ML040850600

Supplement to the Request for License Amendments Related to Application of Alternative Source Term

ML040850600
Person / Time
Site:	Peach Bottom
Issue date:	03/15/2004
From:	Gallagher M Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML040850600 (67)

Text

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Exelon Nuclear 200 Exelon Way www.exelonCoTp.Com1 Nuclear Kennett Square, PA 19348 10 CFR 50.90 March 15, 2004 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 & 3 Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

Supplement to the Request for License Amendments Related to Application of Alternative Source Term

References:

(1) Letter from M. P. Gallagher (Exelon Generation Company, LLC) to US NRC, dated July 14, 2003 (2) Letter from G. F. Wunder (U. S. Nuclear Regulatory Commission) to J.

L. Skolds (Exelon Generation Company, LLC), dated January 16, 2004 This letter is being sent to supplement the License Amendment Request (LAR) to support application of an alternative source term (AST) methodology (Reference 1)at Peach Bottom Atomic Power Station (PBAPS), Units 2 & 3. This LAR proposed certain TS and TS Bases changes for PBAPS Units 2 & 3 as part of implementing an AST methodology.

In the Reference 2 letter, the U. S. Nuclear Regulatory Commission requested additional information. Attachment 1 to this supplemental letter provides a partial response to the request for additional information. The remaining questions will be subsequently answered with a future letter. Attachment 2 to this supplemental letter provides the revised TS Bases Markup Inserts. to this supplemental letter provides the camera-ready TS Bases pages which supercede the corresponding Bases pages previously submitted in Reference 1.

There is no impact to the No Significant Hazards Consideration submitted in the Reference 1 letter. There are no additional commitments contained within this letter.

40 Z)

Supplement to the Request for License Amendments Related to Application of Alternative Source Term March 15, 2004 Page 2 If you have any questions or require additional information, please contact Doug Walker at (610) 765-5726.

I declare under penalty of perjury that the foregoing is true and correct.

Respectfully, Executed on 63(Sf C Michael P. Gallagher Director, Licensing and Regulatory Affairs Attachments: 1. Exelon Response to the Request for Additional Information

2. Revised TS Bases Markup Inserts pages

3. Camera-ready TS Bases pages cc: H. J. Miller, Administrator, Region I, USNRC C. Smith, USNRC Senior Resident Inspector, PBAPS G. Wunder, Senior Project Manager, USNRC (by FedEx)

R. R. Janati - Commonwealth of Pennsylvania

ATTACHMENT 1 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 Supplement to License Amendment Request for "PBAPS Alternative Source Term Implementation",

Response to Request for Additional Information

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 1 of 18 REQUEST FOR ADDITIONAL INFORMATION PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 PROPOSED USE OF ALTERNATIVE SOURCE TERM (AST) METHODOLOGY It is proposed that pages B 3.8-40, B 3.8-74, and B 3.8-94 of the Technical Specifications (TS) bases be changed to indicate that ac and dc electrical power are only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e.,

fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

Pages B 3.8-40, B 3.8-74, and B 3.8-94, however, also indicate that ac and dc are required to ensure (a) the facility can be maintained in the shutdown or refueling condition for extended periods, (b) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status, and (c) adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel. Please clarify your requirements for ac and dc power.

RESPONSE: The TS Bases inserts "Q", "S", "r, for B 3.8-40, B 3.8-74, and B 3.8-94 respectively, have been revised to avoid possible misinterpretation. A clarifying statement has been added to indicate that the inserted statement regarding electrical power requirements only applies to the movement of irradiated fuel. The insert revision is shown below:

INSERT 0:

involving recently irradiated fuel. With respect to moving irradiated fuel assemblies, AC electrical power is only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

INSERT S:

With respect to moving irradiated fuel assemblies, AC and DC electrical power are only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

INSERT T:

With respect to moving irradiated fuel assemblies, DC electrical power is only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e.,

fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

2. TS 3.8.2, 3.8.5, and 3.8.8 (which are currently applicable in Modes 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment) require, in part, immediate suspension of movement of irradiated fuel in secondary containment when both offsite preferred power sources, redundant safety-related electric onsite power sources, or redundant safety-related distribution systems are no longer operable. The proposed TSs relax these TS requirements such that TSs 3.8.2, 3.8.5, and 3.8.8 will be applicable when in Modes 4 and 5 and during movement of recently irradiated fuel. The proposed change thus allows, without TS restrictions, the movement of irradiated fuel assemblies that have decayed at least 24

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 2 of 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> when there is no offsite power, when there is no onsite power, or when there is no ac and dc electric power through the electric distribution system to safety system loads.

a. The application for amendment indicates that the AST analyses take credit for standby liquid control (SLC) system operation. The amendment application also indicates that the SLC system is safety-related, is required to be operable by TSs, and is supplied with emergency power. Justify movement of irradiated fuel assemblies that have decayed at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without the availability of the SLC safety system credited In the AST analyses.

RESPONSE: The SLC system is not credited in any way for the fuel handling accident. The new SLC safety function of pH control only applies to LOCA and ECCS liquid in containment. No pH control is required, assumed, or credited for fuel handling accidents in accordance with the guidance of Reg Guide 1.183.

Therefore, movement of irradiated fuel can be justified without SLC system availability.

b. Justify movement of irradiated fuel assemblies that have decayed at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without the availability of safety systems such as those needed to maintain plant shutdown, for monitoring and maintaining the unit status, or to mitigate events postulated during shutdown.

RESPONSE: The justification to allow movement of irradiated fuel assemblies that have decayed without the availability of those mentioned safety systems is supported by and contained in TSTF-51 Standard Technical Specification Change Traveler. From the TSTF-51:

The purpose of the TSTF is to "remove the TS requirements for ESF features (e.g., primary/secondary containment, standby gas treatment, isolation capability) to be Operable after sufficient radioactive decay has occurred to ensure off-site doses remain below the SRP limits (a small fraction of 10CFR100).

Following reactor shutdown, decay of the short-lived fission products greatly reduces the fission product inventory present in irradiated fuel. The proposed changes are based on performing analyses assuming a longer decay period to take advantage of the reduced radionuclide inventory available for release in the event of a fuel handling accident. Following sufficient decay occurring, the primary success path for mitigating the fuel handling accident no longer includes the functioning of the active containment systems. Therefore, the Operability requirements of the TS are modified to reflect that water level and decay time are the primary success path of mitigating a fuel handling accident (which meets Criterion 3).

Also, the Tech Specs only allow the handling of irradiated fuel in the reactor vessel with the water level in the reactor cavity at the high water level.

Therefore, the proposed changes only affect containment requirements during periods of relatively low shutdown risk during refueling outages.

Therefore, the proposed changes do not significantly increase the shutdown risk.'

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 3 of 18

3. You describe the use of the SLC system to buffer the suppression pool following a loss-of-coolant accident (LOCA) involving significant fission product release. The buffering action of the sodium pentaborate from this system was analyzed to demonstrate that the suppression pool pH remains above 7 for the 30-day LOCA duration. Please provide a more detailed description of this analysis and its assumptions. This description should include the following:

a. Generation of hydrochloric acid by decomposition of the chlorine bearing cables, RESPONSE: The calculation of suppression pool pH uses the same methodology as used and approved for Grand Gulf and addresses the same acid production mechanisms. The simulation tools used for PBAPS were benchmarked against the Grand Gulf published results before use with Peach Bottom parameters. The PBAPS Design Analysis is described in Appendix 6 of this response.

b. Production of nitric acid in the post-accident radiation field, and RESPONSE: See response to part a. above. Nitric acid production is described in Appendix 6 of this response.

c. Determination of the amount of sodium pentaborate required for maintaining the suppression pH below 7 (also listing the input and output to the computer code used in the analysis).

RESPONSE: See response to part a. above. Appendix 6 of this response describes information used to determine suppression pool pH.

4. A review of the ARCON96 meteorological data input files for both Tower 1A and Tower 2 reveals that the 1984 wind speed and direction data are reported to the nearest mph and nearest 5 degrees, respectively, whereas the wind speed and wind direction data for the remaining period (1985-1988) are reported to the nearest 10t mph and nearest degree, respectively. Please explain the data recording and processing procedures that resulted in reduced precision of the reported 1984 data as compared to the 1985-1988 data.

RESPONSE: The upgrade to the meteorological monitoring system was completed in May 1983. For approximately the next year, strip charts were used to record the data in parallel with the break in of the new system. The accuracy of reading these charts was such that 1/10h of a mile per hour could not be readily seen. In the 1984/1985 timeframe, a new digital data logger was purchased and placed into service following the system test service, thereby affording better accuracy.

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 4 of 18

5. A review of the ARCON96 meteorological input files for Tower 1A also shows an unusually high occurrence of low wind speeds (less than 0.5 m/sec) during 1984 as compared to 1985-1988:

Tower 1A Wind Speed Frequency Distributions Tower Period of Wind Speed Range (m/sec)

Level Record <0.5 0.5 -1.0 1.0-1.5 1.5-2.0 2.0-3.0 3.0-5.0 5.0-10.0 >10.0 34-ft 1984 21.7% 19.4% 14.0% 13.9% 14.0% 15.1% 1.9% 0.0%

1985-1988 1.8% 20.8% 20.8% 17.9% 20.5% 15.2% 3.2% 0.0%

92-ft 1984 9.0% 16.7% 11.7% 13.9% 22.0% 21.7% 5.1 % 0.0%

_____1985-1988 1.5% 16.0% 15.4% 13.2% 24.3% 22.5% 7.2% 0.0%

Please explain what might have caused these differences in reported wind speed frequency distributions between the 1984 data set and the 1985-1988 data set.

RESPONSE: The upgrade to the meteorological monitoring system was completed in May 1983. Prior to this, the older Climatronics wind cups would report a wind speed of 0.5 mph if there were no motion of the cups. The use of the strip charts to record data also compounded this effect. Inclusion of the 1984 wind data with more frequent low wind speeds adds conservatism to the calculation.

6. The LOCA analysis assumes that control room isolation and the main control room emergency ventilation system have been initiated by the start of gap release. During the isolation mode, unfiltered inleakage into the control room is assumed to be 1,600 cfm.

This inleakage of unfiltered air, which can occur through doorways, envelope penetrations, and leakage in ventilation system components, was modeled using the control room intake X/Q values. Verify that there are no other potential unfiltered inleakage pathways that could result in X/Q values that are higher than the control room intake X/Q values.

RESPONSE: The unfiltered inleakage allowance is at control room intake equivalent conditions. The control room is maintained at a positive pressure with a once-through ventilation system. Unfiltered inleakage paths are therefore limited to small portions of the ventilation system and control room boundary that are found to be at a negative pressure. The Turbine Building surrounds the control room on 3 lateral sides, as well as above and below, with the remaining side facing the Radwaste building. The infiltration paths will be identified via inleakage testing and/or Control Room Envelope inspections and surveillances similar to that described in Table H-1, Determination of Vulnerability Susceptibility,' of NEI 99-03, "Control Room Habitability Guidance'. If control room equivalent X/Qs are found to be non-conservative, an adjusted effective inleakage evaluation will be performed to ensure the control room remains habitable under all conditions.

The unfiltered inleakage is not a derived value. Rather, it is an allowance that has been determined to be acceptable for control room personnel protection. Unfiltered inleakage will be measured and controlled as described in the Exelon response to NRC Generic Letter 2003-01, "Control Room Habitability' dated 12/9/03.

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 5 of 18

7. Explain the basis for the 131.4 meter release height used in the PAVAN computer runs for main stack releases to the control room.

RESPONSE: PAVAN was used for releases from the main stack to the control room as suggested in Regulatory Guide 1.194, Section 3.2.2. For conservatism, stack grade (85.4 m) was assumed to be equal to plant grade (35.4 m). The effective release height was then determined in accordance with RG-1.194, which states, 'the input parameters should be adjusted such that the effective release height is measured from the elevation of the control room outside air intake rather than plant grade." The stack top is 152.4 m above its assumed plant grade elevation and the control room intake is 21.0 m above plant grade elevation, which results in the following calculation:

Stack height - intake height = effective release height 152.4 m - 21.0 m = 131.4 m Therefore, 131.4 m is the release height that was utilized in the PAVAN analysis.

8. Both Regulatory Guides (RG) 1.145 (Section 5.3) and 1.194 (Section 2) imply that the period with the most adverse release of radioactive materials to the environment should be assumed to occur coincident with the period of most unfavorable atmospheric dispersion. For the main stack releases, the highest control room X/Q values are associated with 0-2 hour flow reversal conditions and the highest offsite X/Q values are associated with the 0-0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> fumigation conditions. Please describe how these highest X/Q values were used coincident with the most limiting portion of the release to the environment to estimate control room and offsite doses.

RESPONSE: Response to this question will be provided in a future supplement.

9. Explain in more detail the methodology used to model steam cloud transport for the main steamline break (MSLB) accident. Please also provide the resulting control room and offsite X/Q values.

RESPONSE: The methodology used to assess the radiological consequences of a Main Steam Line Break (MSLB) accident is the same as that used by Exelon in previous MSLB accident assessments. This approach has been used for Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2, and Clinton AST submittals and is currently being reviewed by the NRC.

Appendix 7 to this response contains additional information regarding the MSLB accident dose consequence analysis.

The postulated MSLB accident assumes a double-ended break of one main steam line outside the primary containment with displacement of the pipe ends that permits maximum blowdown rates. The break mass released includes the amount of steam in the steam line and connecting lines at the time of the break, plus the amount of steam that passes through the valves prior to closure.

The analysis assumes the MSLB accident to be an instantaneous ground level release.

Two models are considered for assessment of MSLB accident radiological consequences. One is for assessing control room dose and the other is for assessing offsite consequences.

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 6 of 18 In the control room model, the released reactor coolant and steam at operating temperature and pressure is conservatively assumed to expand to a hemispheric volume at atmospheric pressure and temperature. No credit is taken for dilution of the steam cloud by the air into which the steam is ejected. Neither the Turbine Building structure nor its ventilation system is assumed to have an effect on the cloud resulting from the MSLB accident. This hemisphere is then assumed to move at a speed of 1 meter per second downwind past the control room intake. No credit is taken for buoyant rise of the steam cloud or for decay in transit. Dilution (i.e., dispersion) of the activity in the plume in transit was also conservatively ignored.

