan 20 gallons per day to the secondary was assumed. in both cases, leakage of water fromn pimnar,' to secondar,' was assumed to continue during cooldown at 75 percent of the pre accident rate during the first 2 hours and at 60 percent Of the pre accident rate UU,:rlg twe     nextO iiuui.
HA          -hese    values Were ueiiveu ii-ur     I~;  to-%enuapfy pressure differentials during cooldown. it was also conscrwatively assum~ed for both cases that the iodine Partition FactoWr in the steam generators releasing steam was 0.01 on a mAsbai (Reference 15). Inaddition, to account for the effect of1 iodine spiking,-
fiul esc,,-,-nape rFate coffic-H-,ie-n6tr for iodines of 30 times, the nrmal operation values given in 15.5-117                    Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE Table 11 .1 9 were used for a period of 8 hours following the etant of the accident.
Other detailed and less Signifcant modeling asrsumption arpeented in Reference 41 The asupin used for meteorology, breathing rates, population den~sity and other common f-actorsq were also described earlierF. Both thec prr, an 6econdar, coolant acti'.itie pro to th.e acc~ident are discussed in Section 1S 5 2.
Inorder to deteFrmine the primnar; coolant activities immediately after the accident, it was assumed that less than 10 percent of the total actiVity contained inthe fuel rod gaps would be immediately, released to the coolant and mixed uni;formlY in the coolant systemn volume. The gap inven~tories used are losted in Table 11.1 :7.
All of the data and assumptions listed above were used with the EMERALD computer progr-am to c*cIIate the activity rIeeases and potet*Ial doses following the accident.
The calculated activity releases are listed hn Table 15.5 41. The potential doses are in Table 15.5 42. The ex(posures are also shown in Figures 15.5 14 and 15.5-15 givenfunction as a:4          of the amount of fuel failue*that occurs. On the left bounda. y of these graphs, in the region of negligible fuel failures, the exposures are just the comnponent resulting from the aGti*ity already pFesent in the secondary system, Or which leaks throu1gh the steam generators at pre accident primar '* c           leveIs. These exposumes Folant correspond to those shown in Figures 15.5 2 through 15.5 5.
HISTORICAL INFORMATION IN ITALICS 1BELOWA.I NOT REQUIRED TO BE REVISED.
Another moede of release felwn a locked rotoF r.iet              or any"G~e        novn signific.ant fuel failure, is the long term rele.ase by way of cleanup and leakage from the pmaY coolant system. The activity going through these pathways, principally Kr 85, Woluld1 result :n som inreetal long termn dose beyond the normal yearly releases.
This pathway of relearse has been evaluated, and the results are presented int Fig9ure 15.5 16. Since the activity released in this way would reac-h t-he- en-vironment over a long termn, the annual average atmospheric dilution factors (Table 15.5 5) and breathing rates, have been used. The amounts of activity released were determ~ined by multiplying the activities released from the gaps following the accident by the release fractions iste in TableIP 15.5-40-.
These long te*rm releae frActions were determined from the nRormal radiOactivity transpeot analysis carried out for Chapter 111, for the anticipated operational occurrences case. in essence, therse fractions are the fractions Of acurie reaching the envir-onment per curie released to the cooelant, for each isotope. The pahY Ineluded Wear Fn~
Table 15.5 40, essentially all flt.h K*r 85 eleased to the coolant il eventually released to the environment, as would be physically expected, and lower fractions Of the other isotopes are released, depending on their respective overall cleanup, leakage, and decay facto~rs in the plant. It can bhe cnclnuded by com~paring these exposures tO the shodt term eXposures in Figure 15.5 12 that the incemmental long term exposures are negligible additions to the radiological consequences of accidents-o-f this kind.
15.5-118                    Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE In addition, it can be cOncudId that accidents of this kind would not result"in significant additions to the annual dosees expected from norma! plant operation.
From these short teFrm and long term analyres, it can also be concluded thatal poteRta! expoasWes from a locked rotor accident will be well below the guideline levels specified in 10 CFR 100.11, and that the occurrence .f such accidents would not result inundue risk to the public. A detailed evaluation of potential exposures to coentrol room personne.l .as. Made in Section 15.*.47, for  .onditi"ns  f"lo.win.g a LBLOCA. Th con-ntaninment shine contribution to control room dose weuld not be applicable following a locked rotor accidenit.
