ML052380407

Comanche Peak, Units 1 and 2, Mark Up of FSAR

ML052380407
Person / Time
Site:	Comanche Peak
Issue date:	08/22/2005
From:	TXU Power
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
00236, CPSES-200501556, GL-03-001, TXX-04127
Download: ML052380407 (90)

Text

Attachment 6 to TXX-05127 Page 1 of 89 ATTACHMENT 6 to TXX-05127 Mark up of FSAR Pages IA(B)-2, 7, 37 1A(B)-61, Insert for Page 61 6.4-1, 6,8,9 Table 6.4-1, Table 6.4-3 Table 6.44 6.5-16 Table 6.5-5 Table 6.5-6 Table 6.5-7 15-xi, xii, xiii Table 15.0-7 (2 Sheets)15.1-15, 16 Table 15.1-3 (3 Sheets)Table 15.1-4 Table 15.1-4A 15.3-7, Insert for Page 7 15.3-8 Table 15.3-1 15.4-28 thru 31 Table 15.44 (4 Sheets)15.6-3, 4, 10,11,17 thru 21, 23, 24, 26 Table 15.6-2 (2 Sheets)Table 15.6-3 Table 15.6-4 Table 15.6-9 (3 Sheets)Table 15.6-10 15.7-2 thru 6 Table 15.7-1 (2 Sheets)Table 15.7-2 Table 15.7-3 (3 Sheets)Table 15.7-4 (2 Sheets)Table 15.7-6 Table 15.7-7 (2 Sheets)15B-1 15B-2, Insert for Pagel5B-2 15B-3, Insert for Page 15B-3 1SB-4, Insert for Page 1SB-4 15B-6, 7 Table 15B-1 (2 Sheets)Insert for Table 15B-1 Attachment 6 to TXX-05127 Page 2 of 89 CPSES/FSAR Regulatory Guide 1.4 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors.Discussion The analysis of the radiological consequences of the loss-of-coolant accident presented in Section 15.6.5 c omC*ie tl h

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• m 2(6,'4) o thi, mes *gaide~ l.\J. sV111Y Mll IWLl119 *dwitiil 1111hibuivi is hv9VIPIV1 ilt MCCUI i-J th611 k,uvvy t i of i~liulW bilt U WJAp~.J.)Ui%.lu.

Ifollows the guidance provided in Regulatory Guide 1.195 instead of Regulatory Guide 1.4.Regulatory Guide 1.5 Assumptions Used for Evaluating Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors.Discussion This regulatory guide is not applicable to the CPSES which has pressurized water reactor steam supply systems.Regulatory Guide 1.6 Independence Between Redundant Standby (Onsite) Power Sources and Between Their Distribution Systems Discussion The CPSES design complies with the requirements of Safety Guide 6 (3/10/71).

For details see Section 8.3.Regulatorv Guide 1.7 Control of Combustible Gas Concentrations in Containment Following a Loss of Coolant Accident Discussion The CPSES design of the hydrogen recombiners and Hydrogen Purge System meet the requirements of Revision 2 (11/78) of this regulatory guide as discussed in Section 6.2.5, with the following exceptions and justifications.

Part C.4 The CPSES design takes exception to this regulatory position by de-classifying portions of the Hydrogen Purge System (HPS) which contain filters to seismic category II. This means that the filter unit is not required to be functional after a seismic event but it must remain in place. This is consistent with the accident scenarios postulated at CPSES. A LOCA is not postulated to occur coincidentally with a seismic event.1A(B)-2 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 3 of 89 CPSES/FSAR Refer to Appendix IA(N).Regulatorv Guide 1.23 Onsite Meteorological Programs Discussion The meteorological monitoring program at CPSES complies with the requirements and those applicable recommendations of the Second Proposed Revision 1 to Regulatory Guide 1.23 (April, 1986) as discussed in Section 2.3. Refer to the CPSES Offsite Dose Calculation Manual (ODCM), Section 1, 4.3.3.6 and Bases 3/4.3.3.6 for an exception.

j Refer to Section 2.3 for a description of the design and siting of the primary meteorological tower. The quality assurance program for meterological monitoring is identified in FSAR Table 17A-1 and Section 17.2.Regulatory Guide 1.24 Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure Discussion The analysis of the radiological consequences of the radioactive gas storage tank failure accident presented in Section 15.7.1 complies with the requirements of Safety Guide 24 (3/23/72) except that only gamma radiation contribution is taken into account in the determination of whole body exposures and the dose calculation methodology is consistent with Regulatory Guide 1.195.Regulatory Guide 1.25 Assumptions used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors Discussion The analysis of the radiological consequences of the fuel handling accident insidtlh, F.,! -ri~in presented in Section 15.7.4 Wulla,,i "ith kit maitemenk.

25 (MC it Z Ifilu VJ a. l follows the guidance provided in Regulatory Guide 1.195linstead of that in Regulatory Guide 1.25.2H 811!jy E"111111" sAvilotll ul ulatibatwuivl ia takell it.VIucVuLi 11k il th detc."inilatioll uf Regulatory Guide 1.26 Quality Group Classifications and Standards for Water-, Steam-, and lA(B)-7 Amendment 97 February 1, 2001

-Attachment 6 to TXX-05127 Page 4 of 89 CPSESIFSAR Design Basis Tornado for Nuclear Power Plants Discussion The CPSES is designed to conform to the requirements of this regulatory guide, dated April 1974, except that it is designed to withstand the effects of a Design Basis Tornado having a maximum wind speed of 360 mph which is made up of a rotational speed of 300 mph and a translational speed of 60 mph. A simultaneous pressure drop of 3.0 psi at the rate of 1.0 psi per second is considered.

The Design Basis Tornado for CPSES was determined prior to the issuance of this Regulatory Guide and was approved for use by the Atomic Energy Commission's Safety Evaluation Report Dated September 3, 1974.Also refer to Section 3.3.Regulatorv Guide 1.77 Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors Discussion The analysis of the radiological consequences of a control rod ejection accident presented in Section 15.4.8 coA..i. EIitL tihe L4UI U ,f ta 1. PFIIeA e B iOull. UF also I bitv .*ide, ate Ma i 9?4 exep un. -i.11U. Vhi.t1 -s. *.s .awrif p"%A1UL&1t si flM. J 1 6, 1 1 1 istiUv 1 O+ a ul %uL ^ V.' follows the guidance provided in Regulatory Guide 1.195 instead of that in Regulatory Guide 1.77.Regulatorv Guide 1.78 Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release Discussion The CPSES design meets the intent of this regulatory guide, dated June 1974, as discussed in Sections 2.2 and 6.4.Regulatory Guide 1.79 Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors Discussion The Initial Test Program, as described in Section 14.2, is in compliance with the provisions of Revision 1 (9/75) of this regulatory guide with the following exception.

lA(B)-37 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 5 of 89 CPSES/FSAR This regulatory guide is not applicable to CPSES.Regulatory Guide 1.146 Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plant Discussion Effective June 1, 1983, CPSES Quality Assurance Audit Personnel are in compliance with regulatory guide 1.146, August 1980.Regulatory Guide 1.148 Functional Specification for Active Valve Assemblies in Systems Important to Safety in Nuclear Power Plants Discussion CPSES is not committed to this Regulatory Guide. The operability program for active valves is covered in Section 3.9N and 3.9B.Regulatory Guide 1.150 Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations.

Discussion Refer to Appendix I A(N).Regulatorv Guide 1.155 Station Blackout Discussion CPSES complies with the guidance of Regulatory Guide 1.155 (August 1988) as described in Appendix 8B.Regulatory Guide 1.163 Performance-Based Containment Leak-Test Program Discussion Effective by August 12, 1996, CPSES complies with Regulatory Guide 1.163, September 1995.I4 IA(B)-61 Amendment 97 February 1, 2001

-Attachment 6 to TXX-05 127 Page 6 of 89 Insert A (after page 1A(B)-61)Regulatorv Guide 1.195 Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light- Water Nuclear Power Reactors, May 2003.Discussion The radiological consequences analysis for the design basis accidents follow the guidance provided in Regulatory Guide 1.195 and CPSES adopts the dose limits defined by Regulatory Guide 1.195. The CPSES design for skin dose calculations DCFs uses DOE/EH-0070 instead of Figure 12 recommended in Regulatory Guide 1. 195.Rejulatorv Guide 1. 196 Control Room Habitability at Light-Water Nuclear Power Reactors, May 2003 Discussion The CPSES control room habitability program follows the guidance provided in Regulatory Guide 1.196 with the following exceptions:

1. An acceptable alternative to the specific TS change identified in RG 1.196 is used which includes CRE integrity testing and periodic assessments.

A Control Room Integrity Program is implemented.

2. CPSES currently is using RG 1.52, Revision 1 for the control room design. CPSES used Revision 2 for testing only.3. CPSES uses RG 1.140 as information only for non-safety related air filtration systems.Regulatory Guide 1. 197 Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, May 2003 Discussion CPSES uses RG 1.197 dated May 2003 for testing of the CRE. CPSES has successfully completed comparison testing as described in section 1.2 of RG 1. 197. Individual component tests will be performed in the future to determine the total CRE leakage as necessary.

If future design changes warrant another ASTM E741 test, they will be performed with the exceptions as defined in NEI 99-03 Revision 1, Appendix EE.

Attachment 6 to TXX-05127 Page 7 of 89 CPSES/FSAR

6.4 HABITABILITY

SYSTEMS 6.4.1 DESIGN BASES 6.4.1.1 Control Room Envelope The Control Room pressurized envelope, as defined in Section 6.4.2.1, includes the Control Room and all areas adjacent to the Control Room on elevation 831' 6" of the Electrical and Control Building containing plant information and equipment that may be needed during an emergency including kitchen, sanitary facilities, and computer rooms.Control Room design is based upon the safe occupation of the Control Room envelope during normal operation and for a period of not less than 30 days after a loss-ofcola 5 accident (LOCA). Habitability systems ensure that th oentol Room during these times will n o ra iatio xeeding 5 rem whole body gamma dose, 4 rem thyroid dose, and eeta skin dose. The allowable unprotected beta skin dose may be increased to 75 rem when special protective clothing and eye protection is used. The Control Room is designed in accordance with NRC General Design Criterion (GDC) 19 [1]. The Control Room envelope contains adequate medical supplies and the necessary kitchen and sanitary facilities to sustain plant personnel for a period of 30 days following a DBA. The necessary food and water for five plant personnel for five days will be permanently stored in the Control Room.6.4.1.2 Radiation and Toxic Gas Protection Control Room shielding is designed to limit the dose from external sources to a level compatible with the dose criteria given in Subsection 6.4.1.1 based on the inventories given in Table 6.4-1. jj The Control Room HVAC system is designed to maintain a positive pressure with respect to the environs during normal and emergency modes of operation.

Airborne radioiodine is limited to levels compatible with the dose criteria given in Subsection 6.4.1.1, based on the radioiodine activities given in Table 6.5-6, and a Containment leak rate of 0. 1- percent for the first 24 hr following an accident and one-half this value for the balance of the accident.

Refer to Subsection 15.6.5.4 for an i analysis of the inhalation dose to the Control Room operators.

In the event of a toxic gas release, the control room may be manually isolated from the outside environmentoby entering a complete recirculation mode of operation.

For CPSES, the probabiliyo simultaneous occurrence of a toxic gas release and radiological release caused by a loss of coolant accident (LOCA) is assumed to be extremely low. Therefore, the event of concurrent releases is not considered in the design basis. See Section 2.2.3 for a discussion of toxic gas releases and analyses.Airborne radioactive material in the Control Room atmosphere is controlled after an accident by the emergency recirculation filtration units and emergency pressurization 6.4-1 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 8 of 89 CPSESJFSAR 6.4.2.5 Shielding Design The shielding design for the Control Room is based on the requirements specified in 10 CFR Part 50, Appendix A, GDC 19. The Control Room is designed to provide radiation 50 protection for personnel occupancy under accident conditions so that no indivd i receive exposures in excess of 5-rem whole-body gamma dose, erem thyroid dose and 75-rem unprotected beta skin dose (with special protective clothing and eye protection).

To achieve this goal, the shielding design of the Control Room considers airborne contaminants within the Control Room and other DBA sources of radiation.

Specifically, these other sources are fission products released to the reactor Containment atmosphere, airborne radioactive contaminants surrounding the Control Room, and sources of radiation caused by potentially contaminated equipment in the vicinity of the Control Room. They are considered to be dominant sources of radiation, and they are among the principal parameters for the shielding design of the Control Room.Shield thicknesses of structural concrete provided for the Control Room are shown on Figure 12.3-14. The 2 ft-0 in. structural shielding walls surrounding the Control Room, combined with the roof and floor slabs above the Control Room, provide more radiation protection for personnel in the Control Room. In addition to shield thicknesses, distances that separate dominant radiation sources from the Control Room are included on the scaled layout and arrangement drawings of the facility in Section 1.2. Radioactive decay for each isotope of the DBA source is taken into consideration in the analysis of the dose to Control Room occupants shown in Subsection 15.6.5.4.

A layout drawing of the Control Room and associated structures is presented on Figure 1.2-33.DBA sources of radiation surrounding the Control Room and shielding related considerations are presented in Section 12.2, Section 12.3, and Subsection 15.6.5.4.

A plan view drawing of the Control Room and associated structures identifying distances and shield thicknesses is shown in Figure 12.3-14.6.4.3 SYSTEM OPERATIONAL PROCEDURES The following modes of operation characterize the Control Room Air-Conditioning System: 1. Normal operation 2. Emergency recirculation

3. Emergency ventilation

4. Isolation (emergency recirculation without pressurization)

The Control Room Air-Conditioning System is automatically switched to the emergency recirculation mode upon receipt of signals as outlined in Section 9.4. In addition, the system may be manually switched to the emergency ventilation mode of operation by the operator from the Control Room ventilation panels. This feature enables the removal of 6.4-6 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 9 of 89 CPSES/FSAR ua riGuide { r',7 6.4.4.3 Evaluation of Heating, Ventilation, Air-Conditioning, and Filtration System The HVAC and Filtration System readiness is ensured by the periodic testing program described in Section 6.4.5. Safe operation is ensured by having redundant equipment for the Control Room HVAC and Filtration System. A complete safety evaluation is given in Section 9.4.6.4.5 TESTING AND INSPECTION Preoperational tests are conducted on the Control Room HVAC and Filtration System to ensure that all equipment satisfies the design criteria during all modes of operation.

est are also performed, as described in Section 9.4,t-o ensure overall system performance.

The leakage tests will be conducted by closing all the access points to the Control Room.Control Room pressure will be established by controlling the outside air intake flow of the emergency pressurization units until the design pressure is achieved.

Should the outside makeup airflow through the emergency pressurization unit exceed the maximum allowable flow of approximately 800 scfm, a survey shall be conducted to locate points of excessive leakage and attempt to seal them. Tests shall be repeated as often as necessary until the above criteria are established.

Control Room pressure is measured by the permanently installed differential pressure transmitters.

The result of the Control Room leak test is considered acceptable if the emergency pressurization airflow does not exceed 800 scfm with the Control Room envelope being maintained at 0.1 25-in. wg.Planned leakage tests will be performed to verify that adequate pressurization of the Control Room envelope is maintained during Emergency Recirculation mode. The Control Room envelope will be maintained at 0.125-in.

wg positive pressure relative to the outside atmosphere with a maximum makeup airflow of approximately 800 scfm.In-place testing of air cleaning components will be performed in accordance with test methods and acceptance criteria described in ANSI N510 [7].Control Room equipment will also be tested in accordance with the methods described in ANSI N509 [6].Testing requirements, acceptability and frequencies for ESF components are established in the Technical Specifications.

< -----6.4.6 INSTRUMENTATION REQUIREMENTS Sufficient indications are provided in the Control Room for the operator to monitor HVAC system prane Annunciators indicate HVAC system or component F6.4-8 Amendment 97--- -IFebruary 1, 2001 Attachment 6 to TXX-05 127 Page 10of89 CPSES/FSAR malfunctions.

See Sections 7.3 and 9.4 for more detailed discussions of the instrumentation.

Fire protection and alarm devices are annunciated in the Control Room as described in Subsection 9.5.1.Ionization and radiation sensors detect smoke and unsafe levels of radiation in the incoming air. The system automatically switches to emergency recirculation if a radiation detector fails. Subsequent to automatic initiation of the emergency recirculation mode, the operator can regain manual control over the system from the Control Room ventilation panels. A Control Room vertical panel mounted selector switch allows the operator to select normal operation or emergency ventilation modes as required for the operation conditions.

REFERENCES

1. 10 CFR Part 50, Appendix A, General Design Criterion 19, Control Room.2. NRC Regulatory Guide 1.4, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors.

Revision 2, June 1974, United States Nuclear Regulatory Commission.