For offsite locations, the buoyant rise of the steam cloud is similarly ignored, and the ground level dispersion is based on the conservative and simplified Regulatory Guide (RG) 1.5, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors," methodology. The "instantaneous release" of the MSLB accident is converted to an equivalent curie release. Since no credit is taken for decay over this release time, or in transit, the calculation accurately models an instantaneous release.

In summary, the following assumptions were used in the control room and offsite dose evaluations for a MSLB accident.

• The release from the break to the environment is assumed to be instantaneous. No holdup in the Turbine Building or dilution by mixing with Turbine Building air volume is credited.

• The steam cloud is assumed to consist solely of the initial steam blowdown and that portion of the liquid reactor coolant release that flashed to steam.

• The released reactor coolant and steam is assumed to expand to a hemispheric volume at atmospheric pressure and temperature consistent with an assumption of no Turbine Building credit.

• This hemisphere is then assumed to move at a speed of 1 meter per second downwind past the control room intake.

• No credit is taken for buoyant rise of the steam cloud or for decay, and dispersion of the activity of the plume was conservatively ignored.

• For offsite locations, the buoyant rise of the steam cloud is similarly ignored, and the ground level dispersion is based on the conservative and simplified Regulatory Guide 1.5 methodology.

The specific information requested is provided below:

a. A sensitivity analysis was not performed since the model maximizes the available concentration and thus yields the highest control room intake. The modeling did not consider heat losses in the expansion.

b. The normal operating reactor vessel steam dome pressure of 1050 psia and corresponding saturation temperature of 550.530 F were used as the temperature and pressure of the steam prior to expansion to atmospheric temperature and pressure.

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 7 of 18

c. Atmospheric pressure and temperature were assumed to be 14.7 psia and its associated saturation temperature.

The dose assessment modeling conservatively assumed that an unprotected individual is located at the control room intake location for the duration of the accident (i.e., during the time of cloud passage). No protection is provided by the control room envelope, except in terms of the geometry factor for external exposure.

10. The proposed change to the Updated Final Safety Analysis Report (UFSAR) (Section 8 of the License Amendment Request (LAR)) states that the temperature profile presented in UFSAR Figure 14.6.12A includes a 2G adder for decay heat. The figure is identified in the UFSAR as Revision 15 dated April 1998. During an amendment review in 2000, it was stated that the figure did not include this adder and additional information was provided to the staff to justify adequate conservatism in the minimum containment pressure available (MCPA) calculation without the adder at that time (Letter from B. C.

Buckley, Sr., Nuclear Regulatory Commission (NRC), to J. A. Hutton, PECO Energy Company, August 14, 2000, "Peach Bottom Atomic Power Station, Units 2 and 3 Issuance of Amendment Regarding Crediting of Containment Overpressure for Net Positive Suction Head Calculations For Emergency Core Cooling Pumps (TAC Nos.

MA629.1 and MA6292).N). Explain the discrepancy in these two statements. Has the analysis and figure been updated to include the adder and approved for use? If the figure does include this adder then why was this not identified during the amendment review in 2000?

RESPONSE: The analyses provided during the amendment review in 2000 did not explicitly include a 2a adder for decay heat, and additional information was provided to the staff to justify adequate conservatism elsewhere in the MCPA calculation without the adder. Subsequent to that review, it had been Exelon's intent to explicitly incorporate a 2a adder for decay heat in future containment analyses. On May 24, 2001, GE issued SIL-636, and on June 6, 2001, SIL-636 revision 1, informing GE BWR owners of errors in the GE specific implementation of the ANSI/ANS 5.1-1979 decay heat standard. New decay heat values and uncertainties specific to PBAPS were regenerated using corrected GE procedures. Reanalysis of the containment response to a DBA-LOCA was performed by GE using their NRC approved SHEX computer code. This analysis included an explicit 2a adder for decay heat. As described in the amendment review in 2000, it is this vendor containment analysis that was used as the basis for the revised utility generated MCPA calculation referred to in this submittal. With the explicit incorporation of the 2cr adder for decay heat, the additional justification of adequate conservatism in the MCPA calculation without the adder has been deleted.

11. The proposed revised UFSAR text identifies a change in methodology regarding how the containment leakage is addressed in the MCPA analysis.

a. Provide the MCPA and containment overpressure license (COPL) calculation for the NRC staff's review.

b. How is it different from the previously reviewed method described in PECO Energy Company's Calculation PM-I1013, NMinimum Containment Pressure Calculation,"

Revision 3, February 2000?

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 8 of 18

c. How are the main steam isolation valve (MSIV) and airlock leakages included in the calculation?

d. How are the leakages conservatively varied with the containment pressure assuming turbulent flow?

RESPONSE: Response to this question will be provided in a future supplement.

12. Previously, containment leakage was assumed to be constant at La=0.50/o/day throughout the event. The containment leakage has been increased to L8=0.7 0/odday for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, based on the proposed change to TS 5.5.12, for a peak post-accident containment pressure of 49.1 psig. This leakage is then reduced to 0.56xL, 0.392 0/o/day from 24 to 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br /> and then reduced to 0,50xLa=0.350 0/o/day, for 38 to 7'20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. In addition, MSIV leakage of 174 scfh is included (based on the proposed change to TS 3.6.1.3) in the MCPA calculation, with leakage measured at a test pressure of 25 psig.

After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the MSIV leak rate is reduced to 77.2%, then to 65.4% at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, to 59.0% at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, to 55.5% at 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, and finally to 50% at 157 hours0.00182 days <br />0.0436 hours <br />2.595899e-4 weeks <br />5.97385e-5 months <br /> for the remainder of the event. Leakage from the personal airlock of 9,000 sccm, for a peak post-accident containment pressure of 49.1 psig, is also included in the proposed change to the MCPA calculation.

a. How are the leakages conservatively varied with the containment pressure assuming turbulent flow?

b. How does this evaluation differ from the MCPA and COPL calculation in question 11 above, which is only carried out to 12.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />?

c. Identify the TS which controls the allowable airlock leakage rate.

RESPONSE: Response to this question will be provided in a future supplement.

13. During the previous amendment review (Hutton, J. A., PECO Energy Company, to NRC, "Peach Bottom Atomic Power Station, Units 2 and 3 Response to May 10, 2000, Telephone Questions Regarding PECO Energy License Amendment Request Related to Generic Letter 97-04,0 June 29, 2000) it was stated that the margin between the MCPA and COPL was set at 1 foot (0.42 psid), The proposed amendment would decrease this margin to about 0.28 psid.

a. Provide a justification for reducing this agreed to margin.

b. Provide a comparison of the COPL value to the COPR (containment overpressure required) value for the residual heat removal (RHR) and core spray pumps for the most limiting event(s), including the margin to the COPL value before and after the proposed change to the MCPAICOPL calculation.

While not directly related to the MCPA calculation, justification for the inclusion of the suppression chamber air space in the mixing of the radioactive release needs to be provided.

RESPONSE: Response to this question will be provided in a future supplement.

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 9 of 18

14. In addressing RG 1.183, Appendix A, LOCA Item 6.1, it is stated in Table B that the radioactive release is mixed with the suppression chamber air space 'based on expected steam flow from the drywell to the suppression chamber, even after the initial blowdown."

a. Is this based on the results of thermal-hydraulic analyses performed for the duration of the release? If so, provide a summary of the analyses for staff review, or

b. Provide justification for this assumption for the duration of the release.

RESPONSE: Response to this question will be provided in a future supplement.

15. What design-basis parameters, assumptions or methodologies (other than those provided in the July 14, 2003, submittal) were changed in the radiological design-basis accident analyses as a result of the proposed change? If there are many changes it would be helpful to compare and contrast them in a table. Also, please provide a justification for any changes.

RESPONSE: See Appendix 8 to this response, "Differences Between Existing Requirements and the Proposed Change" Justifications for these changes are based on results of dose calculations. The justification for the use of SLC to buffer suppression pool pH will be provided once official RAIs regarding SLC are provided.

16. Based upon a preliminary review of the proposed amendment the reviewer is unable to match the calculated doses for the accident analyses. It would be helpful if the licensee would provide their design-basis accident calculations. If the calculations are provided, answers to questions provided in this request for additional information (RAI) may reference the calculation.

RESPONSE: Refer to the following Appendixes to this response for specific information requested.

Appendix 1: Core Isotopic Activity Appendix 2: General AST Parameters Appendix 3: LOCA AST Parameters (will be provided in a future supplement)

Appendix 4: FHA AST Parameters Appendix 5: FHA Water Coverage Appendix 6: Suppression Pool pH Appendix 7: MSLB Information Appendix 8: Proposed Changes Appendix 9: MSIV Leakage Information (will be provided in a future supplement)

17. Appendix B to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, establishes quality assurance requirements for the design, construction, and operation of those structures, systems, and components (SSCs) that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. Appendix B, Criterion IlIl, 'Design Control," requires that design control measures be provided for verifying or checking the adequacy of a design.

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 10 of 18 Appendix B, Criterion XVI, "Corrective Action," requires measures to be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, defective material and equipment, and nonconformances are promptly identified and corrected. GL 2003-01, "Control Room Habitability," addresses current issues with respect to previously assumed values of unfiltered inleakage. Generally, these issues can only be resolved by inleakage testing. In light of your Appendix B requirements and GL 2003-01, provide sufficient justification to explain why the value assumed for your control rooms unfiltered Inleakage is appropriate for this proposed license amendment.

Provide details regarding your control room, design, maintenance and assessments to justify the use of this number.

RESPONSE: A response to GL 2003-01 was developed and submitted on 1219/03.

This response included a commitment for tracer gas testing.

The control room envelope unfiltered inleakage value used in the analyses is conservative, representing a value that is a significant portion of the powered air flow in the emergency mode of operation. A detailed walkdown of the CR HVAC system was performed between 2/23/04 and 2/27/04. Vendor and station personnel performed this walkdown. No significant sources of inleakage were found during this inspection and assessment that would account for inleakage of the magnitude assumed in the dose consequence analysis for the control room. Therefore, this value is considered to be conservative. An integrated tracer gas test is scheduled as indicated in a supplemental response to GL 2003-01 (to by submitted by March 31, 2004).

18. Regarding the conformance to Section 3.1 of RG 1.183, the NRC staff would like Exelon Nuclear to provide additional details regarding the source term utilized for each accident.

Provide the calculated source term values (nuclide and CVMWt).

RESPONSE: Appendix 1 to this response contains the isotopic core inventory for Peach Bottom. The appendix contains tables indicating source term values in Curies, grams, and Curies/MWt.

19. Questions regarding the use of the SLC are currently being developed and will be provide in a future RAI.

RESPONSE: Response to this question will be provided in a future supplement.

20. On Page 15 of Attachment 1 of the submittal, the second paragraph states that Exelon has used the Brockmann-Bixler model for main steamline deposition. The discussion and the data in Table 5 are insufficient to support an NRC staff confirmation. Please provide the following information.

a. A single-line sketch of the four main steamlines and the isolation valves. Annotate this sketch to identify each of the control volumes assumed by Exelon in the deposition model.

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 11 of 18

b. A tabulation of all of the parameters input into the Brockmann-Bixler model for each control volume shown in the sketch (and time step) for which Exelon is crediting deposition. This includes:

• Flow rate

• Gas pressure

• Gas temperature

• Volume

• Inner surface area

• Total pipe bend angle

c. For each of the bulleted parameters in question 20.b., provide a brief derivation and an explanation of why that assumption is adequately conservative for a design-basis calculation. Address changes in parameters over time, e.g., plant cooldown.

d. Clarify if your analysis addresses a single failure of one of the MSIVs. Such a failure could change the control volume parameters that are input to the deposition model. Previous Implementations of main steam deposition have been found acceptable only if the licensee had modeled a limiting single failure. Please explain why Exelon feels that such a limiting failure need not be considered if it is not considered.

e. Since the crediting of main steamline deposition effectively establishes the main steam piping as a fission product mitigation system, the staff expects the piping to meet the requirements of an ESF system, including seismic and single failure considerations. Please confirm that the main steam piping and isolation valves that establish the control volumes for the modeling of deposition were designed and constructed to maintain integrity in the event of the safe shutdown basis earthquake for Peach Bottom. If the design basis for the piping and components does not include integrity during earthquakes, please provide an explanation of how the Peach Bottom design satisfies the prerequisites of the staff-approved NEDC-31 858P-A, BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems. If piping systems and components at Peach Bottom were previously found by the staff to be seismically rugged using the methodology of this BWROG report, please provide a specific reference to the staff's approval.

RESPONSE: Response to this question will be provided in a future supplement.

21. On page 53 of Attachment 1 of your submittal, you state that your submittal is in compliance with paragraph 6.3 of Appendix A to RG 1.183, and reference the RADTRAD Brockman-Bixler approach apparently as establishing that conformance.

However, paragraph 6.3 of RG 1,183 states that the model should be based on well-mixed volumes, but other models such as slug flow may be used if justified. The Brockman-Bixler model is a slug-flow model. This paragraph did not endorse RADTRAD as an acceptable approach. RG 1.183 states that main steamline deposition will be considered on a case-by-case basis.

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 12 of 18 The staff documented its evaluation of the first application of main steamline deposition credit in an AST in Appendix A of NRC staff report, AEB-98-03, Assessment of the Radiological Consequences for the Perry Pilot Plant Application using the Revised (NUREG-1 465) Source Term.' The methodology of this report, which can be found online in ADAMS at ML011230531, was used by at least two additional licensees.

Generally, when the staff has accepted an application of slug flow, the licensee has (1) committed to maintaining a seismically rugged drain path from the 3rd MSIV to and through the condenser, (2) did not assume deposition in piping upstream of the inboard MSIV, (3) assumed a single failure of the one of the inboard MSIVs, (4) did not credit a delay time in the onset of the release, and (5) assumed a constant pressure and temperature in the steamline over 30 days, The added conservatism from the above assumptions provided additional margins to compensate for differences in conservatism in slug flow and well-mixed assumptions. Please provide a justification for your proposed modeling approach or re-perform the analyses.