The LRA is postulated to result in 10% fuel failure resulting in the release of the associated gap activity. As discussed in Section 15.5.3.1.3, the core gap activity is assumed to be comprised of 12% of the core 1-131 inventory, 30% of the core Kr-85 activity, 10% of the remaining noble gas and halogen isotopes, and 17% of the core alkali metals (Cesium and Rubidium). 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                    Revision 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.
15.5.21.2.3    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 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=10.73 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.
Timing 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 Dispersion 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 MSSVs/10% 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                   Revision 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 X/Q 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 2.3-147 and 2.3-148.
The bounding control room dose following a LRA at either unit is presented in Table 15.5-42.
15.5.21.3         Conclusions By comnparing the acti',itY,releases following a !ocked rotor accGident, given On Table 15.5 41, with the ac-tivity releases calculated for a LBLOCA, given in Tables 15.5 13 and 15.5 14, it can be concluded that any control room; exposures following a locked rotor ar.Gident will be well below the GOC 19, 1971, criterion level.
Additionally, the analysis demonstates that the acceptance criteria are met as follows:
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 rem) TEDE as shown in Table 15.5-42.
(1) The radiation dose to the whole body and to the thyroid of an individual lct at any point on the boundary of the exclusion area for the two hou~rs immediately following the onset of the postulated fission product release are well below the dose limits of 10 CFIR 100. 11 as shown in Table 156.5 42.
(2) The radiation dose to- the w~hole body and to the thyro-id Of _An    indi;Fidual loatedpl at any point on the outer bounda.y of the low population zone, who is exprEod to the radioactiVe cloud resulting ferom the postulated fission product release (duFring the ei          d of its passage), are wAell below the dose limits of 10 rCFR 100 11               n T-:blp 16 fi-4 O~eaw 15.5-121                    Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE (3) Since thc actiVity releasee from the !ocked rotor accident given in Table 15.5 41 are less, than thore_ from a LB1LOCA (see Table 15.5 13 and 1 5.5 14), any control           . which m;ight occur would be well within thestahl;hhed, room dose criterki of GDC 19, 197-1 and discu-ssed in Section 15.5.17.
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           Fuel Handling Accident In The Fuel Handling Area 15.5.-224.-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                   Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE The radi;olgiall       coRequenRes of a fuel handling accidentn- the fuel handling area shall not eXceed the dose !imits of 10 CFR 50.67-as outlined below:.
(1) An individual located at any point on the boundar; of the exclusion area for any two hour period following the ontet of the postuated fission perout relea. e shall not Feeive a tetal radiation dose in excess of 0.063 V (6.3 rem) to05t5a1  effctivye dose equivalent (TEDE).
(2)  An individual located at any pient on the outer boundary of the low population zone, who iS eoed nt the Fadioactive cloudresultirg from the postulated fissiont     productrelease (during the entie perioed of its passage), shall net receive a total fadiation dose in exfess of 0.063 SY (6.3 rem) total effective dose equivalent (:TEDE).
(3)  The dto e to the control roeem perator under acrident conditions shall not be in eXcess of 0.05 SY (5 rem) total effective dose equivalent (TEDE) for the duration; of the accident-.
15.5.224.2         Identification of Causes and Accident Description 15.5.22.2.1         Activity Release Pathways This event postulates that a spent fuel assembly is dropped during refueling in the Spent Fuel Pool (SFP) located in the FHB, or in the reactor cavity located in the Containment. All of the fuel rods (264 rods) in the dropped fuel assembly are assumed to be damaged; thus all of the activity in the fuel gap of the dropped assembly is assumed to be instantaneously released into the SFP or into the reactor cavity. As documented in the NRC SER for Amendments 8 and 6 to DCPP Facility Operating -
License Nos. DPR-80 and DPR-82, respectively (Reference 87), the assumption that all fuel rods in one assembly rupture is conservative because the kinetic energy available for causing damage to a fuel assembly dropped through water is fixed by the drop distance. The kinetic energy associated with the maximum drop height for a fuel handling accident is not considered sufficient to rupture the equivalent number of fuel rods of one assembly in both the dropped assembly and the impacted assembly.
During fuel handling operations, containment closure is not required. Generally, the containment ventilation purge system is operational and exhausts air from the containment through two 48-inch containment isolation valves. These two valves are connected in series. This flow of air from the containment is discharged to the environment via the plant vent.
This exhaust stream is monitored for activity by monitors in the plant vent. In the event of a postulated fuel handling accident, the plant vent monitors will alarm and result in the automatic closure of containment ventilation isolation valves. This activity release may result in offsite radiological exposures.
15.5-123                     Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR 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 credited 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.