3. NRC Regulatory Guide 1.52, Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants, Revision 1, July 1976, United States Nuclear Regulatory Commission.

4. NRC Regulatory Guide 1.78, Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, June 1974, United States Nuclear Regulatory Commission.

5. Deleted.6. ANSI N509, Nuclear Power Plant Air Cleaning Units and Components.

7. ANSI N510, Field Testing of Nuclear Air Cleaning Systems.8. Leakage Characteristics of Openings for Reactor Housing Components, NAA-SR-MEMO-5 137, Atomics International, Div. of North American Aviation, Inc., June 20, 1960.9. Deleted.10. ANSI/ASHRAE 15-78, Safety Code forMechanical Refrigeration.(ii ll 1ARC AtveIY4rc 1.196%. Con 4 1?Roonm 11abifi;/7 at)' i/ /e Pewc-kr5 /gfy zooS 6.4-9 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 11 of 89 CPSES/FSAR

-

TABLE 6.4-1 NOBLE GAS AND HALOGEN INVENTORIES RELEASEDAS THE RESULT OF A MA)IMUM CREDIBLE ACCIDENT Noble Gases Curies Released From Core Curies-Nuclide (100% Core Inventory)

Kr-83m 1 x10 7 Kr-85m 3.D 2.7 x 107 Kr-85 1. 3 6.6 x 10 5 Kr-87 5.4 49 x IC0 Kr-88 7.7 7. x 10 7 Xe-131m 7.7 7.0 Xe-133m 3.L 2.9 x 1O 7 Xe-133 1. x lo, Xe-135m 4.4 4. x 10 7 Xe-135 4.& 4. x 10 7 Xe-138 f. 1. x lo, Nuclide Br-83 Br-8 Br-8 Halogens s Released From Core (50% Core Inventory) 8.5 x10 6.0x 106 1-131 574-* 1-132 7,q 1-133 1.1 1-134 t.Z I-135 /1 0 1 o 7 lo, W1, Note: r a. ,dalfof the es o en are assumed lated out with the remainerairne , dava e or re se om containmen mmediately.( re Eqvlentto Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 12 of89 CPSES/FSAR TABLE 6.4-3 LIMITATIONS OF CONTROL ROOM ENVIRONMENT Ambient pressure, in. wg Ambient temperature, FDB Ambient relative humidity, percent+0.125*75 +/-5 35 to 50 Noxious substances Not Applicable N I Maximum Radiation doses to operators:

Whole Body Gamma Dose Beta Skin Dose 5 rem 75 rem**Thyroid Inhalation Dose E5 90remml* Except during detection of smoke or toxic gas, at which time the operator may manually initiate control room isolation mode and maintain the control room at atmospheric pressure.** The 75 rem limit for beta skin dose is allowed due to the use of protective clothing and eye protection.

The dose limit is 50 rem if credit is not taken for protective clothing.I Regulatory Guide 1.195 Control Room Thyroid limit.I i Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 13 of 89 CPSES/FSAR TABLE 6.4-4 (Sheet 2)POTENTIAL LEAK PATHS AND THEIR APPROPRIATE LEAKAGE CHARACTERISTICS*

Components Unidentified Leakage Description of Components Fan margin Leak Rate (ft 3 lmin)464 800 TOTAL* The criteria used to establish leakages is based on a pressure differential of 1/4 (T)-.in.wggas specified in NRC Regulatory Guide 1.78[4].Note: The Control Room is pressurized to 1/8 in. wg in the Emergency Recirculation mode of operation.

The leakrate will be 800 ft 3/min at 1/8 in. wg. positive pressure.( Z. 7/a l in-/eQ 5 75 Can-rme to Le age-ttj7T,,h A/4?CP /Fviary62td

/fi/iz D?.Amendment 97 February I, 2001 Attachment 6 to TXX-05 127 Page 14 of 89 CPSESIFSAR h = drop fall height E = total collection efficiency for a single drop F = spray flow rate d = drop diameter modeled using a value of 10/m (this value is reduced to 1.0/m V containment spray volume after the airborne activity is reduced by a factor of 50).This model embodies assumptions equivalent to most assumptions applied to the washout of elemental iodine. The drop collection efficiency is v Monte u&ilb I= .88f5' H111. "IfIIIIIIIUIIf VALUm ULIlldkiXWU t theintir A 11 VCahL4HwII jVu~ owvar.'4 VJsuvoL[]The particulate iodine removal coefficient is listed in Table 6.5-5. _ _ _ _ _ _ _ _ _ _ _ _ _ _l. The most conservative I For each region calculated, elemental iodine re oval coefficients are above 10 hrC, respectively.

The removal coefficients us d in LOCA offsite dose calculation (see Section 15.6.5.3) have been limited to a ximum value of 10 hrf for 1 elemental iodine [1] "l tparticulate iodine removal oeIiemt0 hrf is also taken into account on the LOCA offsite dose calculation until the inventory is reduced by a factor of 50, after which a value of 1.14 hr -1 is used.4. Single Failure Analysis The containment spray chemical additive subsystem is designed to function reliably in the event of a LOCA. Redundancy requirements are met by providing two electrically separated motor-operated chemical additive tank stop valves in parallel such that, in case of single active failure, there still remains a full 100-percent capacity to reduce the elemental iodine concentration inside the Containment.

In addition, an air-operated normally open valve is provided in series with each of the motor-operated valves to ensure that the line may be closed when the chemical additive tank is empty and thus prevent air from being entrained in the suction of the containment spray pumps. There is no credit for the air-operated valve (passive, fails open) in the pH analysis.

A single failure analysis is presented in Table 6.5-4 for the chemical additive subsystem.

Reproductibility of subsystem design parameters is ensured by using performance tested chemical eductors and spray nozzles, as well as constant speed containment spray pumps. Furthermore, the only moving parts in the system are those associated with the pumps, valves, and motors. System component parameters are verified by testing.6.5-16 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 15 of 89 CPSESIFSAR TABLE 6.5-5 CONTAINMENT SPRAY SYSTEM CHARACTERISTICS Regions A B C D E Total volume, ft 3 Sprayed containment volumes, ft 3 Unsprayed containment volumes 19 Spray drops fall height, ft Number of nozzles per train (b)(272 (273 Elemental iodine removal train B)coefficient) hrWI (d )Particulate iodine removal coefficient, hrX (a) Avcrage value (b) 15.2 gpm/nozzle (c) Unobstructed spray nozzle (train A)2,309,461 1,664,901 644,560 126(a)274 2-Unit 1 train A)3-Unit 1 train B)25.92 CCO -O 4 167,645 35,040 132,605 38 67 [36](c)(64-Unit 2 train A)(67-Unit 2 train B)Q+/-(9 72,686 2,640 70,046 23 14 [5](c)(14-Unit 2 train A)(13-Unit 2 train B)_ , I X3 ( 9.4 124,910 3 3,364 0.121,546 3: 20 -27 [101(c) -(23-Unit 2 train A)(24-Unit 2 train B)Ifot=/&z~q D 7iT, -56,333.0 56,333 Total 3,031,035 1,705,945 1,325,090 382 (380-Unit 1 train A)(381-Unit I train B)(375-Unit 2 train A)(378-Unit 2 train B)GD d) tt;Kwmn X-, D/owde t be used! en Onalys;S is O Ar-'.Amendment 97 Februaty 1, 2001 Attachment 6 to TXX-05 127 Page 16 of 89 CPSESJFSAR

/ -o OUQ7-T r Se r6 IEP FRoJ,41 -T-R COMM11^AW TABLE 6.5-6 lRADIOACTIVE IODINE ISOTOPES AVAILABLE FOR RELEASEAFOLLOWING A LOCA Isotope I-131 1-132 1-133 1-134 1-135 Activity Ci) (X10 7)4.o 5 4 ZL* Based on NRC Regulatory ius tions, is the activity availablef eIaf from containment atmosphere.)

1I4/f of W ,-e ne ;, relea4S -r./ C ie ore r ihe conlainrnenf is 4gV4aIe0 b ?ir re/ea, %6 Ihe oafis Xde dtnosphere.

Amendment 97 February 1, 2001 Attachment~

6 oTX0 2 Attachment 6 to 17(X-05 127 Page 17 of 89 CPSES/FSAR TABLE 6.5-7 ESF FILTRATION UNITS EMPLOYED DURING DESIGN BASIS ACCIDENTS Design Basis Accident Filtration Units Employed Reference Section 1. Loss-of-coolant accident Control Room Emergency pressurization units 15.6.5.4(4)(b)

Emergency filtration units Primary plant exhaust unit 15.6.5.4(4)(b) 15.6.5.4(2)(f)

2. Waste gas system failure 3. Steam generator tube rupture 4. Fuel handling accident 5. Control rod ejection accident 6. CVCS letdown line break 7. Main Steam line 8. Liquid waste tank failure Control Room Emergency Pressurization Unit and Emergency Filtration Units Control Room Emergency Pressurization Unit and Emergency Filtration Units Control Room Emergency Pressurization Unit and Emergency Filtration Units Control Room Emergency Pressurization Unit and Emergency Filtration Units Control Room Emergency Pressurization Unit and Emergency Filtration Units Control Room Emergency Pressurization Unit and Emergency Filtration Units Control Room Emergency Pressurization Unit and Emergency Filtration Units 15.7.1 15.6.3 15.7.4 15.4.8 15.6.2 15.1.5 15.7.2 9-LockeJ Rotor Gtri /Zoecmn Ewer icyI tynrgevcy Filraton Unxfs 15.33 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 18 of 89 CPSES/FSAR 15.0 ACCIDENT ANALYSIS LIST OF TABLES Number Title 15.0-1 NUCLEAR STEAM SUPPLY SYSTEM POWER RATINGS 15.0-2 DELETED 15.0-3 DELETED 15.0-4 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES 15.0-5 DELETED 15.0-6 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT CONDITIONS 15.0-7 CO D GAP ACTIVI B O L OPERATION FOR 650 DAYS 15.0-8 POWER-TEMPERATURE DISTRIBUTION FOR FULL CORE 15.1-1 DELETED 15.1-2 EQUIPMENT REQUIRED FOLLOWING A RUPTURE OF A MAIN STEAM LINE 15.1-3 PARAMETERS FOR POSTULATED MAIN STEAM LINE BREAK ACCIDENT ANALYSIS 15.14 PRIMARY AND SECONDARY SIDE EQUILIBRIUM ACTIVITIES (ACCIDENT ANALYSIS)15.1-4A DARY SIDE EQUILIBRIUM ACTIVITI 15.2-1 DELETED 15.2-2 DELETED 15.3-1 Ad ,/ _15.3-2 DELETED 15.4-1 DELETED 15-xi Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 19 of 89 CPSES/FSAR List of Tables (Continued)

Number Title 15.4-2 DELETED 15.4-3 DELETED 15.4-4 PARAMETERS FOR POSTULATED ROD EJECTION ACCIDENT ANALYSIS 15.5-1 DELETED l 15.6-1 DELETED 15.6-2 PARAMETERS FOR POSTULATED STEAM GENERATOR TUBE RUPTURE ACCIDENT ANALYSIS 15.6-3 PREACCIDENT IODINE SPIKE CONCENTRATION IN THE PRIMARY COOLANT 15.6-4 IODINE APPEARANCE RATE TO THE REACTOR COOLANT AFTER THE ACCIDENT 15.6-5 LOCA ANALYSIS INPUT PARAMETERS 15.6-6 DELETED 15.6-7 DELETED 15.6-8 INVENTORY AND DESIGN BASES GAP ACTITESIES3 15.6-9 PARAMETERS FOR POSTULATED LOCA ANALYSIS 15.6-10 ACTIVITY AVAILABLE FOR RELEASE VIA ESF COMPONENTS AT T=O FOLLOWING LOCA 15.6-11 This table has been revised and renumbered 15.6-10 15.6-12 ATMOSPHERIC DILUTION FACTORS IN CONTROL ROOM DOSE ANALYSIS 15.7-1 PARAMETERS FOR POSTULATED WASTE GAS DECAY TANK RUPTURE ACCIDENT 15.7-2 GAS DECAY TANK INVENTORY FOR ACCIDENT ANALYSIS (ONE GAS DECAY TANK)15.7-3 MAXIMUM RADIOACTIVITY IN A FLOOR DRAIN TANK FOR ACCIDENT ANALYSIS I15-xii August 1, 2002 Attachment 6 to TXX-05 127 Page 20 of 89 CPSES/FSAR List of Tables (Continued)

Number Title 15.74 PARAMETERS FOR POSTULATED RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE 15.7-5 DELETED 15.7-6 Noble Gas and Iodine Activities Released to the Environment As a Result of a Fuel Handling Accident 15.7-7 PARAMETERS FOR POSTULATED FUEL HANDLING ACCIDENT ANALYSIS 15B-1 PHYSICAL DATA FO DOSE CONVERSIONS 15-xiii August 1, 2002 Attachment 6 to TXX-05127 Page 21 of 89 CPSES/FSAR het o Z ( COE ADGAP Anq"lTiES TABLE 15.0-7>CORE AND GAP ACTIVITIES BASED ON FULL POWER OPERATION FOR 650 DAYS@)Core Activity Percent Gap Activity Isotope (Curies) Activity in Gal- (Curies)1-131 8.80 x IC 0.960 8.45 x 10'I-132 1.34 x 108 0.105 1.41 x 105 1-133 1.97 x 10' 0.316 6.24 x 10 5 1-134 2.31 x 108 0.0652 1.50 x Io, 1-135 1.79 x 108 0.180 3.22 x 10'Xe-131m 6.68 x 105 1.17 7.81 x 103 Xe-133 2.03 x 108 0.778 1.58 x 106 Xe-133m 5.16 x 106 0.510 2.63 x 104 Xe-135 5.55 x 107 0.210 1.17 x 105 Xe-135m 5.46 x 107 0.0355 1.94 x 104 Xe-138 1.79 x 108 0.0370 6.63 x 104 Kr-83m 1.64 x 107 0.0964 1.58 x 10 4 Kr-85 9.99 x 105 18.3 1.83 x 10'Kr-85m 3.95 x I0 7 0.145 5.72 x 10 4 Kr-87 7.59 x 107 0.0783 5.94 x 10 4 Kr-88 1.08 x 10 8 0.116 1.25 x 10 5 Kr-89 1.40 x 10 8 0.0161 2.26 x 10 4-rLis tndle i5 Vse- to estma,-re acbv;-,es -he ac vitse V'Cd 11 -the b;e*¶6mnS bn!;;j C1c~i/,0l7a qare ihown In SI eetZ.Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 22 of 89-A G --AND DESIGN BASES GAP ACTIVITIE49r(6)

Cures in Core (x07 ?CC)Curies in Gap (x10 7)I W524 A Iemsr A B-Isotope in Gap Kr-83m Kr-85 Kr-85m Kr-87 Kr-88 1.132 O. 0o'7,2& 0.666.oQ7 -e-?OG 5.39 49 7.7 8.7-1.6 5 10 -1 10 0. O09*2-o.0,07244.4199-o.i4- 5 4.,;-0. Z( I5-o.45~0.83 10 ,0 I o 1 0 I 0 OA3Z2 O ,007L;&o.Zq7 o .I$31 o .77o Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-138 1-131 1-132 1-133 1-134 1-135 0o 077 3.20 4,4oD 4-,bO 4.40 10O 5-+0-5 46-5 IfS-4l 5 C) ,oa3E-r0.AOt&-

i.. 9 t.9-O.ib 0449 OZ3 -&4-o. ZZ -f40 0.90v 1.69 1 0 IO l0 1 0 l0 10 IC0 10 l O l0 I 0 o.ofol7 Z.18 O .3Z o. 44 I ,68 ,.sa azo Z.4Z*.a-7 z2-o Z4.7 7Z1+ 7 10O 8-to- 5 t- 5 C 44-/.21 20.09 J-Zi 2.1.o035rs 6 ,a! Noble Gas and Iodine inventories used for design basis accident analysis.C X.IA EICrt iA/.-5CR T Three region equilibrium cycle core at end of life. The three regions have operated at a specific power of 40.03 MW/MTU for 300, 600, and 900 EFPD, respectively.

• Perce,7-t o -P /cT ity Xc rap i tr ladJ tjectrao A cc.vierft 8 Co rms Yn Gap (K i673 Amendment 97 February 1. 2001 Attachment 6 to TXX-05127 Page 23 of 89 CPSESIFSAR 15.1.5.3 Environmental Consequences The postulated accidents involving release of steam from the secondary system do not result in a release of radioactivity unless there is leakage from the RCS to the secondary system in the steam generators.