RESPONSE: Response to this question will be provided in a future supplement.

22. Page 13 of Attachment 1 of your submittal provides text that states wan initial 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> transport delay is determined. The text suggests that the steamline volume and MSIV leak rate are used to establish this delay. This implies that a delay to fill the steamline is being taken:

a. Your submittal does not identify this as an alternative to the guidance in RG 1.183.

Please explain how this holdup is modeled in the LOCH analysis. Is this modeled as a delay in the onset of the release?

b. Please explain why this delay assumption is consistent with the assumption of slug flow (Item 6.3, Page 53 of Attachment 1).

RESPONSE: Response to this question will be provided in a future supplement.

23. Based on information provided in your submittal, you have assumed an MSIV leakage rate of 0.62 cfm for the 100 scfh lines, and 0.31 cfm for the 50 scfh line, prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-accident and reduced values after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The staff believes that these, values are understated. When the proposed MSIV leakage, in scfh, at test conditions (typically 70 degrees and 25 psig) are scaled to peak containment pressure and temperature (typically 40-50 psig and about 250-350 degrees) the TS leakage past the inboard MSIV has been shown to be 1.3-1.6 cfm, at least double the value you have assumed.

However, the temperature of the fluid in the steamlines is based on the steam piping temperatures, typically 500-600 degrees. At the steam piping conditions, the flow In scfm is even higher, typically 4-8 scfm, Please explain the basis of the values you used and why these values are adequately conservative since the effectiveness of deposition decreases with increasing flow.

RESPONSE: Response to this question will be provided in a future supplement.

24. Paragraph 5.5 of Appendix A of RG 1.183 states that the amount of iodine that becomes airborne should be assumed to be 10%, unless a smaller value can be justified on the actual sump pH history and area ventilation rates. In Figure 2 of Attachment 1 of the amendment request, the ECCS leakage flash fraction is set at 1.41%. On page 52 of

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 13 of 18 Attachment 1 it is stated that the partition factor is based upon ORNL-TM-2412. Please explain why ORNL-TM-2412 is an acceptable alternative to the guidance in RG 1.183.

Is this value in your current licensing basis or is this a new value? If the value is new please provide the details used to calculate this value. Please provide the pH history, and area ventilation rates in the areas of ECCS leakage.

RESPONSE: The ECCS system leakage path is a new release pathway for PBAPS.

The ECCS system leakage flash fraction is derived using methodology historically accepted at the Clinton Power Station (CPS) for ECCS leakage (The use of a flash fraction of 1.36% was previously reviewed and found to be acceptable by the NRC in CPS Amendment 127). The following CPS excerpt (in italics) describes the basis for the use of a flash fraction of 1.36% at CPS.

Iodine Partition Coefficient(IPC) and POSTDBA Purge Filter Efficiency The IPC is used to calculate the fraction of the iodine in the suppressionpool water that becomes airborneupon reaching the environment. This is modeled in POSTDBA by converting the IPC to a filter efficiency and using this as input for the purge filter efficiency in POSTDBA.

The IPC is defined, at equilibrium, as:

IPC = (concentrationin liquid) / (concentrationin gas)

The IPC has been calculated as a function of molar concentration, pH, and temperature using ORNL-TM-2412. The molar concentrationof iodine, 1CM, in the suppression pool wateris determined from the following expression (1 gram-atom =

1 mole) lCM = (dissolved iodine mass in gram-atoms)/(sup pool water volume in liters)

The mass of iodine in the core at shutdown is 182.3 gram-atoms. 50% of this iodine is dissolved in the suppressionpool water. Therefore the mass of dissolved iodine is 91.15 gram atoms. The suppressionpool volume is 146,400 ft3 which is 4.14559E+6 liters.

Therefore the molar concentrationlCM, is:

/CM = (91.15 gram-atoms)/4.14559E+6

/CM = 2.1987E-5 moles/iter The pH of the suppressionpool is 7 (neutral) because no chemicals have not been added to increase the pH to increaseiodine retention. The temperatureof the suppression pool is assumed to be 80°C (1760F). At a pH of 7.0 and 8°'C ORNL-TM-2412 provides the following values for the IPC.

lCM = 3.46E-5 IPC = 59.7 lCM = 1.73E-5 IPC = 78.5 Using linearinterpolation, the IPC is calculatedto be 73.4 at an lCM of 2.1987E-5 moles/liter. This means that the ratio of iodine in the air to that in the wateris 1/73.4 which in turn means that 1.3624% of the iodine is releasedto the air

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 14 of 18 (1/73.4=1,3624E-2). Therefore the iodine released via the suppression pool leakage must be reduced by a factor of 1.3624%. This corresponds to a filter efficiency of 98.6376 (i.e. 1.3624E-2 of the incoming source is transmitted through the filter). This is rounded up to 98.64 and is input as the purge filter efficiency in the POSTDBA case FWLCSSUP.

Application of this approach with PBAPS specific core iodine inventories and suppression pool volumes yield a slightly higher flashing fraction of 1.41% This historic treatment is very conservative for the AST analyses because: (1) Only 30 percent of core iodine is released to the suppression pool rather than the 50 percent assumed previously; and (2) the suppression pool pH stays above 7.0 throughout the duration of the accident, virtually eliminating the elemental iodine that contributes to iodine flashing.

Therefore, the 1.41% flash fraction is considered acceptable in accordance with paragraph 5.5 of Appendix A to RG 1.183 for ECCS releases from ECCS leakage into the Reactor Building.

The areas involved with ECCS leakage are within the Reactor Building (RB). The RB is serviced by the SGTS. Although SGTS filtration is not credited in the analyses, it provides for an elevated release via the main stack after the initial drawdown period.

25. Page 40 of Attachment 1, Table A, contains a comparison of the Peach Bottom analysis to RG 1.183, Section 4.2.1. The comments column of this table states 'SGTS and MCREV filters are well away and/or shielded from the Control Room and have not historically been considered a source for operator doses. AST assessments would reduce filter loadings because of the credited natural deposition in containment.

Therefore, historical conclusions continue to apply." In light of the many changes proposed (no credit for the standby gas treatment system (SGTS) filters, a new allowable MSIV leakage value, and consideration for plateout in containment) provide a more quantitative assessment justifying why the historical conclusions continue to apply.

RESPONSE: The Peach Bottom SGTS filters are located within the radwaste building on the 91'6" elevation, 73.5 feet below the control room envelope. This distance, coupled with the concrete floors of the 116', 135', and 165' elevations and associated equipment, reduce the dose from the SGTS filters to control room operators to a negligible value.

The MCREV filters are located in the radwaste building on the 165' elevation to the west of the control room envelope. The filters are approximately 30 feet away from the control room west wall. This wall consists of 2 feet of concrete. The radiation from the filters would need to pass through this wall at a significant angle of incidence. Therefore, the previous evaluation of negligible impact remains in place.

26. Section 12.3.3, 'Design Considerations," of the UFSAR states 'The main control room, the Technical Support Center (TSC), and the Emergency Operation Facility (EOF) design is based on the airborne fission product inventory in the reactor building following the design-basis LOCA in Unit 2 or 3, using a TID-14844 source term. Shielding and ventilation air treatment are provided such that operators occupying the control room, the TSC, and the EOF and traveling to and from the control room across the site will receive an exposure of less than 5 Rem whole body or its equivalent over the course of the accident." Page 42 of Attachment 1 states "The Technical Support Center at PBAPS is in the Unit 1 Control Room. A review of the current TID-14844-based analysis

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 15 of 18 indicates that it is unnecessary to reanalyze doses therein to assure accessibility. For other areas requiring plant personnel access, a qualitative assessment of the regulatory positions on source terms indicates that, with no new operator actions required, radiation exposures are bounded by those previously analyzed.' Please provide more details regarding these assessments. Justify the conclusions reached by these qualitative assessments.

RESPONSE: Response to this question will be provided in a future supplement.

27. Page 49 of Attachment 1, Table B, contains a comparison of the Peach Bottom analysis to Section 4.5 of RG 1.183. The comment column of this table states 'However, based on revised containment pressure analysis, the revised TS MSIV leakage is limited to 174 scfh." Proposed insert A (for SR 3.6.1.3.14 on TS page B 3.6-29) states that the total leakage through all four main steamlines must be less than 250 scfh. Please explain this apparent inconsistency.

RESPONSE: Response to this question will be provided in a future supplement.

28. Page 52 of Attachment 1, Table B, contains a comparison of the Peach Bottom analysis to Section 6.1 of RG 1.183. The PBAPS analysis column of this table states that it conforms with RG 1.183, but this RG does not endorse mixing between the drywell and the suppression chamber air volume to determine the source term for the MSIV leakage.

The assumption that the radioactive release is assumed to instantaneously mix between these two volumes appears to be inconsistent with the timing of the AST.

a. Is this based on the results of thermal-hydraulic analyses performed for the duration of the release? If so, provide a summary of the analyses for staff review, or

b. Provide justification for this assumption for the duration of the release.

RESPONSE: Response to this question will be provided in a future supplement.

29. From the Peach Bottom UFSAR, Table 5.2.1, Rev. 17, the minimum drywell and suppression pool free volumes are 159,000 and 127,700 cubic feet, respectively. The minimum total containment free volume is therefore, 286,700 cubic feet. Justify the use of 293,900 cubic feet provided in Table 3, on page 27 of Attachment 1. Why is the more conservative UFSAR value not valid for the LOCA analysis?

RESPONSE: The volume includes the 7,200 cubic feet in the reactor vessel above normal water level, and is used for consistency with the Local Leak Rate Test/integrated Leak Rate Test design basis. The total volume is 286,700 cubic feet plus 7,200 cubic feet, or 293,900 cubic feet.

30. On Page 15 of Attachment 1 to your submittal, your first paragraph states that Exelon has used the Powers model for main steamline deposition. The discussion and the data are insufficient to support staff confirmation. Please provide the values used to input into the RADTRAD model and justify these values. Confirm that the statement "at the 10% probability level" corresponds to the 10th percentile decontamination factors or

'lower bound' as discussed in NUREG/CR-6189.

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 16 of 18 RESPONSE: The Powers model was not used for main steam line deposition. The first paragraph on Page 15 of Attachment 1 to the submittal states that the activity of elemental iodine and aerosols released from the core into the primary containment is reduced by deposition (i.e., plateout) and settling utilizing the Powers' natural deposition values identified in the RADTRAD code at the 10% probability level. This 10% value refers to the 10th percentile decontamination factors or *lower bound" as discussed in NUREG/CR-6189.

31. Confirm that the control room and SGTS flow rates assumed in the accident analysis are conservative. For example, the MCREV system flow rates in the TSs appear to allow flow rates from 2700 to 3300 cfm. The value used for the LOCA is 3000 cfm. Does the assumption of 3000 cfm provide the most limiting control room dose for the LOCA?

Confirm that the intake flows assumed for the fuel handling accident provides the limiting control room doses. This evaluation should include other control room intake flow rates if they are allowed by operating procedures. For example, operating procedures may allow normal intake flows (20,600 cfm) for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and then change to emergency flow (3000 cfm). If the control room filters are credited this probably would not lead to a limiting operator dose. If the filters are not credited, this scenario would provide a more limiting dose than assuming the flow remained at 20,600 cfm.

RESPONSE: Use of nominal HVAC system flows is the design basis accident analysis and accepted practice at PBAPS. Use of the worst-case flow range is generally not a significant impact on calculated doses and unduly complicates analyses.

Since secondary containment mixing is not credited in the LOCA analysis, the volume is set artificially low and the flow artificially high. Actual SGTS design flow rate has no meaning in this instance.

The FHA calculation is performed without MCREV filter credit. The normal intake flow rate is used and augmented with the 1600 cfm of inleakage for conservatism. If MCREV is available then isolation would reduce intake flow and provide filtration.

Introduction of MCREV flow after the available activity has been taken into the control room is not desired because normal intake provides a more rapid purge with fresh air.

Appropriate operator decisions in this regard can be expected based on available instrumentation and health physics coverage.

32. Page 55 of Attachment 1, Table C, contains a comparison of the Peach Bottom analysis to Section 2.0 of RG 1.183. The PBAPS analysis column of this table states that it conforms with RG 1.183, but this RG does not find the use of a Decontamination Factor (DF) of 200 acceptable for less than 23 feet of water covering a damaged fuel assembly.

Please provide the DF used for 21.5 feet of water and the parameters, methodology and justification used to calculate this value.

RESPONSE: Attachment 5 to this response contains information that supports use of a DF of 200. The bounding FHA is one based on a drop within the reactor vessel. This drop results in more kinetic energy, which causes more damage. A drop in the spent fuel pool does not cause as much damage as one in the vessel, offsetting the difference in DF. Therefore, the bounding DF of 200 is used.

Supplement to PBAPS AST RAI Attachment 1 March 15,2004 Page 17 of 18

33. More detail regarding the main steamline break (MSLB) accident is needed. Please provide the reactor coolant system (RCS) activity used for the MSLB analysis and the parameters use to determine this activity. The second bulleted item on page 15 of Attachment 1, states that the activity in the steam cloud is based on the total mass of water released from the break. Confirm that the total activity released for this accident is the RCS specific activity times the break discharge mass (190,920 Ibm). If this is not the methodology used, please provide more detail regarding the model utilized. Also, provide the input parameters used to calculate and justify the fraction of liquid water contained in the steam (2%) and the flashing fraction of liquid water released (40%).

RESPONSE: Appendix 7 to this response contains additional information regarding the determination of doses due to the MSLB. No fuel damage is assumed. Therefore, the activity available for release from the break is that present in the reactor coolant and steam lines prior to the break. Two cases are analyzed. Case 1 is for continued full power operation with a maximum equilibrium concentration of 0.2 uCVgm dose equivalent 1-131. Case 2 is for a maximum coolant concentration of 4.0 uCVgm dose equivalent 1-131, based on a pre-accident iodine spike caused by power changes. This accident source term meets the guidance provided in Regulatory Guide 1.183 for analysis of this event.

