Text

Supplement to the Request for License Amendments Related to Application of Alternative Source Term, Dated July 14, 2003
	ML041270056
Person / Time
Site:	Peach Bottom
Issue date:	04/23/2004
From:	Gallagher M; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RG-1.183
	Download: ML041270056 (77)
	v • d • e

Exek1ensm Exelon Nuclear www.exeloncorp.com Nuclear 200 Exelon Way Kennett Square, PA 19348 10 CFR 50.90 April 23, 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, dated July 14, 2003

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 (3)

Letter from M. P. Gallagher (Exelon Generation Company, LLC) to US NRC, dated March 15, 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. In the Reference (3) letter, Exelon provided a partial response to the request for additional information. Attachment 1 to this supplemental letter provides the response to the remaining questions associated with the request for additional information. Attachment 2 to this supplemental letter provides the revised TS Page Markups and TS Markup Inserts pages. to this supplemental letter provides the revised TS Bases inserts. Attachment 4 to this supplemental letter provides the revised Camera-ready TS pages. Attachment 5 to this supplement provides the revised Camera-ready TS Bases pages. Attachment 6 provides the Regulatory Guide 1.183 Comparison Table Revision. Attachment 7 provides the Post-Accident Vital Area Access Considerations Table Revision from the original submittal. Attachment 8 provides the LOCA Radiological Consequences Analysis Revision from the original submittal.

X Co0k

Supplement to the Request for License Amendments Related to Application of Alternative Source Term April 23, 2004 Page 2 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.

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 04 -1 3 o04 Michael P. Gallagher Director, Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments:

1. Exelon Response to the Request for Additional Information

2. Revised TS Pages Markups and TS Markup Inserts pages

3. Revised TS Bases Inserts

4. Revised Camera-ready TS pages

5. Revised Camera-ready TS Bases pages

6. Regulatory Guide 1.183 Comparison Table Revision

7. Post-Accident Vital Area Access Considerations Table Revision

8. LOCA Radiological Consequences Analysis Revision cc:

H. J. Miller, Administrator, Region I, USNRC C. W. Smith, USNRC Senior Resident Inspector, PBAPS G. F. 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 April23, 2004 Page 1 of 15 REQUEST FOR ADDITIONAL INFORMATION PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 PROPOSED USE OF ALTERNATIVE SOURCE TERM (AST) METHODOLOGY The following questions remain from the January 16, 2004 NRC letter regarding the Peach Bottom Alternative Source Term RAI: 8, 11, 12, 13, 14, 19, 20, 21, 22, 23, 26, 27, and 28 Question #

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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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

The 0-2 hour Offgas Stack X/ 0 value of 2.72E-06 sec/m3 was assigned to represent the Control Room distance of 209 meters. This value was predicted by the NRC model PAVAN, and was calculated actually for a distance of 500 meters. In accordance with Regulatory Guide 1.194 (formerly DG-1111), Section 3.2.2, the PAVAN model was executed, in addition to running ARCON96 since the Control Room is relatively close to the base of the tall Offgas Stack, and the ARCON96 model had produced negligibly small x/Q values at the Control Room intake distance. Pursuant to Regulatory Guide 1.194, several distances in addition to the actual 209 m distance were modeled by PAVAN until the maximum 0-2 hour x/Q was determined, as predicted to occur at 500 meters. This maximum 2/a was then assigned as the 0-2 hour Control Room x/0 value. As indicated by Regulatory Guide 1.194, this conservative procedure was performed to account for the possible diurnal wind direction changes, meander, or stagnation.

In the current submittal, the worst two hours were identified only for the EAB.

The 0.5-hr fumigation condition was applied to the beginning of the 2-hour period, rather than the end. This has been corrected in new results submitted.

For LPZ and the Control Room (CR), the cumulative dose is typically dominated by later periods, in which case applying the worst x/a s to short maximum release periods would-not be necessary.

We also note that the RADTRAD Version 3.03 deleted the Control Room and LPZ "worst 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months " period identification in the output. Nonetheless, since the LOCA analysis was modified, maximum (0.5 hr and 2 hr) X/ 0s has been applied to the maximum release periods for the EAB, LPZ, and CR unless otherwise justified.

Supplement to PBAPS AST RAI April 23, 2004 Page 2 of 15

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 staffs review.

b.

How is it different from the previously reviewed method described in PECO Energy Company's Calculation PM-1013, "Minimum Containment Pressure Calculation," Revision 3, February 2000?

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

(a)

Exelon is providing the following information from the calculation to address the issues identified above. This represents the relevant information presented in the calculation relative to containment leakage. Arrangements can be made for the review of Exelon Nuclear Design Analysis PM-1013, "Minimum Containment Pressure Available", Revision 5, if required.