A conservative analysis of the potential offsite doses resulting from this accident is presented considering equilibrium operation based upon Technical Specification limits of primary coolant activity concentration and primary to secondary leakage.The radiological consequences analysis for a postulated rupture of a main steam line assumes a leak rate to the faulted steam generator during the accident that results in calculated consequences approaching a small fraction (10%) of the 1OCFRIOO guideline values for the accident-initiated iodine spike. This leak rate provides a maximum primary-to-secondary leak rate limit against which the predicted end-of-cycle leakage is The following assumptions and parameters are used to calculate the activity releases and offsite doses for a postulated main steam line break.1. Offsite power is lost; thus, the main steam condenser is not available for steam dump.2. Air ejector release and steam generator blowdown do not occur durin t accident.

ro.i he inthc S6s;5 e,,-Ckte t 8 ou 3. [th ou Tteam and activity release to the environment erminate ue to primary and secondary side pressure equalization.

_, 4'The elemental iodine partition factor in the affected steam generator throughout the accident is 1.0.The elemental iodine partition factor in the unaffected steam generators throughout the accident is 0.01.All releases are assumed to be at ground level.Two separate cases are analyzed: a. An iodine spike occurs coincident with the steam line break. The iodine 1.0 appearance rates from the fuel are increased to 500 times the equilibrium 9./ J appearance rates associated with a primary coolant concentration of luCi/gm Dose Equivalent 1-131. The spike appearance rates are given in_ Taliee 15.1-3.the rek~a of QCttiYt,/

fid leaks3 nto, f CvjIees Con7es)£'At;I f*em RCS is coo/d21 Ibe/oa~ ZIZ ley-ees F (zo hc;ros).15.1-15 Amendment 98 Attachment 6 to TXX-05 127 Page 24 of 89 CPSES/FSAR

b. A transient has occurred prior to the steam line break. The transient causes an iodine spike that raises the primary coolant activity to 60 uCi/gm Dose Equivalent I- 131. The preaccident spike concentrations are given in Table 15.1-3.8. Other assumptions are in T 1-3. ( be /.z rein qua £.OOZ Based on the foregoing m o thyroid des and whole body doses for the pre-accident iodine spike caswere calcult t respectively, at the EAB an rem an em, respectively, at the LPZ.These doses are below the applicable limit values of 300 rem thyroid and 25 rem whole body set forth in IOCFR 100. The thyroid doses and whole body doses for the accident-initiated iodine spike case Were calculate to 6.9 rem, respectively, at the EAB andrm and.1Q0 rem, respectively, at the LPZ.These doses are below a "small frtion" of th OCFRO00 values; or 30 rem thyroid and 2.5 rem whole body.15.1.5.4 Conclusions The analysis has shown that the criteria stated earlier in Section 15.1.5.1 are satisfied.

Although DNB and possible clad perforation following a steam line rupture are not necessarily unacceptable and not precluded by the criteria, the above analysis shows that no DNB or clad perforation occurs for any main steam line rupture assuming the most reactive RCCA stuck in its fully withdrawn position.

The radiological consequences are within the guideline values of I OCFRI 00 for the pre-accident iodine spike case and within a "small fraction" (10%) of the IOCFRI 00 guideline values for the accident-initiated iodine spike case.REFERENCES

1. "Westinghouse Anticipated Transients Without Trip Anaylsis," WCAP-8330, August 1974.2. International Commission on Radiological Protection, "Limits for Intakes of Radionuclides by Workers," ICRP Publication 30, Volume 3 No. 1-4, 1979.3. Not Used.15.1-16 Amendment 98 Attachment 6 to TXX-05 127 Attachment 6 to TXX-05127 Page 25 of 89 CPSESIFSAR TABLE 15.1-3 (Sheet 1 of 3)PARAMETERS FOR POSTULATED MAIN STEAM LINE BREAK ACCIDENT ANALYSIS I1. Data and assumptions used to estimate radioactive source from postulated accidents a. Power level (MWt) JJ )b. Primary coolant activity concentration Ci/gm Dose prior to accident Equivalent 1-131 c. Secondary side coolant activity 0.1 uCi/gm Dose concentration prior to accident Equivalent 1-131 d. Primary to Secondary leakage in affected steam generator 9gP1 )I e. Primary to Secondary leakage in 4.unaffected steam generato rs f Offsite power lost lost g. Failed fuel none 2. Data and assumptions used to estimate activity released a. Iodine partition factor for initial and 1.0 long term steam release from affected steam generator b. Iodine partition factor in unaffected 0.01 steam generators prior to and during accident c. Initial steam release fr affected steam 17 generaj 0 t) ( l Note: The e condary sid id inventory in the affected steam generator is assumed to be released.Amendment 100 Attachment 6 to TXX-05 127 Page 26 of 89 CPSES/FSAR TABLE 15.1-3 (Sheet 2)PARAMETERS FOR POSTULATED MAIN STEAM LINE BREAK ACCIDENT@9 ANALYSIS d. ng terms release from affected 0 )steam generato rator}e. steam release from tunaffected steam generators) (0 to 2 hr) 438,000 (2 to 8 hr) 7 3. Dispersion data 'I I a. EAB and LPZ distances 2080m and 4 miles b. x/Q 4. Dose data@ EAB (0 -2 hr)1.6 x IO 4 sec/m 3@ LPZ (0 -8 hr)0i -2x 105 sec/me 3 ( A0 u 4 i )a. Method of dose calculations

b. Dose conversion assumptions

c. Primary and secondary side equilibrium activities

d. Iodine spike appearance rates e. Pre-accident spike iodine concentrations See Appendix 15B See Appendix 15B See Table 15.1@(t)1-131: 103.2 Cumin @I-132: 304.3 Ci/min 730C I-133: 185. Ci/min 4z8 I-134: 128. Ci/min 3 18 1-135: 140. Ci/min 3Z26 I-131: 6.7 uCi/gn 46, 1-132: 6.71 uCi/gm Go-S 1-133: 0.2 Ci/gm 7ZO I-134: 9.53 Ci/gm 9. c 1-135: 8.3 uCi/gm 3 4 Amendment 100

-.Attachment 6 to TXX-05127 Page 27 of 89 CPSES/FSAR TABLE 15.1-3 (Sheet 3)PARAMETERS FOR POSTULATED MAIN STEAM LINE BREAK ACCIDENT ANALYSIS f. Doses (rem)thyroid whole body (gamma dose)@EAB (0-2 hr)concurrent iodine spike preaccident iodine spike@LPZ (0-30 days)concurrent iodine spike preaccident iodine spike.9 l1.4 l 44-01.2 E+1S 2.5 25 0.61l&e.04 001 Amendment 98

-Attachment 6 to TXX-05127 Page 28 of 89 CPSES/FSAR TABLE 15.1-4 PRIMARY AND SECONDARY SIDE EQUILIBRIUM ACTIVITIES (ACCIDENT ANALYSIS)Primary Coolant Concentration (11i/gm), Secondary Liquid Concentrrtion (11ri/gm)seeendal)shc 111 e9VilfClftti Nnluide Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-138-.E1 , *5.06e-1l 0.00 x 10-0 40-A-i8 1*2.20e+O 0.00 x 10-0 1.A 8.03e+00 0° x 10-0 3 1.43e+00 0.00 x 10-0.4 103 +00 0.00 x 10-0^.G4.0o 1 10^ ^10.42e+00 0.00 x 10-°648-0-i 20 .87e+01 0.00 x 10-°11A .28e-O1 0.00 x 10.0 0.00 x 10 .0 1.73 10 °[7.92e+004 0.00 x 10-0 1.JA 1 s 7.0e-0 l0.00 x 1 0°, 9, , iO eq 1.04 ^10-2ti `14 n n 6^4 u+ei X .v+. _,.t..,u .I-131 6.72 x 10' 7.8e-01 1 4-A 8.276e-02

l. 10 I-132 6.72 x 10-1 7.8e-O1 l5.J3 1+ 85 2.853e-021

-A I-133 1.01 x 10+0 1.2e+00 1 '4e-,!-i.9.05e-02 10 1-134 1.37 x 101 1.6e-01 l 1490T 10 1-135 5.52 x 10' 6.4e-O1 e199- 1 e3-371e-o2 l 109r+a. Based on a primary coolant specific activity of 1.0 uCi/gm Dose Equivalent 1-131.NVtt-. +he ssflI -Vvilt 31XS fi t

• b= %9.1 2~""bgkU (Md.C t, 0.45 C.'/&pn D3.,b or We gases are ised on peration with% fuel defects.,~1 a DEH1 vatu%.. ufG.4 eii'Li. Ti.., a.Uip.tiUu

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£1XINA, U iL. VNI.I.IUE L1 .LM tAA.%.P1L9LUi UV L1I, 1i9aiil OLt.ak L..iA1.. ".Sa1 A-a"a1p1Y31 a La 111 (J AI"J all iI...puiuv St., aM 'Laau LAI. a lJL -j I l IItl val.ut. ul 1.Ut ueii grrT.b. Based on a secondary side specific activity of 0.10 uCi/gm Dose Equivalent 1-131.Amendment 98 Attachment 6 to TXX-05127 Page 29 of 89 CPSES/FSAR TADL:E f6.i 4A IPffihfill;kY i( SLAML AJD I QUALyiRflM ETJ)TE)(S T-EM+vnJ f D ERE-EA~i i2*L'b-Y Sf)., 8q if b8191s 3ie- 1 i9~Hef w-1i 9f,- 3 V 1 .VVSallL n .,.VA^W9sXtaLSV

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• ae J.J I s A AU Steac-mal.-lentaif 0.00 A 1 4 9.n_. " lw, V.VVt .tV i Hi I H-i+-13;a. BaPed Oll aItWlant spifie itti v i;ty of O.45 uCriL'l Dose Eii ailen I 131.U. BtD Ol v a Sid.ta.,y sJpd r ufie ti v ity vro.1o uciabelsi DBOS. E-,XivLh I 131.% p. eul1ufl..

ua~illb thy(viv ua1 COLITCY ;Ui ffa.tul. fiulii ftr.,.C ,. 2 D Amendment 98

-Attachment 6 to TXX-05 127 Page 30 of 89 CPSES/FSAR Zirconium-Steam Reaction The zirconium-steam reaction can become significant above 1 800TF (clad temperature).

The Baker-Just parabolic rate equation shown below is used to define the rate of the zirconium-steam reaction.d(w 2) X0 Mt= 33.3 x10 6 exp(- 45,500/1.986T) where w amount reacted (mg/cm )t = time (sec)T = temperature (fF)The heat of reaction is 1510 cal/gm.The effect of the zirconium-steam reaction is included in the calculation of the "hot spot" clad temperature transient.

Plant systems and equipment which are available to mitigate the effects of the accident are discussed in Section 15.0.8 and listed in Table 15.0-6. No single active failure in any of these systems or equipment will adversely affect the consequences of the accident.Evaluation of the Radiological Consequences An evaluation of the number of rods in DNB was performed in order to assess the radiological consequences of this event. Analysis assumptions were made to maximize the heat flux and thus minimize the DNB, consistent with the current FSAR. The number of rods experiencing DNB is determined from a pin census showing the relative peaking factors of each rod.INSERT LOCKED ROTOR 15.3.3.3 Conclusions As noted in Appendices 4A and 4B for Units 1 and 2, respectively, the following conclusions are made: I. Because the peak RCS pressure reached during any of the transients is less than that which would cause stresses to exceed the faulted condition stress limits, the integrity of the primary coolant system is not endangered.

2. Because the peak clad surface temperature calculated for the hot spot during the worst transient remains considerably less than 2700T the core will remain in place and intact with no loss of core cooling capability.

15.3-7 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 31 of 89 shown in sheet 2 of Table 15.0-7 lLOCKED ROTOR The doses resulting from a locked rotor acciden/are based on a conservative fission product release to the reactor coolant of the gap activity'from 10% of the fuel rods in the core. The method follows Regulatory Guide 1.195. Prior to the accident , it is assumed that the plant has been operating with simultaneous fuel defects and steam generator tube leakage for a sufficient amount of time to establish equilibrium levels of activity in the primary and secondary coolants equal to the Table 15.1-4 values.Following the postulated locked rotor accident, the activity released from the pellet-gap due to the fuel failure is assumed to be instantaneously released to the primary coolant and is assumed to be immediately available for release from the RCS.The following assumptions are used in the analysis: 1. The initial Reactor Coolant System (RCS) activity is 1.0 ptCilgm DE 1-131.2. The initial RCS noble gas activity is based on 1% fuel defects.3. The initial secondary activity is 0.1 ItCi/gm DE 1-131 .4. A radial peaking factor (FAh) of 1.65 is applied in the calculation of gap activities released (per Regulatory Guide 1. 195).5. 10% of rods are assumed to undergo clad damage sufficient to release all of their gap activity.6. The primary to secondary leak rate is 1.0 gpm plant total.A summary of the parameters used in the analysis is given in Table 15.3-1.Based on the foregoing model, the thyroid and whole body doses are conservatively calculated to be 1.5 rem and 0.12 rem, respectively, for the exclusion area boundary (EAB), and 2.5 rem and 0.037 rem respectively, for the low population zone (LPZ). These doses are well below the values of 30 rem thyroid and 2.5 rem whole body set forth in Regulatory Guide 1 .195.

Attachment 6 to TXX-05127 Page 32 of 89 CPSESIFSAR 97 TL.. 1 1 1 1 L.1 TL .JI CAFCiVIvi1.

D kiD 1 1 1Xl ka uhil t 1 1 1 1 Lb 1 Ofl 3 ud aa3",Iii.LH i.A fji.,ii.DD fbL till. RodE~tiJ E V %QI~ ak l f 5.4.A5, t1C1f1, .ht *_U1~uo01 Of flt laiuug

• a-lt a- .a f~ t .- Mr n 121dieloLbieud JiJ. beeIAidsV tile dose Val sUII set fiort n 10-R1ie0) applJ ton th:%, V %.lit.15.3.4 REACTOR COOLANT PUMP SHAFT BREAK 15.3.4.1 Identification of Causes and Accident Description The accident is postulated as an instantaneous failure of a reactor coolant pump shaft, such as discussed in Section 5.4. Flow through the affected reactor coolant loop is rapidly reduced, though the initial rate of reduction of coolant flow is greater for the reactor coolant pump rotor seizure event. With a failed shaft, the impeller could conceivably be free to spin in the reverse direction instead of being fixed in position.

The effect of such reverse spinning is a slight decrease in the end point (steady state) core flow.The analysis presented in Section 15.3.3 represents the limiting condition, assuming a locked rotor for forward flow but a free-spinning shaft for reverse flow in the affected loop.15.3.4.2 Conclusions The conclusions of Section 15.3.3.3 apply for the Reactor Coolant Pump Shaft Break event.REFERENCES I. Not Used.2. Baldwin, M. S., Merrian, M. M., Schenkel, H. S. and Van De Walle, D. J., "An Evaluation of Loss of Flow Accidents Caused by Power System Frequency Transients in Westinghouse PWRs," WCAP-8424, Revision 1, June 1975.-Insert Table 15.3-1 after this page 15.3-8 Amendment 97 February 1, 2001

--Attachment 6 to TXX-05127 Page 33 of 89 TABLE 15.3-1 PARAMETERS FOR POSTULATED LOCKED ROTOR ACCIDENT ANALYSIS 1. Data and assumptions used to estimate radioactive source from postulated accidents a. Power level 3565 b. Primary coolant activity concentration 1.0 pCi/gm Dose prior to accident Equivalent 1-131 c. Secondary side coolant activity concentration 0.1 pCi/gm Dose prior to accident Equivalent 1-131 d. Primary to secondary leakage I gpm plant total e. Offsite power lost f. Failed fuel 10%g. Radial Peaking Factor 1.65 2. Data assumptions used to estimate activity released a. Iodine partition factor for steam release from 0.01 secondary side b. Steam release from secondary side 450,000 (0 -2 hours)1,002,000 (2- 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />)3. Dispersion data a. EAB and LPZ distances b X'Q 4. Dose Data a. Method of dose calculations

b. Dose conversion assumptions

c. Doses@EAB (0 -2 hours)@LPZ (0 -8 hours)2080 m and 4 miles@ EAB (0 -2 hours)1.6 x 104 sece/r 3@ LPZ ) (0 -8 hours)2.4 x 10-5 sec/m3 thyroid 1.5 rem 2.5 rem See Appendix 15B See Appendix 15B whole body (gamma dose)0.12 rem 0.037 rem Attchmnt t TX-0512 Attachment 6 to TXX-05127 Page 34 of 89 CPSES/FSAR 15.4.8.3 Environmental Consequences The ejection of an RCCA constitutes a break in the RCS, located in the reactor pressure vessel head. The effects and consequences of loss of coolant accidents are discussed in Section 15.6.5. Following the RCCA ejection, the operator would follow the same emergency instructions as for any other loss of coolant accident to recover from the event.Fission Product Release It is assumed that fission products are released from the gaps of all rods entering DNB. In all cases considered, less than 10 percent of the rods entered DNB.Pressure Surge A detailed calculation of the pressure surge for an ejection worth of one dollar at beginning-of-life, hot full power, indicates that the peak pressure does not exceed that which would cause stress to exceed the faulted condition stress limits [2]. Since the severity of the present analysis does not exceed the "worst case" analysis, the accident for this plant will not result in an excessive pressure rise or further damage to the RCS.Lattice Deformations A large temperature gradient will exist in the region of the hot spot. Since the fuel rods are free to move in the vertical direction, differential expansion between separate rods cannot produce distortion.