The total activity released for this accident is the RCS specific activity times the break discharge mass (165,120 Ibm water + 25,800 Ibm steam = 190,920 Ibm total).

The fraction of liquid water contained in steam, which carries activity into the cloud, is assumed to be 2%, a conservatively high value consistent with current Boiling Water Reactor practice.

A conservatively high flashing fraction of liquid water released of 40% is assumed.

However, all activity in the water is assumed to be released. The flashing fraction (FF) is derived as follows:

FF x (steam enthalpy at 212 F) + (1-FF) x (liquid enthalpy at 212 F) = (liquid enthalpy at temperature of steam at reactor vessel outlet)

A 548 F vessel outlet temperature is used, with liquid enthalpy of 546.9 BTU/lb. At 212 F, a steam enthalpy of 1150.5 BTU/lb and a liquid enthalpy of 180.17 BTU/lb are used (these enthalpies are taken from the ASME Steam Tables).

Substituting, FF = (546.9 - 180.17) / [(1150.5 - 180.17)1 = .378 For conservatism, a value of .40 or 40% is used.

34. In Attachment 1, page 30, Table 8, a value of 0.77% damaged fuel with melt is provided for the control rod drop accident (CRDA). The value typically used for fuel melt with General Electric 14 fuel is 1 % for the CRDA. Please confirm this value of 0.77%.

RESPONSE: This confusion may have resulted from Exelon Round-up of the historical value on the Clinton AST application. For PBAPS, a reference for the higher value could not be found, nor indication of its use by any plant besides Clinton.

Supplement to PBAPS AST RAI Attachment 1 March 15, 2004 Page 18 of 18

35. Please provide the atmospheric dispersion factors used for the CRDA calculations for the exclusion area boundary and low-population zone doses. Are these the values provided in Table 14b on page 32?

RESPONSE: The Exelon CRDA calculation was reviewed to ensure the correct X/Q values were used. These values are correct and are the same as those provided in Table 14b, page 32 of the submittal.

Appendix I Core Isotopic Inventory Appendix 1 Core Isotopic Inventory:

An equilibrium two-year cycle design for Peach Bottom Unit 3 based on the FCYCLE01 code using the actual Cycle 14 design as a reference cycle and a cycle length of 711 EFPD is used as the basis for the source term calculation. In addition, batch average burnups were increased to account for a 1.62% Caldon power uprate. The resulting batch-average burnups for once-burned, twice-burned and thrice-burned fuel batches are shown below.

Avg. Burnup Avg. Enr. per Cycle Power Loading U235 wt j U238 wt. Oxygen wt.

Batch #of FA (Wo U235) (MWd/mtU) NMW) ( MTU) (gMS) (gMS) (gms) l 208 4.105 22800.49 1197.0 37.3256 1,532,215.88 35,793,384.12 5,018,379.87 18256.15 958.4 12116.92 636.1 2 280 4.107 22800.49 1611.5 50.2516 2,063,833.21 48,187,766.79 6,756,264.28 18256.15 1290.3 3 276 4.107 22800.49 1588.5 49.53372 2,034,349.88 47,499,370.12 6,659,746.22 The equilibrium cycle isotopic core inventory is calculated using ORIGEN2.1 and the BWR extended burnup cross-section library BWRUE .

The specific power for a batch in a given cycle is determined by multiplying the batch average burnup for that cycle by the batch loading and then dividing by the number of EFPD in the cycle.

For example, the specific power for Batch 1 in its first cycle of operation is:

(22,800.49

• 37.3256) / 711 = 1197.0 MW.

The grams of U235 and U238 for each batch were determined by the following formulas:

U235 (gins) = Batch loading * (Avg. Enr./100)

• 106 U238 (gins) = Batch loading * (1 - Avg. Enr.l100)

• 106 The corresponding weight of oxygen in U0 2 pellets for each batch is:

0 (gins) = Total batch U weight (gm U) / 238 (gm U/gm atom U)

• 2 (gm atom O/gm atom U)

• 15.9994 gm O/gm atom 0 The ORIGEN2.1 input deck is set up to deplete each fuel batch and write the 100 EFPD and EOC results to temporary storage vectors. Once all batches have been depleted, the results from the temporary vectors are combined to give the results for the entire core. The ORIGEN2.1 core inventory activity and composition results for the equilibrium two-year cycle at 100 EFPD (BOC) and EOC are shown below in Tables I and 3, respectively. The maximum of the 100 EFPD and EOC values for each isotope are selected to generate the bounding isotopic core inventory activity and composition results as shown in Tables 2 and 4, respectively.

1

Appendix I Core Isotopic Inventory Table 1 ORIGEN2.1 Isotopic Activity Results for Peach Bottom Unit 3 100 EFPD EOC ll 100 EFPD EOC Isotope [ (Ci) 0 _ _(Ci) Isotope (Ci) l (Ci)

KR 83M 1.324E+07 1.158E+07 XE131M 1.015E+06 1.056E+06 BR 84 2.373E+07 2.002E+07 TE132 1.322E+08 1.343E+08 BR 85 2.888E+07 2.404E+07 1132 1.338E+08 1.364E+08 KR 85 8.806E+05 1.387E+06 1133 1.953E+08 1.925E+08 KR 85M 2.922E+07 2.436E+07 XE133 1.904E+08 1.930E+08 RB 86 1.118E+05 2.291 E+05 XE133M 5.956E+06 6.007E+06 KR 87 5.739E+07 4.672E+07 1134 2.167E+08 2.118E+08 KR 88 8.096E+07 6.570E+07 CS134 1.335E+07 2.559E+07 RB 88 8.197E+07 6.678E+07 1135 1.825E+08 1.806E+08 SR 89 9.836E+07 8.846E+07 XE135 7.832E+07 7.086E+07 SR 9o 6.982E+06 1.117E+07 XE135M 3.636E+07 3.773E+07 Y 90 7.142E+06 1.150E+07 CS136 3.568E+06 7.123E+06 SR 91 1.336E+08 1.t 1OE+08 CS137 9.460E+06 1.595E+07 Y 91 1.212E+08 1.143E+08 BA137M 8.965E+06 1.510E+07 SR 92 1.412E+08 1.205E+08 XE138 1.679E+08 1.589E+08 Y 92 1.416E+08 1.210E+08 CS138 1.841 E+08 1.760E+08 Y 93 1.591 E+08 1.404E+08 BA139 1.787E+08 1.719E+08 ZR 95 1.521 E+08 1.578E+08 BA140 1.721 E+08 1.661 E+08 NB 95 1.399E+08 1.586E+08 LA140 1.764E+08 1.722E+08 ZR 97 1.637E+08 1.578E+08 LA141 1.631 E+08 1.563E+08 MO 99 1.771 E+08 1.785E+08 CE141 1.579E+08 1.575E+08 TC 99M 1.551 E+08 1.563E+08 LA142 1.593E+08 1.510E+08 RU103 1.234E+08 1.477E+08 CE143 1.556E+08 1.449E+08 RU105 7.642E+07 1.022E+08 PR143 1.509E+08 1.416E+08 RHI05 7.31 OE+07 9.673E+07 CE144 1.012E+08 1.264E+08 RU106 3.856E+07 6.081 E+07 ND147 6.459E+07 6.321 E+07 SB127 8.572E+06 1.018E+07 NP239 1.616E+09 1.897E+09 TE127 8.402E+06 1.01OE+07 PU238 2.639E+05 6.312E+05 TE127M 1.039E+06 1.355E+06 PU239 3.100E+04 4.218E+04 SB129 2.732E+07 3.036E+07 PU240 2.871 E+04 4.526E+04 TE129 2.679E+07 2.988E+07 PU241 1.365E+07 2.173E+07 TE129M 3.875E+06 4.453E+06 AM241 1.634E+04 3.349E+04 1129 2.774E+00 4.816E+O0 CM242 3.645E+06 8.393E+06 TE131M 1.269E+07 1.360E+07 CM244 2.654E+05 9.147E+05 1131 9.139E+07 9.444E+07 2

Appendix I Core Isotopic Inventory Table 2 Bounding Isotopic Core Inventory Peach Bottom Unit 3 Isotopic Isotopic Activity Activity Isotope (Ci) Isotope (Ci)

KR 83M 1.324E+07 XE1 31 M 1.056E+06 BR84 2.373E+07 TE1 32 1.343E+08 BR 85 2.888E+07 1132 1.364E+08 KR 85 1.387E+06 1133 1.953E+08 KR 85M 2.922E+07 XE1 33 1.930E+08 RB 86 2.291 E+05 XE1 33M 6.007E+06 KR 87 5.739E+07 1134 2.1 67E+08 KR88 8.096E+07 CS134 2.559E+07 RB88 8.1 97E+07 1135 1.825E+08 SR 89 9.836E+07 XE1 35 7.832E+07 SR 9o 1.1 17E+07 XE135M 3.773E+07 Y90 1.1 50E+07 CS1 36 7.123E+06 SR 91 1.336E+08 CS137 1.595E+07 Y91 1.21 2E+08 BA137M 1.51 OE+07 SR92 1.41 2E+08 XE1 38 1.679E+08 Y92 1.41 6E+08 CS138 1.841 E+08 Y93 1.591 E+08 BA1 39 1.787E+08 ZR 95 1.578E+08 BA1 40 1.721 E+08 NB 95 1.586E+08 LA140 1.764E+08 ZR 97 1.637E+08 LA141 1.631 E+08 MO 99 1.785E+08 CE141 1.579E+08 TC 99M 1.563E+08 LA1 42 1.593E+08 RU103 1.477E+08 CE143 1.556E+08 RU105 1.022E+08 PR1 43 1.509E+08 RH105 9.673E+07 CE144 1.264E+08 RU106 6.081 E+07 ND147 6.459E+07 SB1 27 1.01 8E+07 NP239 1.897E+09 TE1 27 1.01 OE+07 PU238 6.31 2E+05 TE127M 1.355E+06 PU239 4.21 8E+04 SB1 29 3.036E+07 PU240 4.526E+04 TE129 2.988E+07 PU241 2.1 73E+07 TE129M 4.453E+06 AM241 3.349E+04 1129 4.81 6E+00 CM242 8.393E+06 TE1 31M 1.360E+07 CM244 9.1 47E+05 1131 9.444E+07 3

Appendix I Core Isotopic Inventory Table 3 ORIGEN2.1 Isotopic Concentration Results for Peach Bottom Unit 3 ll 100EFPD l___EOC H 100 EFPD EOC Isotope (grams) (grams) Isotope (grams) (grams)

KR 83M 6.414E-01 5.611E-01 XE131M 1.212E+01 1.260E+01 BR 84 3.370E-01 2.843E-01 TE132 4.352E+02 4.423E+02 BR 85 3.741E-02 3.114E-02 1132 1.296E+01 1.321E+01 KR 85 2.244E+03 3.534E+03 1133 1.723E+02 1.699E+02 KR 85M 3.550E+00 2.959E+00 XE133 1.017E+03 1.031E+03 RB 86 1.373E+00 2.814E+00 XE133M 1.328E+O1 1.339E+01 KR 87 2.025E+00 1.649E+00 CS133 1.025E+05 1.678E+05 KR 88 6.451 E+00 5235E+00 1134 8.118E+00 7.936E+00 RB 88 6.826E-01 5.561 E-01 CS134 1.031 E+04 1.977E+04 SR 89 3.384E+03 3.044E+03 1135 5.195E+01 5.140E+01 SR 90 5.116E+04 8.183E+04 XE135 3.065E+01 2.773E+01 Y90 1.312E+01 2.113E+01 XE135M 3.990E-01 4.140E-01 SR 91 3.683E+01 3.060E+01 CS135 4.502E+04 7.841E+04 Y 91 4.939E+03 4.658E+03 CS136 4.867E+01 9.715E+01 SR 92 1.123E+01 9.579E+00 CS137 1.087E+05 1.832E+05 Y 92 1.471E+01 1.256E+01 BA137M 1.666E-02 2.807E-02 Y 93 4.768E+01 4.206E+01 XE138 1.745E+00 1.652E+00 ZR 95 7.079E+03 7.341 E+03 CS138 4.348E+O0 4.156E+00 NB 95 3.576E+03 4.054E+03 BA139 1.092E+01 1.051E+01 ZR 97 8.560E+01 8.251E+01 BA140 2.359E+03 2.277E+03 MO 99 3.691 E+02 3.720E+02 LA140 3.168E+02 3.093E+02 TC 99M 2.947E+01 2.971 E+01 LA141 2.883E+01 2.762E+01 RU103 3.822E+03 4.575E+03 CE141 5.541 E+03 5.528E+03 RUt05 1.136E+01 1.519E+01 LA142 1.115E+01 1.056E+01 RHI05 8.657E+01 1.146E+02 CE143 2.342E+02 2.181E+02 RU106 1.152E+04 1.817E+04 PR143 2.241E+03 2.102E+03 SB127 3.208E+01 3.811E+01 CE144 3.172E+04 3.962E+04 TE127 3.182E+00 3.826E+00 ND147 8.039E+02 7.867E+02 TE127M 1.101E+02 1.436E+02 NP239 6.963E+03 8.173E+03 1127 4.533E+03 8.040E+03 PU238 1.541 E+04 3.685E+04 SB129 4.856E+00 5.397E+00 PU239 4.985E+05 6.782E+05 TE129 1.279E+00 1.426E+00 PU240 1.259E+05 1.986E+05 TE129M 1.286E+02 1.478E+02 PU241 1.324E+05 2.108E+05 1129 1.571 E+04 2.727E+04 AM241 4.759E+03 9.755E+03 TE131M 1.591E+01 1.704E+01 CM242 1.102E+03 2.537E+03 1131 7.369E+02 7.615E+02 CM244 3.279E+03 1.130E+04 4

Appendix I Core Isotopic Inventory Table 4 Bounding Isotopic Core Inventory Peach Bottom Unit 3 Isotopic Isotopic Concentration Concentration Isotope (grams) Isotope (grams)