(b)(c)(d) The previous containment leakage assumptions considered the Technical Specification 0.5% per day leakage as remaining constant over time, throughout the entire event. The constant nitrogen leak rate from the containment was (Equation 2 from PM-1013, Revision 3):

144*(Pi +P )*L, *V 24*3600*Ra *(T +TO) where m = nitrogen leakage mass flow, Ibm/sec P. = nitrogen leakage pressure, psig Palm = atmospheric pressure, psia La = nitrogen volumetric leakage rate, % per day V = containment free volume, ft3 Ra = nitrogen ideal gas constant, (ft-lbf)/(lbm-0R)

T, = nitrogen leakage temperature, 'F To = temperature conversion, 'F to 'R Values were selected to maximize the leakage (i.e., maximum PI and minimum To). Only nitrogen is assumed to be expelled during the event.

Supplement to PBAPS AST RAI April 23, 2004 Page 3 of 15 The revised containment leakage methodology is based on the proposed PBAPS Technical Specification limit of 0.7% weight per day for general containment leakage at the test pressure of 49.1 psig, plus the proposed PBAPS Technical Specification limit of 158 scfh or less1 for total MSIV leakage at a test pressure of 25 psig, plus the current PBAPS Technical Specification limit of 9000 scc per minute for airlock seal leakage at a test pressure of 49.1 psig. MSIV and airlock leakages are converted to the equivalent percent weight per day as follows (Assumption 5.J.x from PM-1013, Revision 5):

24 *144

QYPalm A

P LA{sO RN -(60 + To)

MaO AP'IX L

_ilck-24

60

144

Q,,iJO¢*

Pa~...

K airlock 30.483

RN - (60 + TO)

Ma J

where L,xsl= MSIV leak rate, % weight per day at APref Lasrloi = airlock leak rate, % weight per day at APf Qjvs 1 = MSIV leak rate, scfh Qairlock = airlock leak rate, sccm P

= atmospheric pressure, psia RN = gas constant for Nitrogen Ma = initial containment Nitrogen mass, Ibm AP,4 = reference containment differential pressure, 49.1 psid AP.es, = test reference differential pressure, psid To = temperature conversion, 'F to 'R Using the PBAPS Technical Specification leakage limits and the above expressions, LMStV and Laidock are calculated as 2.31% and 0.20% per day, respectively.

Combining with the 0.7% per day general containment leakage, a total containment leak rate of 3.21% per day is estimated. Use of the assumed total containment leak rate of 3.21% per day, i.e., with each component at its maximum value, assumed to occur at the same time, is conservative.

With these containment leakage values normalized to the reference containment differential pressure, the % weight leakage at any given time is assumed to be a function of containment pressure:

L(t) = Ma. (Lcontainment + LAjsIv + Lairlck)

At Apd(t) ai24 3600 AJpef The License Amendment Request proposed an MSIV leakage of 174 scfh. Currently, the MSIV leakage rate assumed for AST purposes is under discussion with NRC staff. Therefore, a maximum value of 158 scfh has been assumed for MCPA purposes, and will be confirmed subsequent to approval of the amendment request.

Supplement to PBAPS AST RAI April 23, 2004 Page 4 of 15 where:

L(t) = total leakage during At, Ibm Ma, = initial containment Nitrogen mass, Ibm Lcontainment = containment leak rate, % weight per day at APKI LsV = MSIV leak rate, % weight per day at APref Lairioc& = airlock leak rate, % weight per day at APref At = time step size, seconds APd (t) = containment differential pressure, psid APef = reference differential pressure, 49.1 psid The turbulent (orifice) flow relationship of flow being proportional to the square root of the pressure difference was used based on a review of TID-20583, Leakage Characteristics of Steel Containment Vessels and the Analysis of Leakage Rate Determinations. Per TID-20583, to extrapolate downward from a high test pressure to a low actual pressure, the assumption of orifice flow should be used since it will result in the least change in leakage rate.

Supplement to PBAPS AST RAI April23, 2004 Page 5 of 15

12.

Previously, containment leakage was assumed to be constant at La=0.5%/day throughout the event. The containment leakage has been increased to La=0.7%/day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 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 0.56xLa=0.392%/day from 24 to 38 hours4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months and then reduced to 0.50xLa=0.350%/day, for 38 to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months . 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the MSIV leak rate is reduced to 77.2%, then to 65.4% at 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , to 59.0% at 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , to 55.5% at 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months , and finally to 50% at 157 hours0.00182 days 0.0436 hours 2.595899e-4 weeks 5.97385e-5 months 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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ?

c.

Identify the TS which controls the allowable airlock leakage rate.

RESPONSE

The leakage values stated in the question above are only used to support dose analysis and are governed by the guidance relative to containment leakage in Regulatory Guide 1.183. These leakage assumptions are not the same as those used in the determination of the MCPA in calculation PM-1013, which does not support, take input from nor provide input to the dose calculations.