The radiological con"equences of a fuel handling accident in the fuel handling area were analyzed using the LOCADOSE computer code..
The values assumned for individual fission product inventories are calculatedfoa source termn assumn          approx.Im4ately 105 percent ful! power operation (3580 M thermnal) mmdteypreceding shutdown. The accident is assu~med- to occur-I 100 hours a~ner shutdown. Thnis t-an-er interwai represents approximately the Fm~inmum tme requirFea to preae (colon,head and internlals removal, cavity flooding, etc.) the core for refuelin an            Fsteefore 6omewhat GGnseryative in that it would require thatth accident occur during handling of the first few fuel assemblies.
The source termn ISGonsewatiyely assumed to be a composite of the highest fissio product activity totals, for various, combinations of burnup and enirichment. The. ORIGE=N 2 cmpte cdewas used to calculate these worst case fission product inventories.
TheDBAgaacivty nvnto 3 'is based on NRC Safet Guide 25, March 1972-,
assumptions:' radial peaking factor Of 1.65, gap ftraction of 10 percent for noble gases other than Kr- 85, gap fraction of 30 percent for Kr 85, and gap fraction of 10 percent for edines.
The assumption is mnado for both cases that 100 percent of the activity (consiSting princip.ally of fission product isotopes of the elements, xenon, kr,'pten, and iodine)+
present in the gap betw.een the fuel pellets and the cladding in the damaged rods is immediately Felcased to the pool Or cavity water. This assumption is Gonse~vative for 15.5-124                    Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE elemental iodine because the lOW cladding and gap temperatures would result in a large fraction of it being condensed and temporarily retained within the cladding.
The analysis assumes that the fission produc-'t release occur~s at a water deptho 23 feet, whc       s t he miimmater depth above the top of the fuel as required by TecGhnicGal Specifications. The spn fue p ool, where handling operations are most likely to- result infuel damage, has a wate depth of about 38 feet. Using a depth of 23 feet accounts for cases in Which the release occurs fromB the top of an aGsembly that is resting Ye~tically on the floor, and for releases that occGur near the top of the storage racks. Finally, consistent w'ith Safety Guide 25, March 197-2 the analysis assumes that all activty that escapes from the pool to the fuel hand lin araaipaces iGreleased fromn the area within a 2 hour timne period.
Of the activity reaching the water, 100 percent of the noble gases, xenon and krypten, ace assumed to be immediatel" released to the fuel handling areaiair saes. Howe           are the abiity of the po!o waterto scrub iodine from      the gas bubbles  asthyrse    st    the surfacue has been considered. The pool Dos for the iganic and oani pies aoe 500 and 1, fespectivel, givsing an overall effective DF of 200 (i.e., 99.5 percent of the total released from the damaged rods is retained by the pool water). This difference in DFe fora rgani and organic iodine species       Fesults in the iadit e above the fue! pool being composed of 75 percent inorganic and 25 percont      loganic species. These assumptions are consistent with those suggested in NRC Regulator; Guide 1.183, July 2000. Table 15.5 44 itemizes the gap activity available for release nfo ae            h the atmosphere to the eirnnt-.
Table 15.5 45 itemizes the assumptions and numlerial values used ralcuflateto           the fuel handling accident radiological exposures. The potential releases of activity to the atmosphere are listed in Table 15.5 44. The exposures resulting fromn the postulated fuel handling accident inside the fuelhandling area are presented in Table 15.5 17.
These exposures a5e well below the RegulatRe' Guide 1.183, July 2000 limits and demonstrate the adequacy of the fuel handling safet 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.
Although conservatively neglected for this analysis, the he 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                     Revision 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.
After charcoal filter cleanup (another design feature     Ggnse~vatively neglected inthis analys is), fuel handling area post accident ventilation air exhausts through the plant vent at a height of 70 mneters. Site mneteorology iSsuch that it is ver,' unlikely that any airborne actiVity will en~tcr the control roomA ventilation system.
Spent fuel cask accidents in the fuel handling area causing fuel damage are prFe*luded due to cran~e travel limits and design and eperating features as descrfibed in Sections 9.1.4.3.9 and 9.1.4.2.6. Spent fuel handling accidents in the fuel handling area would 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 12% of the core 1-131 inventory, 30% of the core Kr-85 activity, 10% of the remaining noble gas and halogen isotopes, and 17% of the core alkali metals (Cesium and Rubidium). 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 15.5.22.2.2       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.
15.5-126                    Revision 19 May 2010}}