KR 83M 6.41 4E-01 XE131M 1.260E+01 BR 84 3.370E-01 TE1 32 4.423E+02 BR 85 3.741 E-02 1132 1.321 E+01 KR 85 3.534E+03 1133 1.723E+02 KR 85M 3.550E+00 XE133 1.031 E+03 RB 86 2.81 4E+00 XE133M 1.339E+01 KR 87 2.025E+00 CS133 1.678E+05 KR 88 6.451 E+00 1134 8.118E+00 RB 88 6.826E-01 CS1 34 1.977E+04 SR89 3.384E+03 1135 5.1 95E+01 SR90 8.183E+04 XE1 35 3.065E+01 Y90 2.1 13E+01 XE135M 4.1 40E-01 SR 91 3.683E+01 CS1 35 7.841 E+04 Y91 4.939E+03 CS1 36 9.71 5E+01 SR 92 1.123E+01 CS137 1.832E+05 Y92 1.471 E+01 BA137M 2.807E-02 Y93 4.768E+01 XE1 38 1.745E+00 ZR 95 7.341 E+03 CS138 4.348E+00 NB 95 4.054E+03 BA1 39 1.092E+01 ZR 97 8.560E+01 BA140 2.359E+03 MO 99 3.720E+02 LA1 40 3.1 68E+02 TC 99M 2.971 E+01 LA1 41 2.883E+01 RU1 03 4.575E+03 CE141 5.541 E+03 RU105 1.51 9E+01 LA142 1.115E+01 RH105 1.146E+02 CE143 2.342E+02 RU106 1.817E+04 PR143 2.241 E+03 SB1 27 3.811 E+01 CE1 44 3.962E+04 TE127 3.826E+00 ND1 47 8.039E+02 TE127M 1.436E+02 NP239 8.1 73E+03 1127 8.040E+03 PU238 3.685E+04 SB129 5.397E+00 PU239 6.782E+05 TE1 29 1.426E+00 PU240 1.986E+05 TE129M 1.478E+02 PU241 2.108E+05 1129 2.727E+04 AM241 9.755E+03 TE1 31M 1.704E+01 CM242 2.537E+03 1131 7.61 5E+02 CM244 1.130E+04 5

Appendix I Core Isotopic Inventory Fuel Handling Accident Source Term (.nif) File Nuclide Inventory Name: Source Terms per this calculation Peach Bottom (PBAPS) FHA AST - in Ci/MW Power Level:

0.1000E+01 Nuclides:

60 Nuclide 001:

Co-58 7

0.6117120000E+07 0.5800E+02

0. 1529E+03 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 002:

Co-60 7

0.1663401096E+09 0.6000E+02 0.1830E+03 none 0.0000E+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 003:

Kr-85 1

0.3382974720E+09 0.8500E1+02 0.7892E+03 2.0*LOCA Value for FHA none 0.0000E+00 none 0.OOOOE+00 none 0.0000E+00 Nuclide 004:

Kr-85m I

0.1612800000E1+05 0.8500E+02 0.8313E+04 Kr-85 0.2100E1+00 none 0.0000E+00 none 0.0000E+00 Nuclide 005:

Kr-87 I

0.4578000000E+04 0.8700E+02 0.1633E1+05 Rb-87 0.1000E+01 none 0.OOOOE+00 none 0.0000E+00 Nuclide 006:

Kr-88 I

0.1022400000E+05 0.8800E+02 0.2303E+05 6

Appendix I Core Isotopic Inventory Rb-88 0.1000E+01 none O.OOOOE+00 none 0.0000E+00 Nuclide 007:

Rb-86 3

0.1612224000E+07 0.8600E+02 0.6518E+02 none 0.OOOOE+00 none O.OOOOE+00 none O.OOIOE+00 Nuclide 008:

Sr-89 5

0.4363200000E+07 0.8900E+02 0.2798E+05 none O.OOOOE+00 none O.OOOOE+00 none 0.OOOOE+00 Nuclide 009:

Sr-90 5

0.9189573120E+09 0.9000E+02 0.3178E+04 Y-90 0.1000E+0I none 0.OOOOE+00 none O.OOOOE+00 Nuclide 010:

Sr-91 5

0.3420000000E+05 0.9100E+02 0.3801E+05 Y-91m 0.5800E+00 Y-91 0.4200E+00 none 0.OOOOE+00 Nuclide 011:

Sr-92 5

0.9756000000E+04 Q.9200E+02 0.4017E+05 Y-92 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 012:

Y-90 9

0.2304000000E+06 0.9000E+02 0.3272E+04 none O.OOOOE+00 none 0.0000E+00 none O.OOOOE+00 Nuclide 013:

Y-91 7

Appendix I Core Isotopic Inventory 9

0.5055264000E+07 0.9100E+02 0.3448E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 014:

Y-92 9

0.1274400000E+05 0.9200E+02 0.4029E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 015:

Y-93 9

0.3636000000E+05 0.9300E+02 0.4526E+05 Zr-93 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 016:

Zr-95 9

0.5527872000E+07 0.9500E+02 0.4489E+05 Nb-95m 0.7000E-02 Nb-95 0.9900E+00 none 0.OOOOE+00 Nuclide 017:

Zr-97 9

0.6084000000E+05 0.9700E+02 0.4657E+05 Nb-97m 0.9500E+00 Nb-97 0.5300E-01 none O.OOOOE+00 Nuclide 018:

Nb-95 9

0.3036960000E+07 0.9500E+02 0.4512E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 019:

Mo-99 7

0.2376000000E+06 0.9900E+02 0.5078+05 Tc-99m 0.8800E+00 8

Appendix I Core Isotopic Inventory Tc-99 0.1200E+00 none O.OOOOE+OO Nuclide 020:

Tc-99m 7

0.2167200000E+05 0.9900E+02 0.4447E+05 Tc-99 0.1000E+01 none 0.OOOOE+00 none O.OOOOE+00 Nuclide 021:

Ru-103 7

0.3393792000E+07 0.1030E+03 0.4202E+05 Rh-103m 0.1000E+01 none 0.0000E+00 none O.OOOOE+00 Nuclide 022:

Ru-105 7

0.1 598400000E+05 0.1050E+03 0.2908E+05 Rh- 105 0.I OOOE+O I none 0.0000E+00 none O.OOOOE+O0 Nuclide 023:

Ru-106 7

0.3181248000E+08 0.1060E+03

0. 1730E+05 Rh-106 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 024:

Rh-105 7

0.1272960000E+06

0. 1050E+03 0.2752E+05 none O.OOOOE+O0 none O.OOOOE+00 none O.OOOOE+00 Nuclide 025:

Sb-127 4

0.3326400000E+06 0.1270E+03 0.2896E+04 Te-127m 0.1800E+O0 Te-127 0.8200E+00 none O.OOOOE+O0 Nuclide 026:

Sb-129 4

9

Appendix I Core Isotopic Inventory 0.1 555200000E+05 0.1290E+03 0.8638E+04 Te-129m 0.2200E+00 Te-129 0.7700E+00 none O.OOOOE+00 Nuclide 027:

Te-127 4

0.3366000000E+05 0.1270E+03 0.2873E+04 none O.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 028:

Te-127m 4

0.9417600000E+07 0.1270E+03 0.3855E+03 Te-127 0.9800E+O0 none O.OOOOE+00 none O.OOOOE+00 Nuclide 029:

Te-129 4

0.4176000000E+04 0.1290E+03 0.8501E+04 1-129 0.1000E+01 none O.OOOOE+00 none Q.OOOOE+00 Nuclide 030:

Te-129m 4

0.2903040000E+07 0.1290E+03 0.1267E+04 Te-129 0.6500E+O0 I-129 0.3500E+O0 none O.OOOOE+00 Nuclide 031:

Te-131m 4

0.1080000000E+06 0.13 1OE+03 0.3869E+04 Te-131 0.2200E+O0 I-131 0.7800E+00 none 0.OOOOE+00 Nuclide 032:

Te- 132 4

0.2815200000E+06 0.1320E+03 0.38211E+05 I- 132 0.IOOOE+OI none O.OOOOE+00 10

Appendix I Core Isotopic Inventory none 0.OOOOE+00 Nuclide 033:

1-131 2

0.6946560000E+06 0.13101E+03 0.4299E+05 1.6*LOCA Value for FHA Xe-131m 0.1100E-01 none 0.0000E+00 none 0.OOOOE+00 Nuclide 034:

1-132 2

0.8280000000E+04 0.1320E+03 0.3881E+05 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 035:

1-133 2

0.7488000000E+05 0.1330E+03 0.5556E+05 Xe-133m 0.2900E-01 Xe- 133 0.9700E+00 none 0.0000E+00 Nuclide 036:

1-134 2

0.3156000000E+04 0.1 340E+03 0.6165E+05 none 0.OOOOE+00 none 0.OOOOE+00 none 0.0000E+00 Nuclide 037:

1-135 2

0.2379600000E+05 0.1350E+03 0.5192E+05 Xe-135m 0.1500E+00 Xe-135 0.8500E+00 none 0.0000E+00 Nuclide 038:

Xe-133 I

0.4531680000E+06 0.1330E+03 0.5491E+05 none 0.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 039:

Xe-135 I

0.3272400000E3+05 11

Appendix I Core Isotopic Inventory 0.1350E+03 0.2228E+05 Cs-135 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 040:

Cs-134 3

0.6507177120E+08 0.1340E+03 0.7280E+04 none 0.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 041:

Cs-136 3

0.1 131840000E+07 0.1360E+03 0.2027E+04 none O.OOOOE+00 none 0.OOOOE+00 none O.OOOOE+00 Nuclide 042:

Cs- 137 3

0.9467280000E+09 0.1370E+03 0.4538E+04 Ba-137m 0.9500E+00 none 0.OOOOE+00 none O.OOOOE+00 Nuclide 043:

Ba-139 6

0.4962000000E+04 0.1390E+03 0.5084E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 044:

Ba- 140 6

0.1 100736000E+07 0.1400E+03 0.4896E+05 La-140 0.IOOOE+O I none O.OOOOE+00 none O.OOOOE+00 Nuclide 045:

La-140 9

0.1449792000E+06 0.1400E+03 0.5019E+05 none O.OOOOE+00 none 0.OOOOE+00 none O.OOOOE+00 12

Appendix I Core Isotopic Inventory Nuclide 046:

La-141 9

0.1414800000E+05 0.1410E+03 0.4640E+05 Ce-141 0.IOOOE+01 none O.OOOOE+00 none 0.OOOOE+00 Nuclide 047:

La-142 9

0.5550000000E+04 0.1420E+03 0.4532E+05 none 0.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 048:

Ce-141 8

0.2808086400E+07 0.1410E+03 0.4492E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 049:

Ce- 143 8

0.1 188000000E+06

0. 1430E+03 0.4427E+05 Pr-143 0.IOOOE+01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 050:

Ce-144 8

0.2456352000E+08 0.1440E+03 0.3596E+05 Pr-144m 0.1800E-01 Pr-144 0.9800E+00 none O.OOOOE+00 Nuclide 051:

Pr- 143 9

0.1 171584000E+07

0. 1430E+03 0.4293E+05 none 0.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 052:

Nd-147 9

0.9486720000E+06

0. 1470E+03 13

w.

Appendix I Core Isotopic Inventory 0.1838E+05 Pm-147 0.IOOOE+OI none O.OOOOE+00 none O.OOOOE+00 Nuclide 053:

Np-239 8

0.2034720000E+06 0.2390E+03 0.5397E+06 Pu-239 0.IOOOE+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 054:

Pu-238 8

0.2768863824E+10 0.2380E+03

0. 1796E+03 U-234 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 055:

Pu-239 8

0.7594336440E+ 12 0.2390E+03 0.1200E+02 U-235 0.lOOOE+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 056:

Pu-240 8

0.2062920312E+12 0.2400E+03 0.1288E+02 U-236 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 057:

Pu-241 8

0.4544294400E+09 0.2410E+03 0.6182E+04 U-237 0.2400E-04 Am-241 0.IOOOE+01 none O.OOOOE+00 Nuclide 058:

Am-241 9

0.1363919472E+I I 0.2410E+03 0.9528E+01 Np-237 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 059:

14

Appendix I Core Isotopic Inventory Cm-242 9

0.1406592000E+08 0.2420E+03 0.2388E+04 Pu-238 0.IOOOE+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 060:

Cm-244 9

0.5715081360E+09 0.2440E+03 0.2602E+03 Pu-240 0.1000E+01 none 0.OOOOE+00 none O.OOOOE+00 End of Nuclear Inventory File 15

Appendix 2 General AST Analysis Design Inputs for PBAPS Appendix 2 General AST Analysis Design Inputs for PBAPS General AST Analysis Design Inputs for PBAPS AST Value Assumption/Comment (Applies to All Accident Analyses)

Core Power Level 3528 MWth This value corresponds to the DBA power level, including margin, above the Rated Thermal Power Level of 3514 MWth, to account for instrument uncertainty.

Core Source Terms See Appendix I of this Calculated in PBAPS Design Analysis No.

response PM-1059 Dose Conversion Factors FGR 11 and 12 for Values are built into RADTRAD file Inhalation CEDE and FGR1 1&12.INP for a total of 60 isotopes.

cloud submersion EDE.

EAB - x/Q's Calculated in PBAPS Design Analysis PM-1055, Rev. 0. Representative site meteorological data used is from 1984-1988.

From Main Off-gas Stack Elevated, 500 ft., release from the Main Off-Dispersion Factors: gas Stack at PBAPS.

0 - 0.5 hr 5.30E-05 (sec/m 3 )* *Used during fumigation period.

0 - 2 hr 8.89E-06 (sec/M3 )

From TB/RB Exhaust Vent Dispersion Factors:

0- 2 hr 4.25E-04 (sec/m 3 ) Worst case '/Q from either of two TB/RB Exhaust Vents.

LPZ - '/Q's Meteorological data from 1984-1988.

From Main Off-gas Stack Dispersion Factors: Elevated, 500 ft., release from the Main Off-0 - 0.5 hr 1.75E-05 (sec/m 3 )* gas Stack at PBAPS.

0 - 2 hr 8.87E-06 (sec/n 3 )

• Used during fumigation period.