(a) As described in the response to NRC RAI Question #11 above, the reference containment leakage is assumed to be 3.21% weight per day at a reference containment pressure of 49.1 psig. In addition to the 0.7% weight per day general containment leakage, this also includes the leakage from the MSIVs and personnel airlock (LMsIv and L2idk are 2.31%, 0.20% weight per day, respectively at APrf). During the analysis of the MCPA, the % weight leakage is varied only as a function of the containment pressure:

(L

~+

ailctRA Adt L(t) = MAa i onainnent +LA,,

+ Lairtsk ) *At

Ap 24 -3600 AP1 r (b) The MCPA analysis applies the leakage methodology described above out to 10,000,000 seconds (2777.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) which corresponds to the duration of the GE analysis for the DBA LOCA long term suppression pool temperature response.

Calculation PM-1013 provides documentation of this analysis through 13.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to ensure the point of peak suppression pool temperature and peak MCPA is adequately covered.

Supplement to PBAPS AST RAI April 23, 2004 Page 6 of 15 (c) Peach Bottom Technical Specification 5.5.12 requires that overall airlock leakage is s 9000 scc/min when tested at 2 Pa. This requirement is verified by Surveillance Requirement 3.6.1.2.1.

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," 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 MCPA/COPL 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

The COPL was established in reviews culminating in the previously mentioned NRC SER (Letter from B.C. Buckley, Sr., USNRC to J.A. Hutton, PECO Energy Company, August 14, 2000, "Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 - Issuance of Amendment Regarding Crediting of Containment Overpressure for Net Positive Suction Head Calculations For Emergency Core Cooling Pumps (TAC Nos. MA6291 and MA6292)"). From that SER:

'The NRC staff performed confirmatory calculations of the RHR NPSH analysis. According to our calculations, the minimum margin between the COPL and the COPR for the RHR pumps is 0.88 psig. This occurred at the peak suppression pool temperature of 205. 70F. This margin allows for minor design changes which could affect the COPR. This result is consistent with the licensee's calculations. Additionally, our calculations demonstrated that the minimum margin between the COPL and MCPR was approximately 0.42 psig (1 foot). Because of the way the COPL was defined, i.e., the COPL will be 1 foot less than the MCPA for a design basis LOCA, this minimum margin is maintained over the entire COPL curve.

and:

'For the long term following a LOCA, the staff has approved the use of the containment overpressure depicted on UFSAR Figure 5.2.16 and provided in the table above for both the RHR and core spray pumps."

Supplement to PBAPS AST RAI April 23, 2004 Page 7 of 15 Since that time, some plant changes have been made which were not considered within the original intent of a minor design change. These changes included increases in the TS allowable river water temperature from 900F to 920F, correction of decay heat errors identified by GE in SIL-636 rev.1, the formal incorporation of 2a decay heat uncertainty in the containment calculations, and the currently proposed increases in MSIV and containment leakage as part of AST. These changes had the net effect of decreasing the MCPA. The change in MCPA methodology necessary to accommodate AST proposed leakages, and its potential impact to COPL, is the very reason Exelon Nuclear has requested this NRC review.

(a) Although the COPL line was derived using a 1 foot margin to the MCPA, it is our understanding that, like the original Peach Bottom FSAR containment overpressure limit line, the NRC SER has established the COPL line itself, "depicted on UFSAR Figure 5.2.16" as the limit, rather than the maintaining of a specified margin to the MCPA. With the COPL being a fixed line, the proposed AST changes would have reduced the MCPA margin to the previous COPL from 1 foot of head (7.41 - 6.99=0.42 psid) to essentially zero (7.04 - 6.99=0.05 psid).

Consequently, a new COPL limit needs to be proposed.

If instead of COPL being a fixed line, maintaining a 1 foot of head margin to the MCPA were the case, adequate overpressure margin would still be available to satisfy the NPSH requirements of the RHR pump during the design basis LOCA.

A 1 foot of head margin to the new MCPA would produce a peak COPL of 6.62psig (7.04 - 0.42=6.62 psig), which still provides another foot of margin to the peak COPR for RHR of 6.14 psig (6.62 - 6.14=0.48 psid).

The COPL line that was proposed with the AST amendment request preserved the relative relationship between the COPR, COPL, and MCPA from the previous NRC review. With about a third of the available margin between MCPA and COPR being assigned to the COPL, the new proposed COPL line still maintains about a third of the available margin to the MCPA (0.69 foot, approx. 0.29 psid).

(b) The following table summarizes the COPL, MCPA, and COPR data provided in the attached charts, with a COPL based on a 1 foot margin to the MCPA.

Peak Peak Peak Peak MCPA-COPL-MCPA COPL*

COPR COPR COPL COPR (psig)

(psig)

(RHR, psig)

(CS, psig)

(psid)

PM-1013 r3 7.41 6.99 6.11 4.83 0.42 0.88 PM-1013 r5**

7.04 6.62 6.14 4.78 0.42 0.48

Both COPL values represent a 1 foot margin to the MCPA

- rev. 5 was prepared to address corrections in the calculation write-up of rev. 4 which was prepared for this amendment request.