However, the temperature gradients across individual rods may produce a differential expansion tending to bow the midpoint of the rods toward the hotter side of the rod. Calculations have indicated that this bowing would result in a negative reactivity effect at the hot spot since the core is under-moderated, and bowing will tend to increase the under-moderation at the hot spot. In practice, no significant bowing is anticipated, since the structural rigidity of the core is more than sufficient to withstand the forces produced.

Boiling in the hot spot region would produce a net flow away from that region. However, the heat from the fuel is released to the water relatively slowly, and it is considered inconceivable that cross flow will be sufficient to produce significant lattice forces. Even if massive and rapid boiling, sufficient to distort the lattice, is hypothetically postulated, the large void fraction in the hot spot region would produce a reduction in the total core moderator to fuel ratio, and a large reduction in this ratio at the hot spot. The net effect would therefore be a negative feedback.

It can be concluded that no conceivable mechanism exists for a net positive feedback resulting from lattice deformation.

In fact, a small negative feedback may result. The effect is conservatively ignored in the analysis.

--n 7 Radiological Consequencesunes2 The doses resulting from a rod ej etion accident are basid on a conservative fission product release to the reactor co lant of the gap activity from 1O percent of the fuel rods in the core plus the core activi trom the assumed 0.25 percent core melt. The method of 15.4-28 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 35 of 89 CPSESJFSAR l ofs1itiAonfl3 1 1 .195.lanalysis complies with the requirements of of Regulatory Guide +-1. eeet"a Ilut'd fll £AP.I.dilA fA(B'5.Following a postulated rod ejection accident, two activity release paths contribute to the total radiological consequences of the accident.

The first release path is via Containment leakage resulting from release of activity from the primary coolant to the Containment.

The second path is the contribution of contaminated steam in the secondary system dumped through the relief valves, since offsite power is assumed to be lost.Prior to the accident, it is assumed that the plant has been operating with simultaneous fuel defects and steam generator tube leakage for a sufficient period of time to establish equilibrium levels of activity in the primary and secondary coolants equal to the Table 15.1-4 i3 Following a postulated rod ejection accident, the activity released from the pellet-clad gap due to failure of a portion of the fuel rods and the melted fuel is assumed to be instantaneously released to the primary coolant. The activity released to the primary coolant is assumed to be uniformly mixed throughout the coolant instantaneously.

Thus, the total activity of the primary coolant is assumed to be immediately available for release from the RCS. Of the activity released to the Containment from the coolant through the rupture in the reactor vessel head, 100 percent is assumed to be mixed instantaneously throughout the Containment.

Fifty (50) percent of the iodine activity released from the melted fuel is assumed to immediately plate out on Containment surfaces.

The remaining activity is available for leakage from the Containment at the design leak rate of 0.10 percent of Containment volume per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at a rate of 0.05 percent of Containment volume per day for the duration of the accident.

The only removal processes considered in the Containment for the activity remaining after the above plate out are radioactive decay and leakage from the Containment.

The model for the activity available for release to the atmosphere from the relief valves assumes that the release consists of the activity in the secondary coolant prior to the accident plus that fraction of the activity leaking from the primary coolant through the steam generator tubes following the accident.

The leakage of primary coolant to the secondary side of the steam generator is assumed to continue at its initial rate, which is assumed to be the same rate as the leakage prior to the accident, until the pressures in the primary system and the secondary system are equalized.

No mass transfer from the primary system to the secondary system through steam generator tube leakage is assumed thereafter.

With the assumption of coincident loss of offsite power, activity is assumed to be released to the atmosphere from a steam dump through the relief valves fiv 400"VsId. r8 hours.A summary of parameters used in the analysis is given in Table 15.4-4.Fuel melting, limited to less than the innermost 10 percent of the fuel pellet at the hotspot, is included in the design criteria to ensure that fuel dispersion into the coolant does not occur [1]. Even though centerline melting in a small fraction of the core is not 15.4-29 Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 36 of 89 CPSESFSAR he affected fuel rods are assumed to be operating I at 1.65 times the core average and thus have 1.65 times the average fisson product inventory.

expected, a conservative upper limit of fission prod release from the core as a result of a rod ejection accident can be estimated.

11.1951/The Regulatory Guide l 7,7 limit of fission pro ct release from the core for this very conservative case is determined using the fol wing assumptions:

1. 100 percent of the noble gases an odines in the clad gaps of those fuel rods experiencing clad damage (assu ed to be 10 percent of the rods in the core) are released to the reactor coolant. e gap activities are presented in Table 15.6-8.2. 25 percent of the iodines and 100 percent of the noble gases in the fuel that melts are assumed to be released to the reactor coolant and available for release via the containment building.

50 percent of the iodines and 1 00 percent of the noble gases in the fuel that melts are assumed to be available for release from the secondary systems via primary-to-secondary leakage.3. The fraction of fuel melting is conservatively assumed to be 0.25 percent of the core, as determined by the following method [I]: a. A conservative upper limit of 50 percent of the rods experiencing clad damage also may experience centerline melting (a total of 5 percent of the rods in the core).b. Of rods experiencing centerline melting, only a conservative maximum of the innermost 1O percent of the volume actually melts (equivalent to 0.5 percent of the core that could experience melting).c. A conservative maximum of 50 percent of the axial length of the rod experiences melting due to the power distribution (0.5 of the 0.5 percent of the core = 0.25 percent of the core).4. Prior to the accident, the plant is assumed to be operating at full power with coincident fuel defects and steam generator tube leakage. The steam generator tube leak rate is assumed to be 1.0 gpm. The initial primary coolant and secondary liquid and -a.te activity concentrations are given in Table 15.1-4.5. Instantaneous mixing in the Containment of all activity released from the coolant is assumed.6. Fifty (50) percent of the iodine activity released from the melted fuel to the Containment atmosphere immediately plates out on Containment surfaces.lhours,l7. Primary and secondary system pressures are equalized after A V thus terminating primary to secondary leakage in the steam generators.

The core activities are presented in Sheet 2 of Table 15.0-7.15.4-30 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 37 of 89 CPSES/FSAR 11,452,000 l 8. For the case of loss of offsite power, 3,2608 lb of steam are discharged from the secondary system through the relief valves, during the East following the accident.

Steam dump is terminated after 400 wsv.dsN% .8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />si 9. All releases to the atmosphere are assumed to be at ground level.10. Other assumptions are detailed in Table 15.4-4.lcontainment leakagelBased on the foregoing model, the thyroid and whole body doses from the for this accident are conservatively calculated to be 12.4 s 4.2 % 10 L 14.0 rem and 0.059 rem, respectively, for the exclusion area boundary (EAB), and ..20.0 rem and 0.031 rer respectively, for the low population zone (LPZ). The thyroid and whole body doses fr releases the secondary sidelfor this accident are conservatively calculated to be 3 .eifre¶¶l 2.3 rem and 10.29 9 10 i rem, respectively, for the EAB, and 4.4 .l.. anl 4. 9 .T0 " , 3.8 rem and o.o01rem, respectively, for the LPZ. The doses at these distances are below the dose values o rem whole body and 300 rem thyroid set forth in 1OCFR100.15.4.8.4 Conclusions As noted in Appendices 4A and 4B for Units 1 and 2, respectively, even on a pessimistic basis, the analyses indicate that the described fuel and clad limits are not exceeded.

It is concluded that there is no danger of sudden fuel dispersal into the coolant. Since the peak pressure does not exceed that which would cause stresses to exceed the faulted condition stress limits, it is concluded that there is no danger of further consequential damage to the RCS. The analyses have demonstrated that upper limit in fission product release as a result of a number of fuel rods entering DNB amounts to 10 percent.REFERENCES I. "Westinghouse Anticipated Transients Without Trip Analysis," WCAP-8330, August 1974.2. Risher, D. H., Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1-A, January 1975.3. Bishop, A. A., Sandburg, R. 0. and Tong, L. S., "Forced Convection Heat Transfer at High Pressure After the Critical Heat Flux," ASME 65-HT-3 1, August 1965.15.4-31 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 38 of 89 CPSES/FSAR TABLE 15.4-4 (Sheet 1 of 4)PARAMETERS FOR POSTULATED ROD EJECTION ACCIDENT ANALYSIS I. Data and assumptions used to estimate radioactive source from postulated accidents a. Power level (MWt)b. Primary Side equilibrium activity concentration

c. Secondary Side equilibrium activity concentration

d. Steam generator tube leak rate prior to and during steam dump (gpm)e. Failed fuel[J lRadial peaking factor gm f- Activity released to reactor coolant from failed fuel and available for release I(i.e., gap fraction)Noble gases Iodines 3565 IluCi/gomn Dose _Equivalent 1-131 Ii 0.1 uCi/gm Dose Equivalent 1-131 1 10 percent of fuel rods in core-Iof core inventory I N 10 percent 47 .111%,11tM7 10 percent ..apF .ive-elyi Mh g. Melted fuel In h- Activity released to reactor coolant from melted fuel 0.25 percent of core Noble gases inventory 0.25 percent of core Iodines 0.0625 percent of core inventory (available for release via the containment building Jta 1 pVL%,.S..S

'tLU.a., fi mit V .tbee*4 A IV119e f~educe E i * *. Ie ! r0 1 tilt'. iL1E A ..Y .- L.t-'.. F U IU L L LA L -1.J I V i V U VJ A MI_~ l

• I t * %-,ILL%,&

• - ._

• Am-e A tn dm t 98 .AOrc benfit SV44II 1 A11 AILAA --A1^"11J3*IU UVL 19*Is~kIk~~II AL1 I , A ,, , ,

• hili *.*f*time of i ae CAL .-A A- "-A LAr~ *% _ wt 1 **aanssbaa isJSvaus s .J bIt~ss s~ aaau~~usus stAmendment 98s Attachment 6 to TXX-05127 Page 39 of 89 CPSES/FSAR TABLE 15.4-4 (Sheet 2)PARAMETERS FOR POSTULATED ROD EJECTION ACCIDENT ANALYSIS 0.125 percent of core inventory available for release from the secondary system primary to-seocndary leakage)Mli?Iodine fractions (organic, elemental, and particulate)

Regulatory Guide 4 14 2. Data and assumptions used to estimate activity released a. Plate out of iodine activity 50 percent released to Containment (applied only to fuel melt releases'b. Containment leak rate c. Iodine partition factor in steam generators prior to and during accident d. Offsite power 0.1 percent of containment 0 ~ volume per day (0t24 hr)0 0.05 percent of containment 0.01 volume per day (24 <t<720hr)lost I e. Steam dump from relief valves (lb)f. Duration of dump from relief valves (ael, i,° 1 450,000 (0 -2 Hours)I 11,002,000 (2 -8 Hours) j 46e g. Time between accident and equalization of primary and secondary system pressures (ahour h. Steam dump to condenser (lb)3. Dispersion data a. EAB and LPZ distances 0.0 2080m and 4 miles Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 40 of 89 CPSES/FSAR TABLE 15.4-4 (Sheet 3)PARAMETERS FOR POSTULATED ROD EJECTION ACCIDENT ANALYSIS b. X/Q 1.6 x 10' sec/m3@EAB (0-2 hr)(Q0 sec/m 3@LPZ (0-8 hr)4. Dose data a Method of dose calculation

b. Dose conversion assumptions

c. Doses See Appendix 15B See Appendix I SB@EAB (0-2 hr)thyroid dose-24R Containment leakage Amendment 97 February 1, 2001

--Attachment 6 to TXX-05 127 Page 41 of 89 CPSES/FSAR TABLE 15.4-4 (Sheet 4)PARAMETERS FOR POSTULATED ROD EJECTION ACCIDENT ANALYSIS Containment leakage Primary to secondary leakage Containment leakage Primary to secondary leakage*pt~ vval fvs buv y dvac>thyroid dose 20.0 = 649 rem 3.8 I=4 0A rem-8.49-renm Whole body dose (gamma dose)0 I rem 0.091 = rem Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 42 of 89 CPSESIFSAR 15.6.1.3 Conclusions As noted in Appendices 4A and 4B for Units 1 and 2, respectively, the results of the analysis show that the pressurizer low pressure and the Overtemperature N-16. Reactor Protection System signals provide adequate protection against the RCS depressurization event.15.6.2 BREAK IN INSTRUMENT LINE OR OTHER LINES FROM REACTOR COOLANT PRESSURE BOUNDARY THAT PENETRATE CONTAINMENT There are no instrument lines connected to the RCS that penetrate the Containment.

However, the 3 inch CVCS letdown line and grab sample lines from the hot legs of reactor coolant loops I and 4, and from the steam and liquid space of the pressurizer, do penetrate the Containment.

The grab sample lines are provided with normally closed isolation valves on both sides of the Containment wall and are designed in accordance with the requirements of General Design Criterion 55.The most severe pipe rupture with regard to radioactivity release during normal plant operation occurs in the Chemical and Volume Control System (CVCS). This Condition III event would be a complete severance of the 3 inch letdown line just outside containment, between the outboard letdown isolation valve and letdown heat exchanger (see Figure 9.3-10), at rated power condition.

The occurrence of a complete severance of the letdown line would result in a loss-of-reactor coolant at the rate of about 190 gpm, which is within the makeup capacity of any two of the three charging pumps (see Table 9.3-7, sheet 1). Two pressure switches located in the CVCS letdown line provide the means for detecting a rupture of the line. A low pressure signal, initiated by the rupture, will activate an alarm on the main control board, alerting the operator of the rupture, causing him to manually isolate the system. No credit is taken for Engineered Safety Features System actuation.

Frequent operation of the automatic Reactor Makeup System will also provide the operator some indication of the loss-of-reactor coolant. No transient analysis is performed for this accident.The time required for the operator to identify the accident and initiate the closure of the It is letdown isolation valve is expected to be within 30 minutes after accident initiation conservatively including 10 seconds for the letdown isolation valve closure time. Reactor coolant is assumed that assumed to be released until the isolation valve is fully closed. Ceiklai(tiuX.

..d;ila 1 tlat 20.1 percent of the leaking coolant flashes to steam. All of the iodine in this steam is assumed to become airborne and is available for release to the atmosphere.

All noble gases contained in the leaking primary coolant are available for release to the atmosphere.

A ground level release was postulated with an atmospheric dilution factor of 1.6 x 104 sec/M 3 at the minimum EAB distance.

The equilibrium concentration of radioactive nuclides in the reactor coolant is given in Table 15.14. Effects of a concurrent iodine spike are included in the analysis.15.6-3 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 43 of 89 CPSESJFSAR 0 .039l Based on the foregoing model e doses at the EAB are conservatively calculated to be 3r rem to the thyroid and rem to the whole body. The LPZ doses are Io8 conservatively calculated to be -55 rem to the thyroid and B.i-i 4emto the whole body. As expected, the radiological onsequences resulting from the failure of the 3 inch CVCS letdown line do not exceed a sall fraction of the dose values set forth in IOCFR100.15.6.3 STEAM GENERATOR TUBE FAILURE 15.6.3.1 Identification of Causes and Accident Description The accident examined is the complete severance of a single steam generator tube. This event is considered an ANS Condition IV event, a limiting fault (see Section 15.0.1) and is analyzed to demonstrate that the resulting radiological doses are within the guideline values of I OCFR100. Timely operator response is required to terminate the primary-to-secondary break flow and to ensure that the ruptured steam generator does not fill with water and flood the main steamlines.

This criterion is important because the main steamlines and relief valves are not designed for liquid flow.The accident is assumed to take place at power with the reactor coolant contaminated with fission products corresponding to continuous operation with a limited amount of defective fuel rods. The accident leads to an increase in contamination of the secondary system due to leakage of radioactive coolant from the RCS. In the event of a coincident loss of offsite power, or failure of the condenser dump system, discharge of activity to the atmosphere takes place via the steam generator safety and/or power operated relief valves.In view of the fact that the steam generator tube material is Inconel-600 and is a highly ductile material, it is considered that the assumption of a complete severance of a tube is somewhat conservative.

The more probable mode of tube failure would be one or more minor leaks of undetermined origin. Activity in the Steam and Power Conversion System is subject to continual surveillance and an accumulation of minor leaks which exceeds the limits established in the Technical Specifications is not permitted during unit operation.