2 - 8 (0- 8) hr 3.94E-06 (sec/m 3 )

8 - 24 hr 2.62E-06 (sec/M3 )

I - 4 day 1.09E-06 (sec/m 3 )

4 - 30 day 3.06E-07 (sec/m 3 )

From TB/RB Exhaust Vent Worst case Z/Q from either of two TB/RB Dispersion Factors: Exhaust Vents.

0 - 2 hr 4.8 1E-05 (sec/m 3 )

2 - 8 (0- 8) hr 2.08E-05 (sec/m 3 )

8 - 24 hr 1.37E-05 (sec/m 3 )

1 - 4 day 5.49E-06 (sec/m 3 )

4 - 30 day 1.49E-06 (sec/m 3 )

16

Appendix 2 General AST Analysis Design Inputs for PBAPS General AST Analysis Design Inputs for PBAPS AST Value Assumption/Comment (Applies to All Accident Analyses)

CR - '/Q's Meteorological data from 1984-1988.

From Main Off-gas Stack Elevated, 500 ft., release from the Main Off-Dispersion Factors: gas Stack at PBAPS.

0- 2 hr 2.72E-06 (sec/r 3 )

2- 8 hr 1.00E-09 (sec/m 3 )* *The 2-8 hour and 8-24 hour period x/Q's 8- 24 hr 1.OOE-09 (sec/m 3 )* were conservatively increased from the 1- 4 day 1.46E-08 (sec/m 3 ) calculated value of 1.OOE-15, to an artificial 4- 30 day 4.21E-09 (sec/M3 ) floor of L.OOE-09.

From TB/RB Exhaust Vent Worst case x/Q from either of two TB/RB Dispersion Factors: Exhaust Vents.

0 - 2 hr 1.18E-03 (sec/m3 )

2 - 8 hr 9.08E-04 (sec/n 3 )

8 - 24 hr 4.14E-04 (sec/m 3) 1 -4 day 2.90E-04 (sec/n 3) 4 - 30 day 2.26E-04 (sec/rn3)

Control Room Volume 176,000 ft3 MCREV Initiation Timing 0 sec Automatic isolation of the Control Room is

- (following 2-minute assumed to be within the 2-minute period period, before start of gap preceding the start of gap release.

release)

Intake rate is assumed to be unchanged when 3,000 cfm MCREV initiates.

Intake Rate HEPA Filter requirements will remain Intake Filter Efficiency 90% unaffected by this LOCA analysis. Subject to Elemental Iodine 90% change based on re-analysis for MSIV Organic Iodine 99% leakage. Will be finalized in a future Aerosols response.

Maximized for operational margin. PBAPS Unfiltered Inleakage 800 cfm previously analyzed at 10 cfm. Subject to change based on re-analysis for MSIV leakage. Will be finalized in a future response.

I 17

Appendix 3 LOCA AST Analysis Design Inputs for PBAPS Appendix 3 LOCA AST Analysis Design Inputs for PBAPS LOCA AST Analysis Design AST Value Assumption/Comment Inputs for PBAPS Primary Containment Volumes typically rounded to 4 places when Volume used in RADTRAD analyses.

Containment (Drywell + Because leak rates are tested at 49.1 psig, and Torus Airspace): Local Leak Rate Test (LLRT) results show Peak Containment leakage of 125,417 sccm or 4.43 scfm, back-Pressure, Pa 49.1 psig calculation of the Containment volume is used for analyzing leaks from containment.

LLRT Result 125,417 sccm Calculated Volume 293,900 ft3 This volume indicates the assumed well-mixed condition of the total containment volume, immediately following a LOCA and includes 7200 ft3 of volume within the reactor vessel.

Minimum Suppression Pool 122,900 ft3 Water Volume Releases to Containment No Core Activity Release RADTRAD runs use a starting point of time for first 121 seconds. 0, which is artificially the start of gap release.

Credit for the -2 minute delay can be credited Release Fractions and as additional margin available for credited Timing per NUREG- time dependent system responses. For 1465 as shown in instance, in the RADTRAD analyses, a RADTRAD File drawdown time of 15 minutes is used. Actual BWRDEF.rft available time would be 17 minutes.

[Attachment Al Containment isolation is assumed to be instantaneous in the RADTRAD analyses.

The 2-minute delay can be used in evaluating the acceptability of actual isolation valve closure times.

Containment Activity Natural Deposition Credit for Natural Deposition in Containment Removal Mechanisms is achieved through the use of the Power's Model, as implemented in the RADTRAD code, using the 10% probability level.

18

Appendix 3 LOCA AST Analysis Design Inputs for PBAPS LOCA AST Analysis Design AST Value Assumption/Comment Inputs for PBAPS SGTS Since credit is not taken for mixing in Secondary Containment, an artificially low Flow Rate 10,500 cfm volume (I ft3), and an artificially high SGTS flow rate (104 cfm) are used as RADTRAD input.

Filter Efficiency 0%, for all radionuclides. This LOCA analysis assumes no initialization of SGTS filtration.

Drawdown Timing 15 minutes This analysis assumes this increase over the existing Tech. Spec. value of 2 minutes to provide operational margin.

Primary Containment Leak The PBAPS Containment is assumed to leak Rate 0.70% Iday, 0-24 hrs at this rate corresponding to the peak pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Based on analysis of 0.39% /day, 24-38 hrs Containment leak rate as function of Drywell pressure, the Primary Containment leak rate is 0.35% /day,38-720 hrs reduced to 56% of the initial value after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then to 50% of the initial value after 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br /> and for the accident duration.

MSIV Leakage Rates Acceptable values to be re-determined based Total on results of discussion with NRC reviewer and re-calculation as required. This will be (NPSH Bounding Case) covered in a future response.

MSL D (15' shortest)

MSL C (2nd shortest)

MSL B (3 rd shortest)

MSL A (4h shortest)

Total 19

Appendix 3 LOCA AST Analysis Design Inputs for PBAPS LOCA AST Analysis Design AST Value Assumption/Comment Inputs for PBAPS MSIV Leakage Mitigating Acceptable values to be re-determined based Factors on results of discussion with NRC reviewer Deposition in piping is and re-calculation as required. This will be only credited for covered in a future response.

Horizontal runs using RADTRAD Brockmann-Bixler Algorithms, which require the following pipe parameters:

- Temperature

- Pressure

- Surface Area

- Volume

- Release Timing and Holdup/Delay ECCS Leakage into Values commensurate with at least 2 times Secondary Containment any current acceptance value used as part of T.S. Program 5.5.2 "Primary Coolant Sources Leak Rate 5 gpm Outside Containment".

Fraction Flashed 1.41% Filtration is no longer credited through the SGTS. Only the elevated release of the Off-Filtered by SGTS No gas Stack is credited.

20

Appendix 4 FHA AST Analysis Design Inputs for PBAPS Appendix 4 FHA AST Analysis Design Inputs for PBAPS FHA AST Analysis Parameter or Method for AST Value Assumption/Comment PBAPS Reactor Power 3528 MWth Fuel Assembly Configuration lOx 10 in a 87.33 fuel pin and properties bundle and 172 pins damaged Peaking Factor 1.7 Allowable Fuel Burmup and non- RG 1.183, Table 3 LOCA gap fractions FHA Radionuclide Inventory From Attachment 1 of this See Attachment 1 Calc. for the 60 isotopes forming the standard RADTRAD library, with decay to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Gap activities per R.G. 1.183.

Underwater Decontamination Noble Gases: I See Appendix 5 of this response for Factor additional information regarding water Particulate (cesiums and coverage.

rubidiums): infinity Iodine: 200, corresponding to a 23-ft water depth Iodine chemical distribution From RG 1.183 (95% CsI, Per RG 1.183 (pH not maintained) instantaneously dissociating in the pool water and re-evolving as elemental iodine, unless the pH of the pool water is justified mechanistically to be above 7; 4.85% elemental; 0.15%

organic)

Dose Conversion Factors EPA Federal Guidance EPA Federal Guidance Reports 11 Reports I Iand 12 and 12 Offsite Dose Limit 6.3 rem TEDE after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> RG 1.183 Control Room Dose Limit 5 rem TEDE for the duration 10CFR50 App. A, GDC 19 and of the accident IOCFR50.67 Secondary Containment Not credited for all fuel with RG 1.183 Automatic Isolation and one day decay time Filtration Mitigation by CREF system Not credited for all fuel with RG 1.183 one day decay time Normal Control Room Fresh Design Basis Intake 3000 Subject to change - may decrease to Air Makeup Rate and Volume cfm. 800 cfm, but existing calc then overly 4600 cfm used for conservative, but not restrictive.

conservatism and to allow for 1600 cfm of additional inleakage.

21

Appendix 4 FHA AST Analysis Design Inputs for PBAPS FHA AST Analysis Parameter or Method for AST Value Assumption/Comment PBAPS Volume 176,000 ft3 Refuel Floor Normal Approximately 6 air changes This evacuates 99.9999% of all Ventilation rate and volume per hour and an artificial activity within 2-hours.

value of 100 ft3 is used for simplicity.

CR Release Point Basis and TB/RB ventilation stack and Distance to CR intake 191.7 ft Dispersion Factors 0 - 2 hr 1.18E-03 sec/m 3 EAB Release Point Basis and Normal RB exhaust stack Distance to EAB and 1040 m Dispersion Factors 0 - 2 hr 4.25E-04 sec/m3 LPZ Release Point Basis and Normal RB exhaust stack Distance to LPZ and 7300 m Dispersion Factors 0- 8 hr 4.8 1E-05 sec/m 3 22

Appendix 5 PBAPS Fuel Handling Accident Assessment of Limiting Event (Water Coverage)

Appendix 5 FHA Assessment of Limiting Event and Water Coverage PBAPS Fuel Handling Accident Assessment of Limiting Event This Appendix:

[a] Evaluates water coverage for FHAs over the Reactor Well and over the Spent Fuel Pool.

[b] Evaluates impact of water coverage of less than 23 feet for purposes of pool DF determination.

[c] Justifies that a FHA over the Reactor Well is the limiting event.

Baseline R.G. 1.183 based Analysis of DFs RG 1.183 RG 1.183 Water RG 1.183 RG 1.183 Inorganic Organic Coverage Inorganic Organic iodine Iodine DF Overall (feet) Iodine DF Iodine DF Fraction Fraction DF 23 500 1 0.9985 0.0015 286.0 Case 1: Inorganic iodine DF Guidance Controlling 23 285.3 1 0.9985 0.0015 200.0 Case 2: Overall DF Guidance Controlling DFs determined per Burley Paper with R.G. 1.183 Case 1 assumptions RG 1.183 RG 1.183 Water RG 1.183 RG 1.183 Inorganic Organic Coverage Inorganic Organic Iodine iodine DF Overall (feet) Iodine DF Iodine DF Fraction Fraction DF 23 500 1 0.9985 0.0015 286.0 capped at 200 22.5 436.8 1 0.9985 0.0015 264.1 capped at 200 22 381.6 1 0.9985 0.0015 242.9 capped at 200 21.5 333.4 1 0.9985 0.0015 222.5 capped at 200 21 291.3 1 0.9985 0.0015 202.9 capped at 200 20.5 254.5 1 0.9985 0.0015 184.4 20 222.3 1 0.9985 0.0015 166.9 19.5 194.2 1 0.9985 0.0015 150.6 19 169.7 1 0.9985 0.0015 135.4 All water coverages are more than 21 feet. Therefore, the 200 DF is conservative for all cases.

Appendix 5 PBAPS Fuel Handling Accident Assessment of Limiting Event (Water Coverage)

DFs determined per Burley Paper with R.G. 1.183 Case 2 assumptions RG 1.183 RG 1.183 Water RG 1.183 RG 1.183 Inorganic Organic Coverage Inorganic Organic Iodine Iodine DF Overall (feet) Iodine DF Iodine DF Fraction Fraction DF 23 285.3 1 0.9985 0.0015 200.0 22.5 252.3 1 0.9985 0.0015 183.2 22 223.1 1 0.9985 0.0015 167.4 21.5 197.3 1 0.9985 0.0015 152.4 21 174.5 1 0.9985 0.0015 138.5 20.5 154.3 1 0.9985 0.0015 125.5 20 136.5 1 0.9985 0.0015 113.4 19.5 120.7 1 0.9985 0.0015 102.3 19 106.7 1 0.9985 0.0015 92.1 21.473 196.0 1 0.9985 0.0015 151.7 22.644 261.4 1 0.9985 0.0015 188.0 Overall DF weighted by in rack vs. drop assemblies: 169.5 For Case 2, overall % of 200 DF 84.8%

Fuel failure over SFP vs. over the reactor well. 70%

Based on recent Fermi 2 assessments of a 6 foot drop (and a similar Mark 1 configuration).

Therefore, the drop over the reactor well is bounding when a DF of 200 is used.