The Suppression Chamber air space mixing issue will be addressed in question 14.

Supplement to PBAPS AST RAI April 23, 2004 Page 8 of 15 PBAPS MCPAAnalysis Previous COPL 8.00 7.00 6.00

" 5.00 ea.

en 2! 4.00 CC)

E C

I 3.00 C

0 2.00 1.00 0.00 0

2 4

6 8

10 Time (hours) 12

C. '

Supplement to PBAPS AST RAI April 23, 2004 Page 9 of 15 PBAPS MCPA Analysis Proposed COPL a.F

-e 2

0 as C

a.

CE C0 0

8.00 7.00 6.00 5.00 4.00 3.00 2.00 1.00 0.00 0

2 4

6 8

10 12 Time (hours)

Supplement to PBAPS AST RAI April 23, 2004 Page 10 of 15

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

The original submittal was based on steam driven exchange through the downcomers.

For consistency with the assumptions required for the core activity release (2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of no ECCS flow), no steaming is now assumed for the first two hours. The drywell/suppression chamber air space is assumed to be well-mixed thereafter due to flashing associated with core re-flood.

19.

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

RESPONSE: SLC questions identified per the NRC guidance document will be responded to under a separate cover.

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.

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.

Supplement to PBAPS AST RAI April 23, 2004 Page 1 of 15

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-31858P-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

The use of Brockmann-Bixler approach incorporated in the initial submittal is being abandoned to incorporate a well-mixed methodology to facilitate submittal review, and to credit AEB-98-03 settling velocity treatment.

Furthermore, since deposition in inboard MS piping is being credited, the design basis pipe break is assumed to be a steam line break inside containment in the vicinity of an inboard MSIV.

A single inboard MSIV failure is assumed on the broken line. The MSIV assumed to fail open is the one associated with the broken inboard line that produces the highest dose.

To account for possible containment turbulence in the vicinity of the penetration piping, the first two pipe diameters in the penetration will not be credited.

Figure 1 in this attachment, shows a single-line sketch of the four main steam lines and the isolation valves. This figure shows control volumes and break locations.

Table 1 in this attachment, shows all parameters input into the AEB-98-03 based model of the steam lines, and justifies the conservatism, including consideration of plant cool down effects.

For the analysis, leakage is assumed to be distributed evenly between the two worst steam lines. MSIV leakage limits will be 75 scfh maximum per main steam line with a total acceptance criterion of 150 scfh for all four lines.

Only steam line piping that has been seismically qualified is credited in this analysis.

PBAPS did not pursue the NEDC-31 858P based approach for seismically analyzing and crediting balance of plant equipment, such as turbine shells or the main condenser.

Supplement to PBAPS AST RAI April 23, 2004 Page 12 of 15

21.

On page 53 of Attachment I 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.

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-1465) 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 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

Reference response to question 20 above. AEB-98-03 methodology is now used to assess aerosol removal in steam line piping. Analogous treatment is used for elemental iodine deposition. No deposition is credited for organic iodine.

For aerosol settling only horizontal piping is credited, with the lower half providing the settling area. For elemental iodine deposition, all available piping and surface areas are credited.

Slug flow modeling for MSIV leakage will not be used.

22.

Page 13 of Attachment 1 of your submittal provides text that states "an initial 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 LOCA 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

The 12-hour delay would not be consistent with well-mixed modeling, and therefore, will no longer be credited.

t Supplement to PBAPS AST RAI April 23, 2004 Page 13 of 15

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-accident and reduced values after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 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

The leakage rates of 0.62 and 0.31 cfm result from correcting the measured outboard 100 scfh flow to inboard pressure conditions. This means applying a factor of (14.7/(14.7+25)), where 25 is the MSIV test pressure and 14.7 is atmospheric pressure, both in psi.

This is consistent (in a reverse direction) with the PBAPS and every other BWR approach to La management. For example with current PBAPS limits:

293,900 [cu. ft.]

0.5 [%/day]

0.01 [%] * (14.7 [psia] + 49.1 [psig]) /14.7 [psia] 124 [hr/day] = 265.7 scfh.

This would be an inboard flow rate of 61.23 cfh. This approach is consistent with 10CFR50, Appendix J and its cited ANSI/ANS Standards. No temperature adjustment is required related to leak rate.

Flow rates in inboard MSIV piping will be the same as the volumetric leak rate, which is, in effect, a mass flow rate from a constant volume. Any heating in inboard MS piping would cause expansion back into containment or the reactor vessel until pressures equalize. However, if outboard piping should be hotter than inboard piping then leakage expansion could be greater than that associated with test conditions. Therefore, for conservatism, the flow rate in outboard piping is adjusted as follows:

75 [scfh] * (550[F] + 460[R]) / (68[F] + 460 [R])

No extrapolation upward from the MSIV test pressure to Pa was deemed necessary as actual pressures only exceed the test pressure for approximately the first 6 minutes of the event.