The reactor operators are expected to determine that a steam generator tube rupture has occurred, and to identify and isolate the ruptured steam generator on a restricted time scale in order to minimize the contamination of the secondary system and to ensure the termination of the radioactive release to the atmosphere from the ruptured steam generator.

In the following discussions, the steam generator with the ruptured U-tube is referred to as the ruptured steam generator.

The operator is then expected to carry out the appropriate recovery procedures on a restricted time scale in order to terminate the primary-to-secondary break flow before the water level in the ruptured steam generator rises into the main steam system piping. Sufficient indications and controls are provided to enable the operator to carry out these functions satisfactorily.

Assuming normal operation of the various plant control systems, the following sequence of events is initiated by a tube rupture: 15.6-4 Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 CPSES/FSAR Page 44 of 89 The time dependent mass releases used to assess the radiological consequences of the postulated steam generator tube rupture are calculated from the RETRAN02 thermal-hydraulic analysis described above. Time-dependent values of the leakage rate into the ruptured steam generator and the flashing fraction were also used to assess the radiological consequences for the 0-2 hour time period following the event. Following the closure of the atmospheric relief valve block valve, the additional radiological dose is due to the leakage from the primary system into the intact steam generators and the initial concentration of radioactivity contained in the intact steam generators.

Two separate iodine spikes are considered:

Case I A reactor transient has occurred prior to the tube rupture and raised the primary coolant iodine concentration to 60 uCi/gm Dose Equivalent Iodine-13 1 (DEQ I-13 1). The resulting preaccident isotopic iodine concentrations are shown in Table 15.6-3.Case II The reactor trip or primary system depressurization associated with the postulated accident creates an iodine spike in the primary system. The spike is assumed to increase the iodine appearance rate (inleakage from the defective fuel rods to the primary coolant) to times the equilibrium appearance rate. The concurrent iodine spikeappearance rates are presented in Table 15.64.The assumptions below are used to determine the initial primary and secondary activities and to calculate the activity released and the offsite doses for the postulated steam generator tube rupture accident.1. The initial primary coolant iodine activity (i.e., prior to any iodine spike considerations) is assumed to be at 1.0 uCi/gm DEQ I-13 1.E.tuiivankL 1-1M1 (El-) 1J. IXU t Vwi, d 1 1.ath iabk au ahlayw. av" b%11 +11Ulfs' ,r" ...........................

.. -aly * .t a itl by., ^. A,! MT1.Jx~t.1 is the L, Lt. f 1.0 UCJite, h. D 1a. T 1 i.r. .tfm h0A.ML1UA1 at i11 Ivlaiu 3 t aml L 1 1 1 6 D 1 cak aiial(yol diat Duppuit dIL.ThL 30 ARCe aiiay 4. GUSISii3 B Le-Bt 1 R,, OfC.45 uti 1.Tl R.IlliMP6 the 88 teb. :JC AIt L%.Ibf9LV.LfJ Vil tht. %AA...JU11 Of thts MIaill SMl7 1 lmr B- fIl3al aiialy~iS fL ths 3a ARC, all ipUM'd dubeT a"33U1,UL a DEl-.I i 1..in vi. t VU 1.0 uCit.'..2. The primary coolant activity has been leaking into the secondary side at one gpm for a period of time long enough to establish equilibrium activity concentrations is the steam generators.

3. All noble gas activity transported from the primary system to the secondary system and all noble gas and iodine activity initially in the steam region of the steam generators is immediately released to the environment.

The initial iodine activity in the water region of the ruptured steam generator increases over time due to the unflashed portion of the leakage s scharged 4. Due to the pressure differential between thprimary and secondary sides, a fraction of the primary coolant that caktld o the defective steam generator flashes to steam. This flashed fraction does not mix with the steam generator water and, therefore, is not subjected to any iodine removal process in the steam generator.

However, the flashed fraction experiences iodine removal in the condenser when that path is available.

15.6-10 Amendment 98 Attachment 6 to TXX-05127 Page 45 of 89 CPSESJFSAR

5. Radioactive decay of paren dines to noble gas products i considered during the iodine spiking processes as unflashed iodine accumulates in the steam generators.

tk 1.C laJuadivtive Jay uofiis Pali ll iuJ;T f; LUAI3I o V atiHaiivcly a >ouulJ to,,ot deereftse..

he ta.ti of the parent iodin.6. A ground level release is assumed. No credit is taken for radioactive decay or cloud depletion due to ground deposition during plume transport.

7. Worst case, five percentile atmospheric dispersion factors are assumed.3.5E-04l8. A breathing rate of .fL 3/sec is assumed.9. Conservative iodine partition factors of 0.01 and 0.15 are used in the steam generator and condenser, respectively, to account for iodine removal effects within those components.

10. The dose contribution from the intact generator is based upon the effects of iodine spiking and a 1 gpm primary-to-secondary leak rate lasting for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The leakageis assumed to Lll l. d;,l % tL L L,%.. [mix with the secondary side water.18.8 22.0 1 0.0801 Based on th ng mo me thyroid and vole body doses at the EAB are conservatively calcu .ted to be rem and 0 +1 rem, respectively, for the preaccident iodine spike case and 47 rem and .rem, respectively, for the concurrent iodine spike 3.rem and 0.013 case. The thyroid and whole body dose t the LPZ are conservatively calculated to be.(k 89A 87" rem, respectively, r the preaccident iodine spike case and 6-7I 1 I rem a rem, respectively, for th concurrent iodine spike case. As expected, d oses are well below the values of 300 re to the thyroid and 25 rem to the whole body set forth in IOCFRO. 0077 15.6.4 SPECTRUM OF BWR STEAM SYSTEM PIPING FAILURES OUTSIDE OF CONTAINMENT This section is not applicable to the CPSES.15.6.5 LOSS OF COOLANT ACCIDENTS RESULTING FROM A SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY 15.6.5.1 Identification of Causes and Frequency Classification A LOCA is the result of a pipe rupture of the RCS pressure boundary.

For the analyses reported here, a major pipe break (large break) is defined as a rupture with a total cross-sectional area equal to or greater than 1.0 square feet (ft2) This event is considered an ANS Condition IV event, a limiting fault, in that it is not expected to occur during the lifetime of the plant but is postulated as a conservative design basis (see Section 15.0.1).l11. Other assumptions are detailed in Table 15.6-2.15.6-11 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 46 of 89 CPSESJFSAR Large Break LOCA Evaluation Model For the Unit 1 and Unit 2 large break LOCA analyses, the limiting PCT is less than the 1 OCFR50.46 Acceptance Criteria limit of 2200EF. The maximum local and total core metal-water reaction is well below the embrittlement limit of 17% and 1 % respectively as required by IOCFR50.46.

The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. As a result, the core temperature will continue to drop and the ability to remove decay heat generated in the fuel for an extended period of time will be provided.The containment pressure and temperature response to postulated LOCAs are presented in Section 6.2.1.5.Small Break LOCA Evaluation Model Based on the results of the LOCA sensitivity studies [14] the limiting small break was found to be less than a 10 inch diameter rupture of the RCS cold leg. For Units 1 and 2, a small break spectrum analysis showed that the limiting small break was 3 inch. The worst single failure assumed is one safety injection train.Based on the results reported in Appendices 4A and 4B, the calculated PCT resulting from a small break LOCA is less than the limiting large break PCT and remains well below the Acceptance Criteria limits of 10CFR50.46.

15.6.5.4 Environmental Consequences To demonstrate in a conservative manner that the operation of a nuclear power station does not present any undue radiological hazard to the general public, a hypothetical accident involving a gross release of fission products is evaluated.

No mechanism for such a release has been postulated because it would require a number of simultaneous failures to occur in the engineered safety features.

The core fission product inventory is assumed to be released into the containment as described in A.- bef].Numerical values for the total core fission product inventory of the isqpes considered in calculating the radiation doses are listed in Table 15.6-8. 1Regulatory Guide 1.195 [34].1 The radiological evaluation of this accident is divided into two parts: internal (thyroid)dose from inhalation of iodines in the leakage plume, and external (whole body) exposure as a result of immersion in the leakage plume.The radiological consequences due to the release of core fission products during a postulated loss-of-coolant accident are evaluated in the following sections: 1. Radiological consequences of containment leakage The integrated thyroid doses and the integrated whole body doses are calculated using methods and assumptions in conformance with Regulatory Guide 4# The assumptions used in the analysis are listed below.15.6-17 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 47 of 89 CPSESIFSAR

a. Twenty-five percent of the equilibrium radioactive iodine inventory developed from maximum full-power operation of the core is immediately available for leakage from the reactor containment.

Of this 25 percent, 91 percent is in the form of elemental iodine, 5 percent is in the form of particulate iodine, and 4 percent is in the form of organic iodides.b. All (i.e., 100 percent) of the equilibrium radioactive noble gas inventory developed from maximum full-power operation of the core is immediately available for leakage from the reactor containment.

c. The effects of radiological decay during holdup in the containment are taken into account.d. The containment volume is divided into separate regions by concrete floors at different elevations (see Section 6.5.2). A radial gap between the concrete floors and the inner wall of the Containment Building permits a limited amount of convective mixing between these regions. The region not covered by containment spray is treated as a separate unsprayed volume which is assumed to mix with the volumes in the sprayed areas at a mixing rate of two turnovers per hour.The Containment Spray System is actuated by a high containment pressure signal. For a discussion of the sequence of events of spray system operation, see Section 6.5.2. A sodium hydroxide spray is used to reduce the amount of fission product iodine available for release during theh o LOCA. The containment spray solution is assumed to interact with the el elemental iodine and particulate iodine. The mathematical w hich@ the iodine spray removal f presented in Section Foefficients are 6.5.2. For each region the calculated elemental iodine removal coefficients are above 10 hrf .The removal coefficient for elemental IThe calculated -Le used in the offsite dose calculation is limited to a max of IO hfT2217-i value of Nfr the particulate iodine removal coefficient, Md...gh higll MIL..VM lv%.Iffle 11G.W u a Cahwuiahve The elemental iodine removal effectiveness may be expected 50 is reached diminish after the concentration in the containment atmosphere has been at which time reduced by several orders of magnitude.

The elemental iodine removal the particular effectiveness of the spray system is conservatively assumed to cease after a removal decontamination of 100 in the containment atmosphere has been achieved.

coefficient is reduced by a factor of 10.e. The iodine and noble gases available for release to the environment are assumed to leak from the Containment at a maximum leak rate of 0.10 percent of the containment volume per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 0.05 percent of the containment volume per day for the duration of the accident.f. The duration of the accident is considered to be 30 days.15.6-18 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 48 of 89 CPSES/FSAR

g. A ground-level release is assumed. Atmospheric dilution factors are discussed in Section 2.3 [23] and listed in Table 15.6-9.h. No credit is taken for depletion of fission products in the plume due to ground deposition or radioactive decay in transit.i For the first 8 ho after the accident, the breathing rate of persons offsite is assumed to be i:* x 10-4 cubic meters per second (m3/sec).

Eight to 1.8 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate is assumed to be + x 10-4 m3/sec. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> through 30 days after the accident, the rate is assumed to be M23 x 10-4 m3/sec.j. The mathematical model and dose conversion factors presented in Appendix 15B are used for evaluating the radiological consequences of the LOCA k. Other assumptions are detailed in Table 15.6-9.Using the assumptions presented above and the mathematical models presented in Appendix 15B, the doses at the EAB were conservatively calculated to be ; 349.2 rem to the thyroid and t rem to the whole body; the doses at the LPZ were conservatively calculatedto be 6 em to the thyroid anljt22jem to the whole body. 10.258 2. Radiological consequences of engineered safety features equipment leakage outside containment.

Following a postulated LOCA, a potential source of fission product release is the leakage of water from engineered safety features (ESF) equipment located outside the containment.

Such leakage could occur during the recirculation phase through components such as pump flanges, valves, and heat exchangers.

The fission products could then be released from the water into the atmosphere resulting in offsite radiological consequences that contribute to the total dose from the LOCA.An analysis of the offsite effects attributable to ESF equipment leakage is performed based on the following conservative assumptions:

a. 50 percent of the halogens originally present in the core are intimately mixed with the coolant water and are assumed to be available for release through ESF equipment outside containment (see Table 15.6-10).by ll u .H.C lVIVUV9 E, %., t~u9usvss IVU111 h l the Us*le:*ing w~a 15.6-19 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 49 of 89 CPSES/FSAR Mb v.- The leakage from all ESF components is conservatively assumed to start 10 minutes after the LOCA and continue for the duration of the accident at a rate of 2 gallons per minute.IcJd- An iodine partition factor of 0.1 is assumed. This factor is taken as the fraction of iodine in the leakage that becomes airborne.Md VI Gaseous radioactivity released to rooms housing the leaking components is considered to be immediately swept away by the ventilation system and released to the atmosphere. (See Section 9.4.5 and Figures 1.2-16, 1.2-35 and 9.4-9).MIe An iodine adsorber efficiency of 95 percent is applied since the ventilation exhaust passes through HEPA filters and iodine adsorbers prior to release to the atmosphere.

The iodine adsorbers are designed to the requirements of NRC Regulatory Guide 1.52 (See Appendix lA(B)) as discussed in Section 9.4.3.Lay~ No credit is taken for an elevated release; all meteorological parameters are considered to be identical to those previously defined in this section.l 0.0426 1Based upon the foregoing model, the thyroid and whole body do tributions due to ESF equipment leakage are conservatively calculated to be rem and RELOCATE and 4 rem, respectively, for the EAB. The LPZ doses are conservatively calculated i INSERT #6 HERE as to be +44 rem to the thyroid and 0 n l re m to the whole body.#3 126.31 10.0283 1 4 e Total d ose due to a postulated LOCA 9 e d, and releases through the containment pressure relief line.The total dose attributed to a ostulated LOCA is thyombined doses due to containment leakage ma ESF equipment leakage/Ahe combined EAB doses are 63.0 -99 rem to the thyroid and + t rem to the whole body. The combined LPZ doses i rem to the thyroid and i+ rem to the whole body. As expected, the doses Lo are below the values set fortin OCFR100.I0.291 The dose to personnel engaging in mineral extraction operations within the exclusion area, in the event of a postulated LOCA, would be less than the dose values of 300 rem to the thyroid and 25 rem to the whole body set forth in IOCFR100.@3 t4 Dose to the control room occupants In the event of a Design Basis Accident (DBA), the safety injection actuation signal or a high radiation signal from the control room air intake monitors will initiate emergency recirculation and pressurization of the Control Room air conditioning system. Later, the emergency ventilation air makeup system can be brought into operation as described in Section 9.4.1.15.6-20 Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 50 of 89 CPSESJFSAR The control room doses were analyzed for various design basis accidents.

It was determined that the LOCA doses represent the limiting case. Therefore the methodology and the doses calculated for the LOCA are reported here.The following assumptions are applied in the calculations of the dose to the control room occupants following the LOCA: a. The basic assumptions presented in Items 1 and 2, above, are applied, except a constant breathing rate of &+.4 x 104 m 3/sec is assumed throughout the accident.

EM3 b. The control room pressurization (air intake) and recirculation iodine adsorbers are assigned a 99 percent decontamination efficiency for both elemental and organic iodines in accordance with Table 2 of Regulatory Guide 1.52 (See Appendix 1A(B)). The pressurization adsorbers are arranged in series with the recirculation adsorbers (see Figure 9.4-1)during the emergency pressurization mode, thus providing an equivalent decontaminating efficiency in excess of 99 percent for both elemental and organic iodines from the pressurization makeup air.c. The control room air-conditioning system runs either in the emergency recirculation mode or the emergency ventilation mode during a LOCA.d. During the emergency recirculation mode of operation, a constant air intake flow rate of 800 f13/min is assumed. This makes up for losses caused by leaks and maintains the control room atmosphere at a positive pressure of 0.125 inch water gauge relative to adjacent areas.Since both recirculation trains are actuated by the safety injection signal, the outside air intake flow rate during dual train operation is conservatively estimated to be 1600 cfm. If both trains are assumed to operate for one hour, the calculated thyroid dose would decrease, due to the additional iodine filtration.

The calculated whole body gamma and beta skin doses would increase slightly due to the additional intake of outside air. In both cases, the calculated doses remain below the limits specified in 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants," Criterion

19. The reported control room doses are the higher values of the two cases.e. During the emergency ventilation mode of operation, 3800 f13/min of outdoor air is used to introduce fresh air into the control room.f. The emergency ventilation mode of operation is under administrative control so that the dose to the control room occupants is minimized, and the need for air change is satisfied.

The operating mode sequence used in this analysis is as follows: 15.6-21 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 51 of 89 CPSES/FSAR 1.A = projected area of Containment Building 13265 M 2) for the containment leakage source and 2088 m for the ESF leakage source Table 15.6-12 summarizes the X/Q values calculated utilizing this expression.