24

Appendix 5 PBAPS Fuel Handling Accident Assessment of Limiting Event (Water Coverage)

PBAPS Damaged Fuel Water Coverage Assessment for Fuel Handling Accidents Reference Points MSL Reactor 0 Rx. Inst. 0 SPF Bottom (feet) (feet) (feet) (feet) 232.542 84.292 39.458 37.292 Pool Level for SPF Cooling Operation (scuppers all the way down) 232.250 84.000 39.167 37.000 TS 3.7.7 Minimum Spent Fuel Pool Level for fuel movement 231.250 83.000 38.167 36.000 TS 3.9.6 Minimum Water Level During Refueling 212.869 64.619 19.786 17.619 Bottom of Spent Fuel Assembly at full uplift 211.069 62.819 17.985 15.819 Top of Assembly Lying on Bail Handles 210.623 62.373 17.540 15.373 Top of Assembly Bail Handle in Spent Fuel Pool when assembly is in racks 210.333 62.083 17.250 15.083 Top of Vessel Flange 210.126 61.876 17.043 14.876 Top of Assembly Upper Tie Plate in Spent Fuel Pool 209.898 61.648 16.814 14.648 Top of Fuel Rod Plenum in Spent Fuel Rack 209.077 60.827 15.994 13.827 Top of Active Fuel in Spent Fuel Rack 196.125 47.875 3.042 0.875 Seating Surface in Racks 195.945 47.695 2.861 0.695 Bottom of Assembly Seated in Racks 195.250 47.000 2.167 0.000 Bottom of Spent Fuel Pool 193.083 44.833 0.000 -2.167 Reactor Instrument Zero 179.969 31.719 -13.115 -15.281 Fuel Channel Top Surface 148.250 0.000 -44.833 -47.000 Reactor Vessel Zero 21.473 Coverage (with SPF Cooling) over Assembly Lying on Bails (in SFP) 23.465 Coverage (with SPF Cooling) over Active Fuel (in SFP) 22.644 Coverage (with SPF Cooling) over Plenum Airspace (in SFP) 32.900 Drop Distance over Reactor Well [Less than GESTAR 34 ft. drop assumption]

2.743 Drop Distance over Spent Fuel Racks 52.573 Coverage (with SPF Cooling) over Fuel Channel Top Surface (In Reactor Vessel) 25

Appendix 5 PBAPS Fuel Handling Accident Assessment of Limiting Event (Water Coverage)

References:

(1) Fuel Assembly Dimensions (Inches) Based on GE14 Fuel per GE DWG No. 107E1 593, Rev. 1 176.14 Maximum Fuel Assembly Length 5.96 Minimum Bail Handle Length 1.065 Upper Tie Plate Thickness 1.663 Scaled Box Expansion Spring Length 0.015 Scaled Top clad Thickness 167.437 Length from Bottom of Assembly to top of Plenum (calculated) 18.55 Length from top of Bail to Top of Active Fuel (Part 2 from Table) 157.59 Length from Bottom of Assembly to top of Active Fuel 2.165 Distance from Bottom of assembly to estimated seating surface 5.348 Assembly thickness, lying on its side (2) Drawing S-232, Rev. 5, "Reactor Building Skimmer Surge Tank Sections and Details" (3) Drawing M-352, Sheet 2, Rev. 61, Sheet 4, Rev. 56, "P&ID Nuclear Boller Vessel Instrumentation" (4) Calculation ME-325, Rev. 0, "Calculate the Maximum Distance through which a Fuel Assembly Could Conceivably Fall at PBAPS from the Refueling Platform Grapple to Fuel Assemblies In the Reactor Vessel" (5)Technical Specifications 3.7.7 and 3.9.6 26

Appendix 6 PBAPS Suppression Pool pH Information Appendix 6 PBAPS Suppression Pool pH Information The calculation of suppression pool pH is based on the methodology developed for the equivalent calculation done for the Grand Gulf Nuclear Station, Unit 1 as revised December 2000. The calculation formulas developed in the Exelon documents are accepted without independent verification. The accuracy of translation of the equations in these documents into spreadsheet cell formulas was verified by duplicating the Grand Gulf calculation.

Injection of sodium pentaborate solution by the Standby Liquid Control System is a required function in order to control post-LOCA pH in the suppression pool, and prevent iodine re-evolution. Based on the worst-case beginning of cycle condition, injection should be completed within about 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> (conservatively set at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) after the start of the DBA-LOCA. Therefore, manual initiation is acceptable. Manual initiation of SLC is expected early in a DBA-LOCA as a result of emergency operating procedures and severe accident guidelines, particularly for events resulting in fuel damage that would be consistent with AST source terms.

Acceptance Criteria: Per the guidance of Appendix A of Regulatory Guide 1.183, the Suppression Pool pH should be controlled at values of 7 or greater following loss of coolant accidents.

AssumptionslEngineering Judgments

• The Suppression Pool is assumed to be well mixed so that the pH at any time can be represented by a single value.

• Cable parameters includes the exposed termination length of what is in a raceway. As a conservative estimate of the cable lengths in free air, an additional 5% of the raceway's totals were assumed to be in free air. A 10% contingency on the cable surface is also included.

Radiolysis of surface coatings on the steel and concrete surfaces in the Drywell and Containment would not be significant contributors, since the coatings utilize non-chlorinated polymers.

Temperature Suppression Pool temperatures are taken from UFSAR, Rev.15 Figure 14.6.12A. Since this revised curve extends only to 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, the older UFSAR, Rev 14 curve, Figure 14.6-12 was used to extend the data to 278 hours0.00322 days <br />0.0772 hours <br />4.596561e-4 weeks <br />1.05779e-4 months <br /> and extrapolate to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />. The older curve gives slightly lower temperatures in the area of overlap and is therefore conservative (lower temperatures give higher calculated pH values).

Extrapolation of the semi-log plot is acceptable since the calculated pH is rather insensitive to temperature. At 30 days (720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) it requires a 36 OF increase to reduce the pH by 0.1.

Sodium Pentaborate mass in SBLC Tank Per Technical Specification SR 3.1.7.7, the minimum B-10 stored in the SLC tank is 162.7 lbs. In order to prepare this calculation, total boron is needed. The highest vendor supplied enrichment is 63.5 atom % B-10. For this calculation, 65 atom % B-10 enrichment is assumed. Since B-10 has an atomic weight of 10.01, this gives 7373 gram-atoms of B-10 and 11,342 gram atoms of total boron.

Since the formula of Sodium Pentaborate is Na2 B10 O16 F110H 2 O, there are 1134 gram-mols of the pentaborate in the SBLC Tank. This calculation is performed on the pH calculation spread sheet at the top of columns U -Y.

27

Appendix 6 PBAPS Suppression Pool pH Information Suppression Pool Volume The limiting Tech. Spec. volume is 122,900 cu. ft. from Tech. Spec. Bases B3.6.2.2.

Hydriodic Acid Production Iodine, accompanied by Cesium, is released during the Gap Release and Early In-Vessel Release phases.

The following equation, valid during the Early Vessel Release Phase, includes the release during the Gap Release Phase.

Iodine and cesium core inventories are calculated for both beginning and end of cycle (BOC and EOC) conditions. Since EOC conditions result in increased inventory of both acidic (iodine) and basic (cesium) compounds, pH values are calculated for both conditions. For conservatism, the EOC radiation doses are used for the BOC calculation.

The hydriodic acid concentration is governed by the following equation:

[HI](t) = ml / (120

• VpOOL) * [t - (0.5 + tgap)] + ml/ (400

• VPOOL) [Equation I1 where:

[HI](t) = concentration of Hydriodic Acid at time t (molesAiter) ml = core iodine inventory (gram-moles)

VPOOL= Suppression Pool volume (liters) t = time after start of accident (hrs) (includes tg, + Gap Release [0.5 hrs] + Early In-Vessel Release [1.5 hrs] durations for a t,,. = 2.0336 hrs) tgap = time of onset of gap release = 121 seconds = 0.0336 hrs

t. = 2.0336 hrs = end of Early In-Vessel Release Nitric Acid Production Nitric Acid is produced by radiolysis of the water in the Suppression Pool with a G value of 0.007 molecules HNO 3 / 100 eV absorbed dose or 7.3E-6 g moles / megarad- liter.

The nitric acid concentration is governed by the following equation:

[HNO 3 1(t) = 7.3E-6

• D(t)pwl [Equation 2]

where:

[HNO3 ](t) = nitric acid concentration at time t (moles/liter)

D(t)pol = Total accumulated dose in Suppression Pool at time t (megarad) 28

Appendix 6 PBAPS Suppression Pool pH Information Hydrochloric Acid Production Hydrochloric Acid is produced by radiolysis of chlorinated polymer cable jacketing. Radiolysis of surface coatings on the steel and concrete surfaces in the Drywell and Containment would not be significant contributors, since the coatings utilize non-chlorinated polymers.

The calculation of the resulting concentration in the Suppression Pool is based on the equations in the GGNS Calculation. These equations are in turn based on the following G value for HCO production in Hypalon chlorinated polymer.

GHCI = 2.115 molecules/lOOeV = 3.512E-20 g moles HCI / MeV The hydrochloric acid concentration is governed by the following equations:

Doses from beta and gamma radiation are calculated separately.

[HCl]p(t) = GHCI / VPOOL * (Stray / 2 + Sfa) / pair

• Dp(t) [Equation 31 where the effective cable surface area for J dose is:

Stray / 2 + Sfa = 7c

• Do * (Ltry / 2 + Lfa)

[HCl]a(t) = GHCI / VPOOL * (Stray + Sfa) * (1- e -pXair*rX) / ai,

• (I1- e - XHypalon

• th)* Dt) [Equation 4]

where:

Stray + Sfa=* Do * (ray + Lfa)

[HCl]p(t) = HC1 concentration from Beta radiation at time t (g moles/liter)

[HClI](t) = HCI concentration from Gamma radiation at time t (g moles/liter)

Do = cable diameter (cm)

L4,ay = cable length in trays (raceways) (cm)

Lra = cable length in free air (cm) f air = linear beta absorption coefficient in air (1/cm)

F~air= linear gamma absorption coefficient in air (1/cm) rX = gamma free path (cm)

Jt) Hypalon = linear gamma absorption coefficient in Hypalon ( 1/cm) th = Hypalon jacket thickness (cm)

D>(t) = accumulated beta dose per unit volume at time t (MeV/cm 3 )

Dy(t) = accumulated gamma dose per unit volume at time t (MeV/cm 3 )

GHCI= 3.512E-20 (g moles HCO / MeV)

VPOOL = Suppression Pool volume (Liters)

St = Cable surface area in trays (cm2)

Sfa = Cable surface area in free air (cm2 )

29

Appendix 6 PBAPS Suppression Pool pH Information Cesium Hydroxide Production Cesium, accompanied by iodine, is released during the Gap Release and Early In-Vessel Release phases. The following equation, valid during the Early Vessel Release Phase, includes the release during the Gap Release Phase.

Iodine and cesium core inventories are calculated for both beginning and end of cycle (BOC and EOC) conditions. Since EOC conditions result in increased inventory of both acidic (iodine) and basic (cesium) compounds, pH values are calculated for both conditions. For conservatism, the EOC radiation doses are used for the BOC calculation.

The cesium hydroxide concentration is governed by the following equation:

[CsOH](t) = (0.4

• mc,- 0.475

• in) / 3

• VPOOL) * [t - (0-5 + tgap)]

+( 0.05

• mc, - 0.0475

• mI) / VPOOL [Equation 5]

where:

[CsOH](t) = concentration of Cesium Hydroxide at time t (g moles/liter) ml = core Iodine inventory (grain-moles) mc, =core Cesium inventory (gram-moles)

VPOOL = Suppression Pool volume (liters) t = time after start of accident (hrs) (includes tgap + Gap Release [0.5 hrs] + Early In-Vessel Release [1.5 hrs] durations for a tnax = 2.0336 hrs) [per, Table 4, page 1.183-15]

tgap= time of onset of gap release = 121 seconds = 0.0336 hrs tmax = 2.0336 hrs = end of Early In-Vessel Release 30

I Appendix 7 PBAPS MSLB Information Appendix 7 Main Steam Line Break (MSLB) Accident Information:

No or minimal fuel damage is expected for the limiting MSLB. Iodine concentrations used were 0.2 pCi/g and 4.0 pCi/g per RG 1.183 in determining the consequences of the main steam line break. All of the radioactivity in the released coolant is assumed to be released to the atmosphere instantaneously as a ground-level release. No credit is taken for plateout, holdup, or dilution within facility buildings.

Cloud Volumes, Masses, and Control Room Intake Transit Times The cloud is assumed to consist of the initial steam blowdown and that portion of the liquid reactor coolant release that flashes to steam. The flashing fraction (FF) is derived as follows:

FF x (steam enthalpy at 212 F) + (1-FE) x (liquid enthalpy at 212 F) =

(liquid enthalpy at temperature of steam at reactor vessel outlet)

A 548 F vessel outlet temperature is used, with liquid enthalpy of 546.9 BTU/lb.

At 212 F, a steam enthalpy of 1150.5 BTUAb and a liquid enthalpy of 180.17 BTU/lb are used (these enthalpies are taken from the ASME Steam Tables).

Substituting, FF = (546.9 - 180.17) / (1150.5 - 180.17) = .378 For conservatism, a value of .40 or 40% is used below.

Mass of water carrying activity into the cloud is calculated as the sum of the fraction of water in the steam and the liquid blowdown:

The mass steam released = 25,800 lb The mass liquid water released = 165,120 lb Flashing fraction for calculating cloud volume = 40%

The mass water contained in steam released = (25,800 lb)

• 2%

= 516 lb The mass of water carrying activity into the cloud = 516 + 165,120 lb

= 165,636 lb

= (165,636 lb)(453.59 glib)

= 7.5131E7 g The mass of steam in the cloud = (25,800 - 516) + 40%* 165,120 lb

=25,284 + 66,048

= 91,332 lb The release is assumed to be a hemisphere with a uniform concentration. The cloud dimensions (based on 91,332 lb of steam at 14.7 psi and 212 'F, vg = 26.799 ft3 fAb) are calculated as follows:

Volume = (91,332 lb)(26.799 ft 3 /lb)

= 2,447,600 ft3

= (2,447,600 ft3 )/(35.3 ft3 /m3 )

= 69,337 m3 The volume of a hemisphere is ixd3 /12. Thus, the diameter of the hemispherical cloud is 64.2 meters.

31

Appendix 7 PBAPS MSLB Information The period of time required for the cloud to pass over the control room intake, assuming a wind speed of 1 m/s is 64.2 s (=(64.2 m)/(1 W/s)).

Therefore, at a wind speed of 1 m/s, the base of the hemispherical cloud will pass over the control room intake in 64.2 seconds.

Dispersion for Offsite Dose Assessment The following formulation was used for Offsite Dose X/Q assessment, with F Pasquill Stability and a 1 m/sec wind speed.

X 0.0133 Q a3yu where ay = horizontal standard deviation of the plume (meters) u = wind velocity(meters/second)

As calculated by PAVAN, at the 1040 meter EAB distance (Ty is 38.3, and at the 7300 meter LPZ distance cYy is 222.6. The resulting EAB and LPZ X/Qs are 3.47E-04 and 5.97E-05 sec/M3 ,

respectively.