However, based on a review of the Staff endorsed NEDC-32091 and NEDC-31858P documents, an alternative method of evaluating leak rates is now being applied. This method accounts for partial pressures of water vapor, initial containment non-condensables based on containment response to a Recirculation Suction Line Break, plus H2 from Zirconium-Water reaction. Use of this methodology for a 100 scffh MSIV leakage acceptance criteria for PBAPS results in a predicted leak rate of 0.58 cfm at containment conditions.

Supplement to PBAPS AST RAI April 23, 2004 Page 14 of 15 The proposed MSIV leakage limit is now 150 scfh total with a maximum of 75 scfh in any one line. The above method results in a proportionally lower leak rate of 0.437 cfm in the maximum line at containment conditions.

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 I states "The Technical Support Center at PBAPS is in the Unit 1 Control Room. A review of the current TID-14844-based analysis 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

Doses to personnel in the TSC have been reanalyzed using the same release modeling as used for the control room. Primary differences between the TSC and control room are improved XIQs at the TSC, and the availability of a recirculation filter, in addition to the intake filter. An unfiltered inleakage allowance of 50% of the filtered intake flow rate is analyzed. Direct shine from external sources such as airborne activity in the unshielded reactor building refuel floor and external cloud are assessed to quantify this contributor.

Table 2 (Attachment 7) provides additional detail on the qualitative and semi-quantitative assessment for the analyzed activities as described in the UFSAR Section 12.3.5.

Further discussion of quantitative and qualitative assessments of control room dose contributors is included in Attachment 6 (Table A, Item 4.2.1). Other vital areas have been reassessed using AST source terms and have been determined to be accessible consistent with their use.

The EOF is located in Coatsville, PA (approximately 30 miles away from the site).

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 scffh." 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

These TS pages are revised to reflect leakage rates that are now limited based on LOCA Dose analyses. (see Attachment 5 to this supplement).

Supplement to PBAPS AST RAI April23, 2004 Page 15 of 15

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

The original submittal was based on steam driven exchange through the downcomers.

For consistency with the assumptions required for the core activity release (2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of no ECCS flow), no steaming is assumed for the first two hours. The drywell/suppression chamber air space is assumed to be well-mixed thereafter due to flashing associated with core re-flood.

Reactor Building Wall Notes:

1. Posulated MSLB and failed MSiV Is assumed to be In shortest ine.

2. Two Pipe Diameters (24-) of the Penetration pipe following this failed MSiV Is conservatively not assumed for deposition credit due to turbulence In the Inlet vicinity. For the broken line, Inboard pipe Is from this point to the Outboard MSiV. Outboard piping continues from this point through selsmically supported pipe, out through the Turbine Stop Valve.

3. For Intact steam fines inboard piping is from the RPV through the Inboard MSiV. Outboard piping Is from this point through the seismicanfy supported piping, out through the Turbine Stop Valve.

Turbine Building Wall FIGURE 1 Reactor Pressure Vessel (RPV)

Inboard Piping Peeratio PBAPS Design Basis AST LOCA Schematic Assumes Accident Initiation by MSLB and Single Failure of Inboard MSIV Deposition Credited In Two Shortest Intact Lines Using an AEB-98-03 Well-Mixed Model Outboard I MSLB

f0$

I PBAPS Unit 3 Determination of MSL Decontamination Factors Due to Iodine Deposition I

I I

Unit 3 1

Unit 3 3

Unit 3

1 Unit 3

1_

Unit3 unit 3

1 Unit3 Unit 3 _

Inboard A I Inboard B I Inboard C I "Inboard 0 Outboard A I Outboard I Outboard C I "Outboard D d Pipn Surface Are (

Total PIpe Volume (

e87 e20 i17 115 2101 2036 T

1971 I

1e18 33

8 305 3

304 toiotlTotal Pioe Su.t-A-.s ftn'1 302) 258A I255 5

i=

1034 2031 1015 49 19w 1901 1002 970 1

796 1548

.I-----------

4 Horizontal Settllng Plpe Surface Area (fell 151.19 1

128 87 127.60 1

57.58

'8298 950 i9 774 04

'Horizontal Pine Volume (ft'll 149 1

127 1

126 1

58 1000 1

968 938 1

762 rosol Settling Velocity (mls )

1.170E-031 1.170E-03 1.170E-03 I 1.170E-03 I _

1.170E-03 l

1.170E-03 1.170E-03 I 1.170E-03 mrosol Settling Velocfty_(tth)'

1 3 839E-03 3.839E-03 1 3 839E-03 l 3 839E-03 I 1 3 839E-03 l 3 839E-03

_3 839E-03 3 839E-03 5 'Elementar Depositlon Velocity 0-24hrs (m/sec l 5.359E-06 I 5.359E-i I 5.359E-05 S

5 359E-06 I 5 359E-06 l

5 359E-06 I 5.359E-08 1 5 359E-06 I 248E-05 I 1.248E-05 I 1.248E-05 I 1.248E-05 1.248E-05 1.248E-05 I 1.248E-05 I 1.248E-05 I

.4E0 7.4E0 t 7.94E05 0

7.45E- 0 7.4SE-05 I 7_________05 I

.4E0

.4E0 n

i 110-3 I

110---

I 1.3O- 03

1. ~-i--J3 1.758E-05 I 1.758E435 I 1.758E-05 I 1.758E-05 I 1.758E-05.