The CPSES CREFS design is zone isolation, with filtered recirculation air, and with a positive pressure.

This design maximizes the iodine protection factors and minimizes the dose from iodine. The total unfiltered infiltration rate in the control room is conservatively assumed to be +2 27 cfm, including 10 cfm due to ingress/egress vnd 2 cfm leakage from the ductwork passing through the control room pressure boundaI Filtered inleakage through the c osed dampers due to the pressure diflrential is also included.

The damper leakage air will be filtered by th recirculation filtration units.l, and 15 cfm from other sources.Because the control room door ingress/egress is to a stairwell which is equivalent to a two-door vestibule, backflow will not occur with the CPSES CREFS design and the 10 cfm is not applicable per SRP 6.4. The ductwork has all wefded joints which were leak tested prior to operation.

Therefore, the assumed unfiltered inleakage from adjacent areas is conservative with respect to the SRP review criteria.p j. Habitability of the control room is based on the following occupancy factors: Time Period 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1 to 4 days 4 to 30 days Occupancy Factor 1.0 0.6 0.4 k. The air volume in the control room used to determine exposures to operators is 423,032 fM3.l. The models for the major contributors to the control room dose are provided in Appendix 15B.Using the above assumptions and procedures, the thyroid dose is conservatively calculated to be 4 rem in the control room for the duration of the accident.

The thyroid dose can be further reduced by the m use of the available respiratory protection equipment.

The tot body gamma dose is conservatively calculated to be Em.This calculated dose includes whole body dose contributions from containment sources (both direct and scattered radiation), the external passing cloud, control room atmosphere, activity buildup on filters, and streaming through doors and penetrations.

These calculated doses are less than the limiting values specified in lC.rn_0, AylJe.IA A, "C a "Tr C 1;it 1;. N.i " F s lAi., L./ 1. Reglatory Guide 1.195.The skin dos eived in the rI room during the accident period is conservativell calculated to be rem. This calculated beta skin dose is less than the rem limit allowedif special protective clothing and eye protection we used. 9lhvr~fbrC speNI protective clothing and eye s n However, 15.6-23 Amendment 99 by Regulatory Guide 1.195 Attachment 6 to TXX-05127 Page 52 of 89 CPSES/FSAR protection are provided for use, if required, to reduce the beta skin doses to the operators U itlin ilM.tnb ak limlX 1 1 i i;Xedane wait i86 Roc5R8o,*ppen 1 i. A, 8DC 19.Environmental consequences of containment purging to control containment hydrogen concentration after a LOCA Purging of the containment atmosphere provides a backup method for controlling potential hydrogen accumulation in the Containment following a postulated LOCA. The use of the hydrogen purge system (see Section 6.2.5.2.2) is precluded by redundant electric hydrogen recombiners located in the Containment Building (see Section 6.2.5.2.1).

The electric hydrogen recombiners are the primary means of controlling post-LOCA hydrogen buildup. Thus, an analysis of the radiological conseouences of containment nurwine is not nrovided.-9 Environmental consequences of releases through the containment pressure relief line in the event of a LOCA RELOCATE to PAGE 15.6-20 as #3 An analysis of the radioactive effluents escaping the Containment to the environment after a LOCA, via the line through the controlled access area exhaust system, was performed using the following assumptions:

a. The maximum containment air/steam mass release to the environment was conservatively calculated assuming containment pressures for a large break LOCA, and critical flow via the 3-3/8 inch orifice plate at the inlet to the pressure relief system ductwork.

No credit was taken for line losses in the ductwork or two butterfly valves.It was conservatively assumed that all iodine and noble gas activity in the primary coolant was instantaneously released to the containment atmosphere.

b. Only reactor coolant activity is assumed to be released.

HeS il eg&fine b f 1 inc. to 6 in..3 Vr a.aL..c.?f ..1f J%..IV Il LSJ6IL1'-

Op % , V U3a ai~ J k~. 1 J~ -.L 1..kA 11 1E,~ LIL%. P LU M MYk~.....1 I ..I cnvnif-Lativuuina l h1ate1 i Tatbk. 5.6 3. The ngbac s activity concentrations are presented in Table 15.14.d. The containment pressure relief line isolation valve closure time including instrumentation delays will not exceed 5 seconds. Tl. indionetive fiss.i pjjk>a uj%., 1 J;hfL; 1 , fbu a FVsiVu U filui. ...1 ual k a i y 1.Xwal aILaLa br-Cak LOCE ie.aiut. trip Dirial thriUI (plLus 5 sA.AIhJ3.

The analysis did not consider the reduction of mass flow during the valve closure time of five seconds; full flow was assumed until the lapse of five seconds.e. No credit was taken for radioactive decay.The primary coolant iodine activity is assumed to be 1.0 micro curie per gram dose equivalent 1-131. The noble gas activity is based on operation with 1% fuel defects.f. No credit was taken for an elevated release.--15.6-24 Amendment 97 February 1, 2001

-Attachment 6 to TXX-05127 Page 53 of 89 CPSES/FSAR 17a. Deleted 18. Deleted 19. Deleted 20. Deleted 20a. Deleted 21. DiNunno, J.J., Anderson, F. D., Baker, P.E. and Waterfield, R. L., "Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, U. S. Nuclear Regulatory Commission, Division of Licensing and Regulation, March 1962.22. Postma, A. K., and Pasedag, W. F., "A Review of Mathematical Models for Predicting Spray Removal of Fission Products in Reactor Containment Vessels," WASH-1329, U. S. Nuclear Regulatory Commission, Accident Analysis Branch.23. Meteorology and Atomic Energy," U. S. Nuclear Regulatory Commission, Division of Technical Information, 1968.24. Murphy, K. G. and Campe, Dr. K. M., ANuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19,@ U. S. Nuclear Regulatory Commission

25. Deleted 26. Boatwright, W. J., Maier, S. M. and Lo, S. S., "Design Basis Analysis of a Postulated Steam Generator Tube Rupture Event for Comanche Peak Steam Electric Station, Unit 1", RXE-88-101, TU Electric, March, 1988.27. Deleted 28. Deleted 29. Deleted 30. Deleted 31. Deleted 32. Deleted 33. Deleted 134. Regulatory Guide 1.195, "Methods and Assumptions For Evaluating Radiological I Consequences Of Design Basis Accidents At Light-Water Nuclear Power Reactors," May 2003.15.6-26 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 54 of 89 CPSES/FSAR TABLE 15.6-2 (Sheet 1 of 2)PARAMETERS FOR POSTULATED STEAM GENERATOR TUBE RUPTURE ACCIDENT ANALYSIS Parameter Value 1. Data and Assumptions Used to Estimate Radioactive Source Term For the Postulated Steam Generator Tube Rupture Accident a. Steam generator tube leak prior to and during 1.0 accident (gpm)b. Offsite power lost 2. Data and Assumptions Used to Estimate Activity Released to the Atmosphere

a. Iodine partition factor in steam generators prior to 0.01 and during accident b. Iodine partition factor in condensers prior to 0.15 accident C. Duration of plant cooldown by secondary system 8 after accident (hr)3. Dispersion Data a. Exclusion area boundary and low population zone 2080 m and 4 distances miles b. x/Q (sec/m3)0-2 hours 1.6x 104 0-8 hours 2. x10 5 4. Dose Data a. Method of dose calculation See Appendix 15.B b. Dose conversion assumptions See Appendix 15.B Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 55 of 89 CPSES/FSAR TABLE 15.6-2 (Sheet 2)PARAMETERS FOR POSTULATED STEAM GENERATOR TUBE RUPTURE ACCIDENT ANALYSIS Parameter Value c. Doses (rem)With preaccident iodine spike@ exclusion area boundary (0 -2 hours)Thyroid inhalation Whole body gamma@ low population zone (0 -8 hours)Thyroid inhalation Whole body gamma With concurrent iodine spike@ exclusion area boundary (0 -2 hours)Thyroid inhalation Whole body gamma@ low population zone (0 -8 hours)Thyroid inhalation Whole body gamma 46- 22.f.99 A -20.013 10.0771 6.5 1.4o;J.J As V -0.012 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 56 of 89 CPSES/FSAR TABLE 15.6-3 PREACCIDENT IODINE SPIKE CONCENTRATION IN THE PRIMARY COOLANT (a)Spike Isotope Preaccident Iodine Concentration (RCi/gm)1-131 1-132 I-133 1-134 I-135~94 72.&93.4 (a) This concentration corresponds to 60 gCi/gm Dose Equivalent 1-13 1, the maximum allowed at Rated Thermal Power by the Technical Specifications.

Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 57 of 89 CPSES/FSAR TABLE 15.6-4 IODINE APPEARANCE RATE TO THE REACTOR COOLANT AFTER THE ACCIDENT Iodine Appearance Rate(a)-nt (C 1-131 +46 1-132 e 48 1-133 9 28 1-134 4-5e 1-135 3 219 (a) Corresponds to 5 times the equilibrium appearance rate based on operation with reactor coolant activity at 1.0 uCi/gm Dose Equivalent 1-131.1115TE. The Priimiary

Uun4u1l TS li,,.il hb b. dhiangvd (i'MuiLcd) tu 6.4. uaii'g,, Dusu Equivaeldit I-13t1 (DEI- 13). Numenc, dh 1 e lapt ativ dvl dltalya lYS vcr T 1...0 ,R ... D .-L .1 .,f ..^ .TIIL urr.%IL uL51UtawUd tw be conlservativ, uashig t.v Uneblf LnI- 13. T hec Usilty wavMsptullft,

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• UkE DEI-L S6 Mtilatyuh, 1.0 zCL'.a DEMM Vatukaw if 0., t.. T aS.upil MaiiL, th _S6*w *RE bmft Thr1i. W it th. -~ti__ -ths Mah Stea hint., BL ek Sia u i -lrip va 8 ufem u.i ubpuz tl.. dulamar.-)E * * .i fil T _

• Oft .* se.i .Amendment 98 Attachment 6 to TXX-05127 Page 58 of 89 CPSES/FSAR TABLE 15.6-9 (Sheet 1 of 3)PARAMETERS FOR POSTULATED LOCA ANALYSIS(a)

Data and assumptions used to estimate radioactive source from postulated accidents a. Power level (MWt) 3565 b. Pe lrc offafl dfid 1,--e-. Plateout of iodine activity 50%released to containment

__ Activity released to containment and available for release Noble gases 100% of cor lodines 25% of core Ito environment I.e inventory inventory A } Iodine fractions (organic, Regulatory Guide e Re m PivitXiv 1 l 1.195 elemental, and particulate)

+.2. Data and assumptions used to estimate activity released a. Containment volume l1.706l Directly sprayed volumes t~y x 106 f3 Unsprayed volumes 1 .x 106 ft 3 affected by spray Total free volume 3.031 x 106 ft 3 b. Mixing rate between 2 turnovers per hour sprayed and unsprayed volumes (a) Per Regulatory Guide 44 1.195 Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 59 of 89 CPSES/FSAR TABLE 15.6-9 (Sheet 2)PARAMETERS FOR POSTULATED LOCA ANALYSIS(a)

c. Spray removal coefficient Elemental iodine 10 hrf Organic iodines 0 hrf Particulte iodine f' hr-i for DF less than or equal to 50 14 hr-1 I [for DF greater than 50 d. Effective decontamination 100 factor of spray Ifor elemental iodine e. Containment leak rate 0.1% containment volume per day, (0<t<24 hrs) 0.05% of containment volume per day, (t > 24 hrs)f. Duration of removal effectiveness Elemental iodine Particulate iodine 3. Dispersion data a. Exclusion area boundary and LPZ distances b. x/Q's (for time intervals of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 4 days, 30 days)Ii hrJ hrl5hr hr 28m a hr 4i 2080 m and 4 miles (DF of 100 attained)I(DF of 50 attained)

I Isprays assumed to be terminated I ( EAB, onsite 5 percentile data 1.6 x 104 sec/m 3 (0-2 hrs)@LPZ, onsite 5 percentile data l 2.4e-05 l:-- -I sec/M 3 (0-8 hrs)1 .6e-05 l4. sec/M 3 (8-24 hrs)6.2e-06 l sec/mi 3 (1-4 days)l1 .7e-06 AI sec/M 3 (4-30 days)Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 60 of 89 CPSES/FSAR TABLE 15.6-9 (Sheet 3)PARAMETERS FOR POSTULATED LOCA ANALYSIS(a)

4. Dose due to containment and ESF equipment leakage I and the containment pressure relief line.a. Method of dose calculation See Appendix 15B b. Dose conversion assumptions

c. Doses See Appendix 15B@EAB, (0-2 hrs) I=thyroid = 6.9 rem 63.0 whole body gamma = +-4 rem 0.79@LPZ (0-30 days)thyroid = 3 rem 47.0 whole body gamma = +-+ rem 0.29 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 61 of 89 CPSESIFSAR TABLE 15.6-10 ACTIVITY AVAILABLE FOR RELEASE VIA ESF COMPONENTS AT T=O FOLLOWING LOCA ISOTOPE ACTIVITY IN CURIES I-131 4~9it10z 1-132 I-133 1-134 I-135 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 62 of 89 CPSES/FSAR content of a single waste gas decay tank inventory of 200,000 Ci of noble gases (considered as dose equivalent Xe-133) is alarm in me Control Koom. I ne gaseous activity in Me tanks tMen decays wnile the other tanks in the system are bein filled with gaseous radioactivity.

1. • 5# The maximum activity that can be released s a result of a gas decay tank rupture is the activity stored in one gas decay tank immediat ly after it has been isolated from the GWPS.The entire content. of assumed to be released to the Auxiliary Building, and all of the noble gases are assumed to leak from the building at ground level over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period. Based on this model and an onsite 0 to 2 hr atmospheric dilution factor (1.6x104 sec/M 3), the whole body dose at the nearest point on the exclusion area boundary (EAB) 0 is conservatively calculated to be OH-S iem. Tnis dose is substantially below the 25 rem whole body value set forth in I OCFR100; it may be concluded that such an incident would not interrupt or restrict public use of areas beyond the EAB.X apalf1tilltt "tu d , sonlvo*sL thi Singlelb VaVo ftst ge.w deellyirk nV efllv ell~ of -vvvvv i 01F AIJJIu.. ECo (evlaideled no iie e-.. I v 1 a) MI U T 1u ul( ia i a ill" v a 1 v elJ etleal.ittLe V1v411v. buvy uria dv 1 O.f 0.Z 111 at l 1%, LAD, uail,6 HIv lzVIh LUJ, U r~tv .iil r 'Ri ran't, 15.7.2 RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE 15.7.2.1 Identification of Causes and Accident Description The accident is defined as the uncontrolled atmospheric release from the 30,000 gallon floor drain tank due to the postulated rupture of the tank. This tank is the highest potential atmospheric release source term because of its large volume and the fact that it is assumed to be 80 percent full of reactor coolant.15.7.2.2 Analysis of Effects and Consequences The activity in the 30,000 gallon floor drain tank, the assumptions on which the activity is based, and the radiological consequences resulting from the release of the activity are discussed in Section 15.7.2.3.15.7.2.3 Environmental Consequences This analysis assumes that the plant has been operating with 1 percent failed fuel for an extended period sufficient to achieve equilibrium radioactive concentrations.

Floor drain tank III is assumed to contain the inventory as indicated in Table 15.7-3. The entire contents of the tank are assumed to be released to the atmosphere at ground level over a 2 hr period. Other conservative assumptions are detailed in Table 15.74.Based on the foregoing model, the thyroid and whole body doses at the EAB are conservatively calculated to be 544 rem and .rem, respectively.

The doses from this accident are well withinlthe values et forth in 10CFR100.E 32303 15.7-2 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 63 of 89 CPSESJFSAR 15.7.3 POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID TANK FAILURES The analysis is presented in Section 2.4.12 and 2.4.13.3.15.7.4 DESIGN BASIS FUEL HANDLING ACCIDENTS 15.7.4.1 Identification of Causes and Accident Description The accident is defined as dropping of a spent fuel assembly in the Containment Building or spent fuel pool fuel storage area floor resulting in the rupture of the cladding of all the fuel rods in the assembly despite many administrative controls and physical limitations imposed on fuel handling operations.

All refueling operations are conducted in accordance with prescribed procedures under the direct surveillance of a supervisor.

15.7.4.2 Analysis of Effects and Consequences Method of Analysis The method of analysis used for evaluating the potential radiological consequences of a fuel handling accident is in compliance with Regulatory Guide +.6vowelc rt R, ! 1.195.piUV;J;UIAO lL; ill A~~kd;f 1A(D). A two hour, ground level release is assumed for te analysis.The following assumptions are postulated in the calculation of the radiological consequences of a fuel handling accident: 1. The accident occurs at 100 hr following the reactor shutdown, the minimum time at which spent fuel could be first moved into the fuel storage area.2. The accident results in the rupture of the cladding of all fuel rods in a single assembly.3. The damaged assembly is, coincidentally, the one operating at the highest power level in the core region to be discharged.