Release Isotopics and Quantification The iodine isotopic distribution previously stated was used. The concentrations of this mix are adjusted to I- 131 equivalence, using the inhalation Committed Effective Dose Equivalent (CEDE)

Dose Conversion Factors (DCFs) from Federal Guidance Report No. 11. This is a more conservative set of DCF assumptions for Control Room and off-site dose calculation than the use of ICRP 2 DCFs.

It is also more conservative for these calculations than use of R. G. 1.109 or Federal Guidance Report No. 12 DCFs.

This 1-131 equivalent mix is adjusted to the activity yielding the two design basis MSLB accident reactor coolant activities of 0.2 ttCi/cc and 4.0 I.Ci/cc. The released activities are these concentrations times the 7.5 lE+07 grams of water carrying activity released, with the assumption that TS activities are based on laboratory temperature and pressure conditions.

The PBAPS UFSAR Section 14.6.5.2.1 provides the following design basis concentrations of significant radionuclides contained in the coolant:

Iodine Isotope Activity (VCi/cc) 1-131 0.17 1-132 1.02 1-133 1.04 1-134 1.47 1-135 1.30 32

Appendix 7 PBAPS MSLB Information For the Noble Gases, the isotopic distribution is given below. The released activities are these concentrations times the 25,800 lb mass of steam released, converted to 1.17E+07 grams using the 453.59 g/lb conversion factor.

The MSLB Power Rerate Calculation provides the following Noble Gas concentrations for potentially significant radionuclides contained in the coolant:

Noble Gas Concentration Isotope uCi/g Kr-83M 1.92E-03 Kr-85M 3.44E-03 Kr-85 1.13E-05 Kr-87 1.13E-02 Kr-88 1.13E-02 Kr-89 7.33E-02 Xe-131M 8.46E-06 Xe-133M 1.63E-04 Xe-133 4.62E-03 Xe-135M 1.47E-02 Xe-135 1.24E-02 Xe-137 8.46E-02 Xe-138 5.02E-02 33

Appendix 8 Differences Between Existing Requirements and the Proposed Change Tech Spec Requirement Existing Post-AST Primary Containment Leak 0.5% per day 0.7% per day Rate (La)

MSIV Leakage Limit 11.5 scfh per line Acceptable values to be re-46 scfh total for 4 lines determined based on results of discussion with NRC reviewer and re-calculation as required.

This will be covered in a future response.

MCREV Charcoal Efficiency 90% 90%

Subject to change based on MSIV leakage re-analysis MCREV Charcoal 5% 5%

Penetration Test Acceptance Subject to change based on Criteria MSIV leakage re-analysis SGTS Charcoal Efficiency 90% Not credited SGTS HEPA Efficiency 99% Not credited SGTS Charcoal Penetration 5% Not required Test SGTS Heater ST Required Not required SGTS Availability During fuel movement Not required Secondary Containment Required during fuel NOT required during fuel Integrity movement movement Secondary Containment 2 minutes 15 minutes Drawdown Time Credit for SLC as a method Not required Required as a new design of pH control function to maintain suppression pool pH >7 during a LOCA

• MCREV will be credited only for the LOCA.

• Requirements for SGTS penetration testing being removed from TS. Will still require delta-P testing to ensure adequate flow to main stack.

• Since SGTS charcoal not credited, heater ST no longer required. However, Exelon does not plan to remove filter media or heaters from the system.

34

ATTACHMENT 2 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 Supplement to License Amendment Request for "PBAPS Alternative Source Term Implementation",

Markup of Technical Specification Bases Pages (Forinformation only)

UNITS 2 & 3 PBAPS Units 2 and 3 Technical Specification Bases Markup Inserts

PBAPS Units 2 and 3 Technical Specification Bases Markup Inserts INSERT A (pa. B 3.6-291 Total leakage through all four main steam lines must be < 250 scfh, and < 100 scfh for any one steam line, when tested at > 25 psig. The analysis in Reference 1 is based on treatment of MSIV leakage as secondary containment bypass leakage, independent of the primary to secondary containment leakage analyzed at La. The Frequency is in accordance with the Primary Containment Leakage Rate Testing Program.

INSERT B {nQ. B 3.1-391 The SLC System is also used to maintain suppression pool pH at or above 7 following a loss of coolant accident (LOCA) involving significant fission product releases. Maintaining suppression pool pH levels at or above 7 following an accident ensures that iodine will be retained in the suppression pool water.

INSERT C {pa. B 3.1-411 In MODES 1,2, and 3, the SLC System must be OPERABLE to ensure that offsite doses remain within 10 CFR 50.67 (Ref. 3) limits following a LOCA involving significant fission product releases. The SLC System is designed to maintain suppression pool pH at or above 7 following a LOCA to ensure that iodine will be retained in the suppression pool water.

INSERT D f{n. B 3.3-1561

. Both channels are also required to be OPERABLE in MODES 1,2, and 3, since the SLC System is also designed to maintain suppression pool pH above 7 following a LOCA to ensure that iodine will be retained in the suppression pool water. These INSERT E fpq. B 3.6-731 The function of the secondary containment is to receive fission products that may leak from primary containment or from systems in secondary containment following a Design Basis Accident (DBA) and, in conjunction with the Standby Gas Treatment System (SGT) and closure of certain valves whose lines penetrate the secondary containment, to provide for elevated release through the Main Stack.

INSERT F fpp. B 3.6-761 The SGT System exhausts the secondary containment atmosphere to the environment through the elevated release point provided by the Main Stack.

To ensure that this exhaust pathway is used, SR 3.6.4.1.3

V INSERT G {(p. B 3.6-851 The primary function of the SGT System is to ensure that radioactive materials that leak from primary containment into the secondary containment following a Design Basis Accident (DBA) are discharged through the elevated release provided by the Main Stack.

INSERT H {pp. B 3.6-851 These filters are not credited in any DBA analysis.

INSERT I {pp. B 3.6-861 The design basis for the SGT System is to mitigate the consequences of a loss of coolant accident by providing a controlled, elevated release path. The SGT system also provides this function for OPDRVs. For all events where required, the SGT System automatically initiates to reduce, via an elevated release, the consequences of radioactive material released to the environment.

The HEPA filter and charcoal adsorber provided in the SGT System are not credited for any DBA analysis.

INSERT J (pa. B 3.6-901 The only credited safety function of the SGT System is to provide a secondary containment vacuum sufficient to assure that discharges from the secondary containment will be through the Main Stack. The VFTP test 5.5.7.d. provides verification that the pressure drop across the combined HEPA filters, the prefifters, and the charcoal adsorbers is acceptable. SR 3.6.4.1.3 and SR 3.6.4.1.4 provide assurance that sufficient vacuum in the secondary containment is established with the time period as used in the DBA LOCA analysis.

INSERT K {pg. B 3.7-161 Additionally, the MCREV System is designed to maintain the control room environment for a 30-day occupancy after a DBA without exceeding 5 rem TEDE.

INSERT L {pa. B 3.7-161 The MCREV System is credited as operating following a loss of coolant accident. The MCREV System is not credited in the analysis of the fuel handling accident, the main steam line break, or the control rod drop accident,

INSERT M fpq B 3.6-74)

Secondary containment is only required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

INSERT N fpa B 3.6-871 The SGT System is only required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

INSERT P f{pa B 3.6-791 SCIVs are only required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

INSERT Q f{p B 3.8-401 involving recently irradiated fuel. With respect to moving irradiated fuel assemblies, AC electrical power is only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

INSERT R fPG B 3.8-42.43,72,73,74.94, and 951 involving recently irradiated fuel INSERT S f{P B 3.8-941 With respect to moving irradiated fuel assemblies, AC and DC electrical power are only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

INSERT T fpa B 3.8-741 With respect to moving irradiated fuel assemblies, DC electrical power is only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

INSERT U fpo B 3.6-75. 3.6-82, 3.6-88, 3.6-89. 3.7-18,3.7-191

, since the movement of recently irradiated fuel can only be performed in MODES 4 and 5.

INSERT V fpx B 3.8-44,741 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply since the movement of recently irradiated fuel can only be performed in MODES 4 and 5.

INSERT W fpq B 3.3-1741 The Functions are only required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

INSERT X {pa B 3.3-1821 The MCREV System is only required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

INSERT Y fpg B 3.1-401 The sodium pentaborate solution in the SLC System is also used, post-LOCA, to maintain ECCS fluid pH above 7. The system parameters used inthe calculation are the Boron-10 minimum mass of 162.7 Ibm, and an upper bound Boron-1 0 enrichment of 65%.

INSERT Z {pq B 3.7-171 The MCREV System is only required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

INSERT AA {pg B 3.8-22 3.8-38.3.8-701 (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

ATTACHMENT 3 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 Supplement to License Amendment Request for "PBAPS Alternative Source Term Implementation",

Camera-ready Technical Specification Bases Pages (For information only)

UNITS 2 & 3 B 3.8-40 B 3.8-74 B 3.8-94

AC Sources-Shutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources-Shutdown BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC Sources-Operating."

APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 SAFETY ANALYSES and 5 and during movement of recently irradiated fuel assemblies in secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate AC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident involving recently irradiated fuel. With respect to moving irradiated fuel assemblies, AC electrical power is only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

In general, when the unit is shut down the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5. Worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and corresponding stresses result in the probabilities of occurrences significantly reduced or eliminated, and minimal consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCO for required systems.

During MODES 1, 2, and 3, various deviations from the analysis assumptions and design requirements are allowed within the ACTIONS. This allowance is in recognition that (continued)

PBAPS UNIT 2 B 3.8-40 Revision No.

DC Sources-Shutdown B 3.8.5 BASES APPLICABILITY b. Required features needed to mitigate a fuel handling (continued) accident involving recently irradiated fuel are available;

c. Required features necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

With respect to moving irradiated fuel assemblies, DC electrical power is only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

The DC electrical power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.4.

ACTIONS The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply since the movement of recently irradiated fuel can only be performed in MODES 4 and 5.

A.1. A.2.1. A.2.2. A.2.3. and A.2.4 If more than one DC distribution subsystem is required according to LCO 3.8.8, the DC electrical power subsystems remaining OPERABLE with one or more DC electrical power subsystems inoperable may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, recently irradiated fuel movement, and operations with a potential for draining the reactor vessel.

By allowance of the option to declare required features inoperable with associated DC electrical power subsystems inoperable, appropriate restrictions are implemented in accordance with the affected system LCOs' ACTIONS. However, in many instances, this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement of recently irradiated fuel assemblies in secondary containment, and any activities that could result in inadvertent draining of the reactor vessel).

(continued)

PBAPS UNIT 2 B 3.8-74 Revision No.

Distribution Systems-Shutdown B 3.8.8 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Distribution Systems-Shutdown BASES BACKGROUND A description of the AC and DC electrical power distribution system is provided in the Bases for LCO 3.8.7, "Distribution Systems-Operating."

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 14 (Ref. 1), assume Engineered Safety Feature (ESF) systems are OPERABLE. The AC and DC electrical power distribution systems are designed to-provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System, and containment design limits are not exceeded.

The OPERABILITY of the AC and DC electrical power distribution system is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum AC and DC electrical power sources and associated power distribution subsystems during MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident involving recently irradiated fuel. I With respect to moving irradiated fuel assemblies, AC and DC electrical power are only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

The AC and DC electrical power distribution systems satisfy Criterion 3 of the NRC Policy Statement.

(continued)

PBAPS UNIT 2 B 3.8-94 Revision No.

AC Sources-Shutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources-Shutdown BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC Sources-Operating."

APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 SAFETY ANALYSES and 5 and during movement of recently irradiated fuel assemblies in secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate AC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident involving recently irradiated fuel. With respect to moving irradiated fuel assemblies, AC electrical power is only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

In general, when the unit is shut down the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5. Worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and corresponding stresses result in the probabilities of occurrences significantly reduced or eliminated, and minimal consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCO for required systems.

During MODES 1, 2, and 3, various deviations from the analysis assumptions and design requirements are allowed within the ACTIONS. This allowance is in recognition that (continued)

PBAPS UNIT 3 B 3.8-40 Revision No.

DC Sources-Shutdown B 3.8.5 BASES APPLICABILITY b. Required features needed to mitigate a fuel handling (continued) accident involving recently irradiated fuel are available;

c. Required features necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

With respect to moving irradiated fuel assemblies, DC electrical power is only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

The DC electrical power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.4.

ACTIONS The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply since the movement of recently irradiated fuel can only be performed in MODES 4 and 5.

A.1. A.2.1. A.2.2. A.2.3. and A.2.4 If more than one DC distribution subsystem is required according to LCO 3.8.8, the DC electrical power subsystems remaining OPERABLE with one or more DC electrical power subsystems inoperable may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, recently irradiated fuel movement, and operations with a potential for draining the reactor vessel.

By allowance of the option to declare required features inoperable with associated DC electrical power subsystems inoperable, appropriate restrictions are implemented in accordance with the affected system LCOs' ACTIONS. However, in many instances, this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement of recently irradiated fuel assemblies in secondary containment, and any activities that could result in inadvertent draining of the reactor vessel).

(continued)

PBAPS UNIT 3 B 3.8-74 Revision No.

Distribution Systems-Shutdown B 3.8.8 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Distribution Systems-Shutdown BASES BACKGROUND A description of the AC and DC electrical power distribution system is provided in the Bases for LCO 3.8.7, "Distribution Systems-Operating."

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 14 (Ref. 1), assume Engineered Safety Feature (ESF) systems are OPERABLE. The AC and DC electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System, and containment design limits are not exceeded.

The OPERABILITY of the AC and DC electrical power distribution system is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum AC and DC electrical power sources and associated power distribution subsystems during MODES 4 and 5 and during movement of recently irradiated I fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident involving recently irradiated fuel. I With respect to moving irradiated fuel assemblies, AC and DC electrical power are only required to mitigate fuel handling accidents involving recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

The AC and DC electrical power distribution systems satisfy Criterion 3 of the NRC Policy Statement.

(continued)

PBAPS UNIT 3 B 3.8-94 Revision No.