I 1.758t: 0s I1.758E-05 1.7F8t-u5 4 095E-05 4 095E-05 I4 095E-05 l 4 095E-05 4.09SE-(

1 2 607E-04 1 I.,

v 0-24hrs (mlsec)!

5 919E-09 S 980E49 IS 989E-09 5.969E-09 1.3909-E8 "Oroanlc Deoostion Velocity 24-96hrs (mlsc) 1 390E-08 1.390E-08 I 1.390E-08 I 1.390E-08 560,oanic Deposition Velocitv 96-720hrs mlmseell 8 8495-08 8.849E-08 8 849E-08 8 849E-08 OrganIc Deposition Velocity o0 1 958E-C 4.561E-C 2.903E.C 0

4 561E-08 1 4 561E-08 Uncorrected Flow Rate (scfh)l 2 903E-07 2.9035E-07 2.903E-07 0

77 07 0 0000 0.7736 10.7736 "Ar 0 0000 2.903E-07 0

N/A N/A N/A 00000 00000 0.0000 0.0000 N/A I

N/A N/A NtA t RIA l

N/A l

N/A N/A N/A v I 0 0000 I 0 0000 I 0 4375 1 0 4375 Pipe Fn 0 0000 07813 144 6023 00000 1 28.2489 I 26.2489 Pipe Flow Rate 24-96 hn; (cfhl I Npe Flow Rate 24-720 hrs (dh) i iol Settling Rate Cons ant (hr")l 0.78t13 144.M023 61.7898 46 8750 1.40E+01 -4l 1.40E+01 I 1.40E+01 1.40E+01 I 1.41E+01 1.40E+01 1.40E-01 I 1.40E+01 ental Deposition Rate Constant 0-2 I'-l 1.29E.01 l 1.29E401 1.29E401 I 1.29E-01 1.29E-01 1.29E-01 1.29E.-01 1.29E-01

'Elemenbl Deaos~iton Rate Constant 24-96hr Ih Il

'Eemental Depositon Rate Constant 96-720hr (hr'ii

'Organic Deposition Rate Constant0-24hr(ht')l 009-01 3.00-01 I 301E-01 3.00E-01 3 00E-01 3.00E-01 300O-01

,4 i

I_

1 919+00 1 S91E+00 1.91E+00 1.91 E+00 1 43E-04 I 1.43E-04 1.4 1.43E-04 143E-04 I

f'il 3.34E-04 3.34E1-4 0

324E1-044 3.35E-04 f'jl I2.12E403 I2.12E403 I2.12E403 I2.13E-03 IC 1.43E-04 3.34E-04 2.12E-03 98 91%

i 43E-04 3 34E-04

'Aerosol Fitter Efficiency (0-24 hrsl)

'Aerosol Fillter E

)cncy (24-98 hfl!

'Aerosol Filter Efficiency (96-720 hn)r 000%

I 000%

99 26%

QQ NM1

=

0.00%

I__ O 00%

0 00%

0.00%

nOrim I

98 87%

99 38A%

Elemental Filter Efficienyjn2

'Elemental Filter Efficiency (24-96 hrs)

'Elemental Filter Efficiency 196-720 hrsl)

I I-I I

0 00%

0.00%

000%

0 00%i 598 2%

_ 21.73%

000% I 87.40%

_ 5340%

000%

48 32%

99 43%

99 56%

41.46%

79 43%

97 01%

zzi_

000%

I 0.00%

l 8246%

0 00%

97.78%

89 17%

0 17%

003%

000%

0.00%

97.53%

000%

I 000%

I 0.10%

I oo08%

'0 0 00%

0 77%

0.14%

0.00%

4 69%

I0 91% _

000%

0 52%

043%

3.48%

v (98-720 hrs)l

'PpeWal empeatre,0-24hr F4158 00 Sinine T.k-ff R.ll Suma rtarhd i

r--

-z

.en1lww-wlt vj.1-4 e

~

z_ _

'Pine Wall Temperature. 24-96hr fF141000 JSNRC Document AEB-98-03. 12/91998. Page

-3. MedIan Value PlpoWall Temperature, 24-98hr (Kii483.1S IUSNRC Document AEB-9843, 12/311998, Pag8 -,9 owl 7Plpa Wall Temperature. 9S 720hr (Fj1200.00 USNRC Documeant AEB-98-3, 12/911998, Page A-2, Formula 4_

Pipe Walt Tempenatune, 96720hr (K)1388.48 ClineJ.E -MSIV Lesksge Iodine Transport Anahrsis-3/2t1991_

Condenser Temperature, constant (F)'Nrt Credit NUREGtCR-6804, RADTRAD Mnul.411998 Suplement 1, 6/13999_

Condenser Tempenatur, constant (K) 'Not Creditd Cline. J.E. 8/1990tW Standard Temperature, constant (F) '88 Condenser Not CreditedJ

'MSItr Test Pressure, constant (psi ) 25 PBAPS Te=h*cal Specifictio

'"Peak P. Containment Pressure, constant (pslg) 491 to IPBAPS Technical Specificstion Atmospherlc Pressure, eonstant (psls)j 14.7

[Value uued to simulate an un mix ed containment U oume for the firs*t 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the LOCA.