4. The power in this assembly, and the corresponding fuel temperatures establish the total fission product inventory and the fraction of this inventory which is present in the fuel pellet-cladding gap at the time of reactor shutdown.5. The fuel pellet-cladding gap inventory of fission products is released to the refueling cavity or spent fuel pool at the time of the accident.6. The refueling cavity or spent fuel pool retains a large fraction of the gap activity of halogens by virtue of their solubility and hydrolysis.

Noble gases are not retained by the water as they are not subject to hydrolysis reactions.

Fission Product Inventories 15.7-3 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 64 of 89 CPSESIFSAR l1.195.l The actual fission product gap inventory in the ue ssembly is dependent on the linear heat generation rate of the assembly and the temyerature of the fuel. However, the gap inventories assumed in fuel handling accident nalyses were based on the conservative guidance contained in Regulatory Guide Table 15.7-6 lists the fuel assembly fission product activities at the time of the fuel handling accident.

These activities are consistent with the assumptions used in analyzing the environmental consequences of postulated fuel handling accidents detailed in Section 15.7.4.3.Iodine Decontamination Factors An experimental test program [I] was conducted to evaluate the extent of removal of iodine released from a damaged irradiated fuel assembly.

Iodine removal from the released gas takes place as the gas rises through the body of solution in the fuel storage area to the pool surface. The extent of iodine removal is determined by mass transfer from the gas phase to the surrounding liquid and is controlled by the bubble diameter and the contact time of the bubble in the solution.In order to obtain all the necessary information regarding this mass transfer process, a number of small scale tests were conducted, using trace iodine and carbon dioxide in an inert carrier gas. Iodine testing was performed at the design basis solution conditions (temperature and chemistry) and data were collected for various bubble diameters and solution depths. This work resulted in the formulation of a mathematical expression for the iodine decontamination factor in terms of bubble size and bubble rise time.Similar tests were conducted with carbon dioxide in an inert carrier, except that the solution temperature and chemistry were patterned after that of a deep pool where large scale tests were also performed with carbon dioxide. The small scale carbon dioxide tests also resulted in a mathematical expression for the decontamination factor in terms of bubble size and bubble rise time through the solution.To complete the experimental program, a full size fuel assembly simulator was fabricated and placed in a deep pool for testing, where the gas released would be typical of that from the postulated damaged assembly.

Tests were conducted with trace carbon dioxide in an inert carrier gas and overall decontamination factors were measured as a function of the total gas volume released.

These measurements, combined with the analytical expression derived from small scale tests with carbon dioxide, permitted an in-situ measurement of both the effective bubble diameter and rise time, both as a function of the volume of gas released.

Having measured the characteristics of large scale gas releases, the decontamination factor for iodine was obtained, using the analytical expression from small scale iodine testing.Decontamination Factor = 7.3 e 0.3 1 3 t/d where t = rise time (sec)15.7-4 Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 65 of 89 CPSES/FSAR fior A fowl rod t preSSvre o f IesS than /SO9ps4 d = effective bubble diameter (cm)The overall test results clearly indicate that iodine will be readily removed from the gas rising through the fuel storage area solution and that the efficiency of removal will depend on the volume of gas released instantaneously from the fuel void space.With consideration given to the total quantity of gas released from a fuel assembly, a typical pool decontamination factor for iodine is at least 60 for the 26 foot depth. Ibo However, a much lower decontamination factor o1 s conservatively selected to provide for deviations in the factors which contro iodine absorption by the pool water.15.7.4.3 Environmental Consequences

,.15.7.4.3.1 Postulated Fuel Handling Accident Outside Containment The analysis of a postulated fuel handling accident is performed as follows: 1. The accident is assumed to occur 100 hr after plant shutdown.2. All of the rods in one fuel assembly are ruptured.3. The damaged assembly is the highest powered assembly in the core region to be discharged.

The values for individual fission product inventories in the damaged assembly are calculated assuming full power operation at the end of core life and a 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> shutdown.

A radial peaking factor of 1.65 is used.4. The minimum water depth between the top of the damaged fuel rods and the spent fuel pool surface is 5. All of the gap activi i the damaged rods is released to the spent fuel pool. This activity consists of, _Frcent of the total noble gases other than krypton-85,,0) percent of the krypton-85, an rcent of the total radioactive iodine in the rods at the time of the accident.

Th gap activities released as a result of this accident are presented in Table 15.7-6. (6. All activity released from the spent fuel pool is release at groun eve to e environment over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period.7. The iodine gap inventory is composed of inorganic species (99.75 percent) and organic species (0.25 percent).8. The overall decontamination factor for the spent fuel pool is 9. No credit is taken for iodine filtration by the primary plant ventilation system.10. Atmospheric diffusion conditions are assumed to be the 0 to 2 hr ground level case.15.7-5 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 66 of 89 CPSESJFSAR The parameters used for this analysis are listed in Table 15.7-7.22.0 0.087 Based on the foregoing assumptions

^Pe thyroid a ole body doses at the EAB are conservatively calculated to be 5+/- rem and rem, respectively.

The corresponding doses at the LPZ are conservatively calculated to be em and 69 S l6)firem.

The calculated doses are within the values set forth in 10CFRl 1 3l3 15.7.4.3.2 Postulated Fuel Handling Accident Inside Containment An analysis of the radiological consequences of a fuel handling accident inside the Containment Building would use the same assumptions and yield the same results as those of a fuel handling accident outside the Containment Building.

The accident described in the preceding section is considered to represent the limiting case; therefore, no specific analysis of such an accident inside the Containment is provided.15.7.4.4 Conclusions The possibility of a fuel handling accident is relatively small due to the many physical, administrative, and safety restrictions imposed on fuel handling operations.

However, based on conservative design basis parameters, the calculated doses from a postulated fuel handling accident are within the values set forth in I0CFRO00.15.7.5 SPENT FUEL CASK DROP ACCIDENTS The Comanche Peak Steam Electric Station Fuel Handling Building overhead crane satisfies NUREG-0554 single-failure-proof requirements and is designed to the requirements of seismic Category I (See Section 3.2). As such it can retain the maximum design load during a Safe Shutdown Earthquake and remain in place under all postulated seismic loadings.

The crane is also provided with interlocks which prevent a fuel cask from being lifted more than 29.25 ft above floor elevation or from passing over the new fuel storage area during the spent fuel cask mode of operation (see Section 9.1.2.2).

The crane does not pass over the spent fuel pool in any mode of operation.

Based on this design approach, the radiological consequences of a spent fuel cask drop accident need not be evaluated.

REFERENCES

1. Bell, M. J., Duhn, E. R., Locante, J. and Malinowski, D. D., "Radiological Consequences of a Fuel Handling Accident," WCAP- 7518-L (Proprietary) and WCAP-7828 (Non-Proprietary), June 1970.2. "Radiation Analysis Manual, Standard Plant Model 412, "Rev. 3, Westinghouse Electric Corporation, November 1978.3. Killough, G. G., Begovich, C. L., Sjoreen, A. L. and Bell, L. W., "A Guide for the TACT III Computer Code," NUREG/CR-3287, Oak Ridge National Laboratory, May 1983.15.7-6 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 67 of 89 CPSESIFSAR TABLE 15.7-1 (Sheet I of 2)PARAMETERS FOR POSTULATED WASTE GAS DECAY TANK RUPTURE ACCIDENT 1. Data and assumptions used to estimate radioactive source from postulated accidents a. Power level (MWt) 3565 b. Percent of fuel defected I (%)c. Release of activity by Table 15.7-2 Values increased to equ nuclide of dose equivalent Xe-1V 2. All pertinent data and assumptions used to estimate activity released a. The maximum content of the decay tank is considered for computing noble gas inventory b. Radiological decay is considered for minimum time required to transfer the gases from the primary system to the decay tank c. Entire content of one tank is assumed to be released to the building due to failure d. All of the noble gas is assumed to be released to the atmosphere over a 2 hr period 3. Dispersion data a. Exclusion area boundary 2080 m and 4 miles (EAB) and low population zone (LPZ)distances b. x/Q 1.6 x I0Osec/m3 (0 -2 hr)al 200,00 Ci 33 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 68 of 89 CPSES/FSAR TABLE 15.7-1 (Sheet 2)PARAMETERS FOR POSTULATED WASTE GAS DECAY TANK RUPTURE ACCIDENT 4. Dose data I Appendix 15B I a. Method of dose calculations

b. Doses ftSuiatlly Guild 1.240 dose = 0.191@EAB, whole body rem (gamma dose).,riau , .i .1%J 1 .. A 1.. .)U. ., .1I 1.. ., t' ..1 1 1 1 1 .t.fiVILIWl~at llgIUU3i.

tndint~ivf

• Ya 1 1 l, Siv1 11ual f2].Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 69 of 89 CPSES/FSAR bo~seA on TABLE 15.7-2 GAS DECAY TANK INVENTORY FOR ACCIDENT ANALYSIS (ONE GAS DECAY TANK)Isotope Tank Activity (Ci ° C Kr-85 2.72x 103 +.49 X to Kr-85m 3.23x 10 2 5 :3b3 x IO Z Kr-87 4.42x10 1 7.30 gX I l Kr-88 4.08x 102 (0,74 X lo l-Xe-131m 6.80 x IO' i,1 Y2 X lC3 lXe-133 7.82x 10 4 1.29 X to5 l Xe-133m 5.10x 102 8.42 X 102 lXe-135 1.67x IO 2 .7 x I0 J Assumptions

1. 3565 MWT 2. 40 Year continuous operation at I percent fuel defects 3. 100 percent stripping efficiency

4. Volume control tank purge rate of 0.7 standard cubic feet per minute (scfm)5. Tanks switched at regular intervals 6. 8 gas decay tanks shared between two units Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 70 of 89 CPSES/FSAR TABLE 15.7-3 (Sheet I of 3)MAXIMUM RADIOACTIVITY IN A FLOOR DRAIN TANK FOR ACCIDENT ANAT V.Y1.R-Isotope Br-83 Br-84 Br-85 I-130 1-131 1-132 1-133 1-134 1-135 Rb-86 Rb-88 Rb-89 Cs-134 Cs-136 Cs-137 Cs-138 Ba-137m H-3 Cr-51 Mn-54 Mn-56 Fe-55 Fe-59 Co-58 Co-60 Floor Drain Tank Activity (Ci)8.63 x 10°4.27 x 100 5.45 x IO-0 1.91 x 100 2.54 x 102 2.54x 102 3.82 x 102 5.18 x 101 2.09 x 102 2.00x 100 4.36x 102 1.91 x 1O0 2.09 x 102 2.63 x 102 1.36 x 102 8.72x 10'1.27 x 102 3.18 x 102 5.00 x 10'4.00x 10-2 1.82 x 100 1.82 x 10'4.72 x 10.2 1.36x 100 1.73x 10'PIMor DrC61 Tqnk Acfiv#. v tLsed -to Ddtemihe -the 'Otsfte, Atli codtri Rcom Dv°Ze C;2ce;Ki/A HA/NA N/A Z ari x IO'5 .1c X10 ' C tMA NI/A N/A s/A W1/A N /A N/A M/4 tJ/A m/A N/A W/A N/A NIA NIA^(6 k I Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 71 of 89 CPSES/FSAR TABLE 15.7-3 (Sheet 2)Isot Sr-4 Sr-5 Sr-'Sr-'y g y g y g y g Zr-!Nb-Mo Tc-Ru-Ru-Rh-Rh-Ag-Te-Te-Te-Te-Te-Te-Te-Te-MAXIMUM RADIOACTIVITY IN A FLOOR DRAIN TANK FOR ACCIDENT ANALYSIS F/or Dr;i 7n k (jcv t Floor Drain Ta / seJ to Deteriine -theitsfs Activity (Ci) and Co-tri 1ROOA Dos ieGAS s(C)89 3.91x10 1 I-/A 90 1.09x 10-2 tQ/A 9 1 5.63 x 10-' N /A 92 1.18x 10' "/&'0 3.09 x 10 3 W1 /A Pim 3.00 x 10 I1 1 5.18 x 10-2 N /a 12 1.09 x 10'- N/A'3 3.45 x 10 2 I' /A 95 5.90 x 102 1N/A.95 5.90 x 10-2 N /A-99 6.81 x 10' I'/A 99m 6.27x 10I' N/A.103 5.18 x 10-2 N/A-106 1.27 x 10-2 N/A-103m 5.18 x 10-2 N/A-106 1.27x 10-2 N/A.1 lom 1.27 x 10' N/A*125m 2.54 x IO-2 N/A 127m 2.63 x IO1' N/A 127 1.09 x 10° N/A 129m 1.73 x 10° N/A 129 1.64 x 10 NI/A 131m 2.36 x 100 /4 131 1.09 x 100 tJ/4 /132 2.63x 10' _4/Amendment 97 February 1, 2001)

Attachment 6 to TXX-05127 Page 72 of 89 CPSES/FSAR TABLE 15.7-3 (Sheet 3)MAXIMUM RADIOACTIVITY IN A FLOOR DRAIN TANK FOR ACCIDENT ANALYSIS 5ozo- tjr. lAk Ac'tdi tope Floor Drain Tank UsL to Dtle r&ine, 1le e Si Artivitv (Cil l n 1 0ontrvtocAj 1 esCout~ces Iso Te-134 Ba-140 La-140 Ce-141 Ce-143 Ce-144 Pr-143 Pr-144 2.73 x 10°3.81 x 10 1.27 x 10"-5.72 x 10.2 4.54 x 102 3.54 x 10-2 5.72 x 10.2 3.54 x 10-2 t/A Assumptions

a. 3565 MWt b. Based on I percent failed fuel c. Floor drain tank of 30,000 gal capacity assumed 80 percent full of reactor coolant Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 73 of 89 CPSES/FSAR TABLE 15.7-4 (Sheet I of 2)PARAMETERS FOR POSTULATED RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE 1. Data and assumptions used to estimate radioactive source from postulated accident a. Power level (MWt)b. Percent of fuel defected (%)c. Release of activity by nuclide 2. Data and assumptions used to estimate activity release 3565 Table 15.7-3 a. Entire content of tank is assumed to be released to the building due to failure Activity release to the atmosphere takes place over a two hour period at ground level.3. Dispersion data a. EAB distance b. x/Q 4. Dose data a. Method of dose calculations

b. Doses thyroid dose =b .uiatiu 11 ..... ,1 I ... _. 11.._MN to tile X~i 61 alsX c. Iodine Partition factor is 0. 1 2080 m 1.6 x 104 seC/M3 (0-2 hr)See Appendix 15B~LEAB M2 93.1 rem Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 74 of 89 CPSES/FSAR TABLE 15.7-4 (Sheet 2)PARAMETERS FOR POSTULATED RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE whole body dose =l0.0038 14 A 18-l rem (gamma dose)EAIA that E," .,,.,,La tW ,,' fSAWl w^a v u,11EuuS.

Radialiuii Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 75 of 89 CPSES/FSAR TABLE 15.7-6 1from Damaged Fuel Rods Noble Gas and Iodine Activities Released tv th, iL:. A& mnIIXt As a Result of a Fuel Handling Accident'Nuclide Curies Released Kr-85 414 +02 Xe-131m 5t4A 10 3+02 Xe-133m 5.56e+03 Xe-133 .GA1 69e0lXe-135m G.G 1 l.6e-1 Xe-135 .x 0 2e2lI-131 I-132 i. AO8+1-133 G.31 A+03 I-135 +79 ,O0O 1. Based on the following assumptions:

a. The total gap inventory of one fuel assembly in the discharge region is released to the spent fuel pool.b. A radial peaking factor of 1.65 is applied.c. The accident occurs 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown.d. 8.p, LkiitietIMSOldW to f896... OWUh Col ifvlftU C'C103 l.WX fsvss WAM rI I8EI 1VII1 ;>il .JUS V1 6 'F il WIvl rh"=ftrVSy.

Gap activities are 8% for 1-131 and 5% for all other iodines and are 10% for Kr-85 and 5% for all other noble gases.Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 76 of 89 CPSES/FSAR TABLE 15.7-7 (Sheet 1 of 2)PARAMETERS FOR POSTULATED FUEL HANDLING ACCIDENT ANALYSIS 1. Data and assumptions used to estimate radioactive source from postulated accidents a. Power level (MWt) 3565 b. Release of activity Regulatory Guide6 c. Damage to fuel assembly All rods ruptured d. Iodine fractions (inorganic and Regulatory GuideT!: organic)2. Data and assumptions used to estimate activity released a. Spent fuel pool 10 decontamination factor b. All other pertinent data and Regulatory Guidej assumptions

3. Dispersion data a. EAB and LPZ distances 2080 m and 4 miles b. x/Q 1.6 x 10-4 sec/m 3 0 5 sec/rn 3 A Tlee ts 2Q 23 D (0-2hr)(0 -8 hr)a.b.C.Method of dose calculation Dose conversion assumptions Doses See Appendix 15B See Appendix 1 5B@EAB, Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 77 of 89 CPSES/FSAR TABLE 15.7-7 (Sheet 2)PARAMETERS FOR POSTULATED FUEL HANDLING ACCIDENT ANALYSIS thyroid dose = 69 rem, 22.0 whole body dose = 0-44 rem 0.087 (gamma dose)@ LPZ thyroid dose = .rem, l 3.3 whole body dose = 6.29 x 19-2 rem 0.013 (gamma dose)Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 78 of 89 CPSESJFSAR

9. Locked Reactor Coolant Pump Rotor.APPENDIX 15B DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15B. 1 INTRODUCTION This section identifies the models used to calculate offsite radiological doses that would result from releases of radioactivity due to various postulated accidents.

The postulated accidents are: 1. Steam Line Break 2. Rod Ejection Accident 3. Steam Generator Tube Failure 4. Loss-of-Coolant Accident 5. Radioactive Gas Waste System Leak or Failure 6. Radioactive Liquid Waste System Leak or Failure 7. Design Basis Fuel Handling Accident 8. Primary Coolant Small Line Break Outside Containment 15B.2 ASSUMPTIONS The following assumptions are basic to the model for the gamma and beta dose due to immersion in a cloud of radioactivity and the model for the thyroid dose due to inhalation of radioactivity.

1. The dose contribution of direct radiation from sources other than the leakage cloud is negligible compared to the dose due to immersion in the leakage cloud.2. All radioactivity releases are treated as ground level releases regardless of the point of discharge.

3. The dose receptor is a standard man as defined by the International Commission on Radiological Protection (ICRP)[l].

4. Radioactive decay from the point of release to the dose receptor is neglected.

5. Isotopic beta and ,Mal*1 IM l uv1a.3 It( fiuJ11i LULL [] allu fieili 1Lf .[l7. whole body and beta skin dose conversion factors are taken from references

[13]and [14] respectivily.

6. Atmospheric dispersion lactors used in Ite analyses are presented in Section z.i.15B-1 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 79 of 89 CPSESJFSAR 15B.3 GAMMA DOSE The gamma dose is obtained by considering the dose receptor to be immersed in a radioactive cloud which is infinite in all directions above the ground plane, i.e., a semi-infinite cloud. The concentration of radioactive material within this cloud is taken to be uniform and equal to the maximum centerline ground level concentration that would exist in the cloud at the appropriate distance from the point of release.The whole body dose is a result of exposure to external gamma radiation.

The whole body dose due to immersion in a semi-infinite cloud is given by fej?eF = e'25- t/Q -tAR, -R (1513-1)w INSERT EQUATION 15B-1 AND DEFINITIONS FROM THE FOLLOWING PAGE where:'Dy is the whole body gamma dose from immersion in a semi-infinite cloud for a given time period, rem.ARN is the activity of isotope i released during a given time period, curies X/Q is the atmospheric dilution factor for a given time period, sec/m3 Tlh bgl..iA. W..' -lsle il. ftb iLD 1i ;1 liade t1 1.. iE lays wl d a....;latio rE"ml""i ;f a li.,y IiJ.U it ~ill tile I1vi1IubiiLz t~ L'.15B.4 THYROID INHALATION DOSE The thyroid inhalation dose is obtained from the following expression E+]-B -, iQI ~ Drs(1513-2)

INSERT EQUATION 15B-2 AND DEFINITIONS FROM FOLLOWING PAGE where: DyHy = thyroid inhalation dose, rem X/Q = atmospheric dilution factor for a given time period, sec/M 3 B = breathing rate for a given time period t, m 3/sec Qj = total activity of iodine isotope i released in time period t, curies (DCF)i = dose conversion factor for iodine isotope i, inhaled 15B-2 Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 80 of 89 Equations for page 1513-2 Replace Equation (15B-1)Dy = K

• X/Q ARi

• DCF 7 Y Insert K is the conversion factor 3.71312 rem-Bq/Ci-Sv DCFf, is the whole body gamma dose conversion factor for isotope I, Sv-m 3/Bq-sec Replace Equation (15B-2)DTHY = K

• X/Q

• B

• E Qi * (DCF), Insert K the conversion factor 3.7E12 rem-Bq/Ci-Sv Attachment 6 to TXX-05127 Page 81 of 89 CPSES/FSAR The thyroid inhalation dose conversion factors and offsite breathing rates used in the above model are given in Table 15B-1.15B.5 CONTROL ROOM DOSE The thyroid inhalation, whole body gamma, and beta skin dose models for the major contributors to the Control Room dose are described below. The dose to the Control Room occupants due to a postulated accident is calculated on the basis of source strength, atmospheric transport, dosimetry and Control Room emergency pressurization and filtration considerations as illustrated in the following equations fflj The thyroid inhalation dose is obtained from the following expression:

Tj (Il 5B-3)a"" =Z f3R B Ant ro Ij\L; j T .INSERT EQUATION 15B-3 AND DEFINITIONS FROM FOLLOWING PAGE where: i isotope index j = time interval index BR = breathing rate, m 3/sec DCF; = dose conversion factor for iodine isotope i, inhaled Ai(t) = airborne concentration of iodine isotope i at time t (sec), in the Control Room, curies/m 3 The thyroid inhalation dose conversion factors and Control Room breathing rate used in the above model are presented in Table 15B-1.The whole body gamma dose due to inleakage is calculated using the following equation: E)h = (Rin P0 A~ (PER (I 5B-4)INSERT EQUATION 15B-4 AND DEFINITIONS FROM FOLLOWING PAGE where: DR = Whole body gamma dose, rem el iadiuO ufanl Ljuivakill 1 1 il;ia L Cul iuunk, in A; = time integrated concentration of nuclide i, i h" I Ci-seclm3 (K* d03 e~"e i feet" 1 f~. !"Ilde i, Rel II'e a defillet b, e~ili 15B-3 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 82 of 89 Equations for pagce 1511-3 Replace Equation (15B-3)Tj DTHY=KZ- BR-DCF;- AjA(t)dt ij Ti-I Insert K = the conversion factor 3.71312 rem-Bq/Ci-Sv Replace Equation (15B14)DR = K

• I/GF

• E Ai * (DCF),, Insert K = the conversion factor 3.7E12 rem-Bq/Ci-Sv GF = Geometry Factor to adjust the dose to reflect the finite cloud in the control room DCFr, = the whole body gamma dose conversion factor for isotope i, Sv-m 3/Bq-sec Attachment 6 to TXX-05127 Page 83 of 89 CPSES/FSAR INSERT EQUATION 1 5B-5 AND DEFINITIONS FROM FOLLOWING PAGE<-i @ _ w _1 06 .A %(IS B1 -5)I-brkiu122 ell'elliY eiitteP. disilyte61U1t fol f 1 ulide. i; 2tl~le-aEg

,-flU &t 1- ~1 b""" A~b vee V4 isa C- -lv lva toh wulse e ovetlsiv fneaelfl 6811112 ea .............

e.... iwg, EJ, Zhole bd I Thelgamma dose conversion factors ffrzuY ban rft 1 1 1"?t eil. gies EI:, are presented in Table 15.B-1 The beta skin dose due to inleakage is calculated using the following equation f B3B =+t 6 EB; (15B-6)INSERT EQUATION 15B-6 AND DEFINITIONS FROM FOLLOWING PAGE where: DB = beta skin dose from immersion in a semi-infinite cloud, rem Ai = time integrated concentration of nuclide i, Ci-sec/m3 Er. a -eAige btau eneigy eni 1;tLJ by Li top ;, MeNv is'The beta eie ina een i ei e... iF ll., a. lu..;.ein

.il.Weskin dose conversion factors wsw;V~ssb0.;.AtL41 M~xids IM. C011pte "VvIUwa 011Csiv thidHC12Iw fflvssufll vvtc talelry fol a b"'-" jJLwAftAULL.

TlHe vc a f are presented in Table 15.B-l.The whole body gamma dose to Control Room personnel due to a cloud external to the Control Room is calculated using the following equation: Dc =*X/Qj* Aj; *CF) (I 5B-7)where: DC whole body gamma dose due to external cloud shine, rem X/Qj = atmospheric dispersion factor for the time period j, calculated at the Control Room air intake assuming a ground level release, sec/m3 Aij = total activity of nuclide i released during time period j, Ci 15B-4 Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 84 of 89 Equations for page 15B14 Replace Equation (15B-5)GF= 1 173NV 0 3 3 8 Insert V = control room volume, ft 3 Replace Equation (15B-6)DB=K*ZA; * (DCF)Bj Insert K = the conversion factor 3.71312 rem-Bq/Ci-Sv DCFBi = the beta skin dose conversion factor for isotope i, Sv-m3/Bq-sec Attachment 6 to TXX-05127 Page 85 of 89 CPSES/FSAR ANISND solves the one-dimensional Boltzmann transport equation for neutrons or gamma rays in rectangular, spherical or cylindrical geometry.

The source may be fixed, fission, or a subcritical combination of the twvo. Criticality search may be performed on any one of several parameters.

Cross sections may be weighted using the space and energy dependent flux generated in solving the transport equation.

The external cloud dose calculations use Bugle 80, a P-3 cross section library.5. EFFLMDA [10]EFFLMDA calculates the effective spray removal constant as a function of time.The calculation is based on a multi- region model with intermixing among the sprayed and unsprayed regions in a reactor containment.

The EFFLMDA results are merged with DRAGON input data. This enables the DRAGON program, a single spray region computer code, to simulate a multi-region spray model with step changes in the effective spray constants.

6. TITAki5 v -- RADTRAD [12]Program -lAIr pulforms similar analysis functions as described above for i DRAGON. i REFERENCES I. ICRP Publication 2, Report of Committee II on Permissible Dose for Internal Radiation (1959). The International Commission on Radiological Protection.

2. Tfame u t~uupes, zsixt ditionl, bye.It:I.

L J.lM. IIullad11., I. PM ..liL,. Not Used.3. R..iul. .Uiy C eAif...1, ABerkle;Vk VmU.,J Radia tV.LI f taskbu t;vla R*tJlVagIeal

_VeUseene V- i EVEN OI C.aehtt AeeIUent forA

• v.e.4 t... B 4MeNot Used.Reacturs, Re visionv -2 Ju.sw 1974, b.3. biucien, Rve~umat...

euznmisaossi..

4.4 D TE T _ ' F.D An t sn-TE. Bae an R.n Waserfd Di v#-o 1 f*.: ..5 , -.... ai-U, a .\...>.d e ,. S ..., ^ ^^ .N .6, .L ... Nt U sed.ZiI 5 n\^ *. .-B ?64R.tO hild, B.E Mani, Mac .5, 6. RADIOISOTOPE, SWEC proprietary computer code NU-007, Version 01, Le 02, April, 1985.7. DRAGON, SWEC proprietary computer code NU-1 15, Version 04, Level 02, September 1984 and Version 05 Level 00, April, 1986.8. QADMOD, SWEC proprietary computer code NU-137U, Version 00, Level 0 November 1985.9. ANISND, SWEC proprietary computer code NU-146, Version 01, Level 01, January 1985.INot Used. I vel 3, 1513-6 Amendment 98 Attachment 6 to TXX-05127 Page 86 of 89 CPSES/FSAR

10. EFFLMDA, SWEC proprietary computer code NU-159, Version 01, Level 00, December, 1984.11. International Commission on Radiological Protection, "Limits for Intakes of Radionuclides by Workers," ICRP Publication 30, Volume 3 No. 1-4, 1979. [+1 f-iWatingEto asepot ie-m eepie ee *eo 4. .*, Beot

• a .r i 995x* 3.ssV~ VIs}WAPLXWV-W JU=.sV;UXI NUREG/CR-6604, "RADTRAD:

A Simplified Model for RADionuclide Transport and Removal And Dose Estimation," December 1997.NUREG/CR-6604, Supplement 1, "RADTRAD:

A Simplified Model for RADionuclide Transport and Removal And Dose Estimation," June 1999.NUREG/CR-6604, Supplement 2, "RADTRAD:

A Simplified Model for RADionuclide Transport and Removal And Dose Estimation,'

October 2002.K. F. Eckerman and J.C. Ryman, External Exposure to Radionuclides in Air, Water, and Soil," Federal Guidance Report No. 12, EPA 402-R-93-081, Environmental Protection Agency, September 1993.DOE/EH-0070, "Extemal Dose-Rate Conversion Factors for Calculation of Dose to the Public," July 1988.I NRC Regulatory Issue Summary 2001 -19: Deficiencies in the Documentation of Design Basis Radiological Analyses Submitted in Conjunction with License Amendment Requests," 10/18/01.1513-7 Amendment 98 Attachment 6 to TXX-05 127 Page 87 of 89 CPSES/FSAR TABLE 15B-1 (Sheet I of 2)PHYSICAL DATA FOR DOSE 1122 IC 1 -0;..f-r'4*ro" KT -W ,*i1 so E' 'ge Bet" Ellogy}PffeWdiRt-rrv

-+5 0.101 1..00 O."+J.II _.IE.7rr7 e+s6 aw.221 V.VJV eW7T 2e~0.022^ "91 DCr.JULV M ., O' .JUll F lU V I. J f .;sxi6 5.9 w-\t 41 4-Uex I W 2.4fb W 40 102+ e e@ A e t v DCFY* s: t t *~ ._ t _ _ g ..Fiee .-3 fbi .f! fo. Ru*,*m *.E 10 1i-51 U U a U I 0.01G 0.02J;it v0.>q v U U Amendment 97 February 1, 2001 Attachment 6 to TXX-05 127 Page 88 of 89 CPSES/FSAR TABLE 15B-1 (Sheet 2)PHYSICAL DATA FOR DOSE CONVERSIONS Delete and Replace with following Table 1 5B-1---.I UVtLJL.t+/-*"4-*M~*ec 134 geC;-3, K"1 l30 A9 v / IJJ If*, Bet E0.e0g P 1fe;.iq2f7 0-65 0.110 o.WAn o-%a-1 11l I VlIVl FmJm 6-e

• i4v l ' 'DCF I LIJIVJJV4 X...UVI4 6rem_ W+- rH _ 1-tzval~sB ..io__J. 1 VJI WI JW '-WAIVWIJVII L UffVM 4-9 BD<lEA+Hi~ib icr+F E Finles perie4l&-24 2 4 720 3.47, 10-_Mr 73'U10 44-*x 4O" I., -...I.r ere e fil GfleUfat t43 "', eqn ;V. -5 6-2 * ^* se as E.ehe -M11 bL.C lr! .20 *,1 I Ft., r." *Amendment 97 February 1, 2001 Attachment 6 to TXX-05127 Page 89 of 89 Table 15B-1 Isotope 1-131 1-132 1-133 1-134 1-135 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Xe-131m Xe-133m Xe-1 33 Xe-135m Xe-135 Xe-1 38 Dose Conversion Factors Thyroid DCF(') Whole Body DCF(2)(Sv/Bq) (Sv-m 3/Bq-sec)2.9E-7 1.82E-14 1.7E-9 1.12E-13 4.9E-8 2.94E-14 2.9E-10 1.30E-13 8.5E-9 7.98E-14 N/A 1.50E-18 N/A 7.48E-15 N/A 1.19E-16 N/A 4.12E-14 N/A 1.02E-13 N/A 3.89E-16 N/A 1.37E-15 N/A 1.56E-15 N/A 2.04E-14 N/A 1.19E-14 N/A 5.77E-14 Beta-Skin DCF(3)(Sv-m 3/Bq-sec)8.66E-15 3.02E-14 2.44E-14 3.87E-14 2.16E-14 0.OOE+00 1.38E-14 1.35E-14 9.08E-14 2.13E-14 4.07E-1 5 8.46E-1 5 2.85E-1 5 5.85E-15 1.75E-14 3.99E-1 4 Time Period (hours)0-8 8 -24 24 -720 Breathing Offsite 3.5E-4 1.8E-4 2.3E-4 Control Room 3.5E-4 3.5E-4 3.5E-4 Rates (m 3/sec)(1) Thyroid DCFs from ICRP-30 (Reference i1)(2) Whole-body DCFs from Federal Guidance Report 12 (Reference 13)(3) Beta-skin DCFs from DOE/EH-0070 (Reference 14). Skin doses are modeled without the contribution from photon emissions as supported by the NRC in Reference 15.