IExtrapotatlon Factor, eonsbandl1.00 lMin Steam Une Break of "nest hinoard line assumed for LOCA Deposthon ere~di reduction.

=

N extrapolation from Test pressure to P. *ppheC due to sma1l differential. short bime of that d-Iferential, and obiher conservabsins Table 1: Main Steam Line Depoition Parameters-Page I of 7

no~tcrminntinn nf lnhnnurr MRlV I goik Rntp-iminn INFFlC.51R!RP anin N1766-52hp mnihndonlny Constants 68 Standard Temperature (OF) 558 Main Steam Pipe Wall Temp 0-24 hours (°F) 410 Main Steam Pipe Wall Temp 24-96 hours (°F) 200 Main Steam Pipe Wall Temp 96-720 hours (OF) 14.7 Conversion Factor (atm to psi)

I Containment Volumes 159,000 Drywell Volume (ft) 127,700 Wetwell Volume (Rft) 7,200 Reactor Vessel (R3) s ace above nominal water level vs. (GE 14,000 ft3 value) 293,900 Total Volume (f3) 8322.3663 Total Volume (mi) 1.7684 Ratio of Total Volume to Drywell Volume including RPV Containment Temperatures and Pressures per Containment Analysis for RSLB In PM-1061, RO I

I I

276 DW Temp (OF) at minimum DW-WW differential (at - 69 seconds) 131 WWTemp (°F) at minimum DW-WW differential (at - 69 seconds) 213.0 Average Bulk Temperature (CF) 46.1 DW Pressure (psia) (use for pressure vessel as well) 43.9 WW Pressure (psia)

I.

-_I 45.1 Average Bulk Pressure (psia)

I__

3.07 Average Bulk Pressure (atmospheres)

Hydrogen Contribution from Zirconium Water Reaction 764 assemblies T

{PBAPS Value) 102.00 lbs Zr/assembly j

NEDC-31858P_

7.87 cubic feet H2 per lb Zr

{NEDC-31858P) 0.20 fraction of Zr undergoing metal water reaction (NEDC-31858P) 122658.67 Total Hydrogen (f3) l 1

(Calculated PBAPS Value) 167782.42 Corrected to bulk average temperature (Calculated PBAPS Value) 0.5708827 Partial Pressure of Hydrogen (atmospheres)

( {Calculated PBAPS Value) 3.64 Total {H2, N2, H20} Pressure (atmospheres)

{Calculated PBAPS Value)

I I-I I

I I

I Inboard Leak Rate Determination per NEDC-32091, Section B.1.3, Duane Arnold Example based.

A B

C D

I I

01 0J 75 751Containment Leak Rate (scfh) {use as basis for outboard flow rate) 0 01 0.21435 0.21435 Leak Rate In %Iday IlI 0.0000 0.00001 0.4375 0.43751 Inboard Leak Flow Rate (cfm)

I I

II 0.0000 0.00001 26.2489 26.2489lInboard Leak Flow Rate (cfh) l I

i I

I I

I Note that no extrapolation from test pressure to Pa Is required based on the NEDC-31858P note that these containment conditions are essentially equivalent to test conditions.

I Table 1: Main Steam Line Deposition Parameters - Page 2 of 7

Main Steam Piping Summary 23.624 Main Steam 24 inch pipe ID

__i

_i A

B C_

D_

PBAPS Unit 2 Nodalization (Horizon Is) 296 254 254 300 Node 1 Surface Area (sq. ft.)

146 125 125 148 Node 1 Volume (cu. ft.) l 153 140 140 153 Node 2 Surface Area (sq. ft 75 69 69 75 Node 2 Volume (cu. ft.)

1794 1838 1882 1927 Node 3 Surface Area s K.)

883 905 926 948 Node 3 Volume (cu. ft.) I Nodalization (Totals) _

l l

667 616 616 671 Node I Surface Area (sq. ft.)

328 303 303 330 Node 1 Volume (cu. ft. l 153 140 140 153 Node 2 Surface Area (s5. at 75 69 69 75 Node 2 Volume (cu. ft.) I 1863 1907 1952 1997 Node 3 Surface Area (s. ft.)

917 939 961 983 Node 3 Volume (cu. ft.)

A B

C D

PBAPS Unit 3 Nodalization (Horizontals) 302 258 255 307 Node I Surface Area (sq ft.)

149 127 126 151 Node 1 Volume (cu. ft.) l 140 140 140 140 Node 2 Surface Area sq. ft.

69 69 69 69 Node 2 Volume (cu. ft.)

1891 1826 1761 1548 Node 3 Surface Area (sq ft.)

931 1

899 867 762 Node 3 Volume (cu. ft.)l Nodalization (Totals) 687 620 617 685 Node l Surface Area (sq. ft.)

338 305 304 337 Node 1 Volume (cu. ft.)

140 140 140 140 Node 2 Surface Area sq. ft.

69 69 69 69 Node 2 Volume (cu. ft.) I 1961 1896 1831 1618 Node 3 Surface Area s ft.)

965 933 901 796 Node 3 Volume (cu. ft.)]

==

l l

Table 1: Main Steam Line Deposition Parameters - Page 3 of 7

-I PBAPS Unit 3 Main Steam Line A Inner Diameter (in.)= 23.624 Horizontal Horizontal Location Horizontal Volume (ft3)

Surface Area (ft2)

Volume (ft3)

Surface Area (ft2)

Inboard TRUE 4.81 9.77 4.81 9.77 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard FALSE 100.55 204.30 0

0.00 Inboard FALSE 39.52 80.30 0

0.00 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard TRUE 83.98 170.63 83.98 170.63 Inboard TRUE 9.896 20.11 9.896 20.11 Inboard FALSE 49.15 99.86 0

0.00 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard TRUE 19.03 38.67 19.03 38.67 Penetration TRUE 68.85 139.89 68.85 139.89 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard FALSE 34.26 69.61 0

0.00 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 120.28 244.39 120.28 244.39 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 424.02 861.54 424.02 861.54 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 241.83 491.36 241.83 491.36 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 92.75 188.45 92.75 188.45 Totals 1371.886 2787.44 1148.41 2333.37 Horizontal Horizontal Total Volume Total Surface Volume Surface Area (ft_)

Area (f2)

(ft3)

(ft2)

Inboard (Node 1) 338.05 686.85 148.83 302.39 Penetration (Node 2) 68.85 139.89 68.85 139.89 Outboard (Node 3) 964.99 1960.70 930.73 1891.09 l

Totals 1371.89 2787.44 1148.41 2333.37 I

_ I I

I I

II Table 1: Main Steam Line Deposition Parameters - Page 4 of 7

lb I PBAPS Unit 3 Main Steam Line B Inner Diameter (in.)= 23.624 Horizontal Horizontal Location Horizontal Volume (ft3)

Surface Area (ft2)

Volume (ft3)

Surface Area (ft2)

Inboard TRUE 5.45 11.07 5.45 11.07 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard FALSE 100.03 203.24 0

0.00 Inboard FALSE 35.57 72.27 0

0.00 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard TRUE 47.11 95.72 47.11 95.72 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard FALSE 42.69 86.74 0

0.00 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard TRUE 22.44 45.59 22.44 45.59 Penetration TRUE 68.85 139.89 68.85 139.89 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard FALSE 34.5 70.10 0

0.00 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 131.94 268.08 131.94 268.08 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 424.02 861.54 424.02 861.54 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 210.35 427.40 210.35 427.40 Outboard TRUE 10.37 21.07 10.37 21.07 i

Outboard TRUE 80.57 163.70 80.57 163.70 Totals 1307.22 2656.05 1094.43 2223.70 Horizontal Horizontal Total Volume Total Surface Volume Surface Area (ft3)

Area (ft2)

(ft3)

(ft2)

Inboard (Node 1) 305.14 619.99 126.85 257.74 Penetration (Node 2) 68.85 139.89 68.85 139.89 Outboard (Node 3) 933.23 1896.17 898.73 1826.07 Totals 1307.22 2656.05 1094.43 2223.70 Table 1: Main Steam Line Deposition Parameters - Page 5 of 7

Il I PBAPS Unit 3 Main Steam Line C Inner Diameter (in.)= 23.624 l

=

Horizontal Horizontal Location Horizontal Volume (ft3)

Surface Area (ft2)

Volume (ft3)

Surface Area (ft?)

Inboard TRUE 5.45 11.07 5.45 11.07 l

Inboard TRUE 10.37 21.07 10.37 21.07 Inboard FALSE 100.03 203.24 0

0.00 Inboard FALSE 35.57 72.27 0

0.00 Inboard TRUE 10.37 21.07 10.37 21.07

]

Inboard TRUE 47.11 95.72 47.11 95.72 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard FALSE 42.69 86.74 0

0.00 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard TRUE 21.19 43.05 21.19 43.05 Penetration TRUE 68.85 139.89 68.85 139.89 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard FALSE 34.5 70.10 0

0.00 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 143.6 291.77 143.6 291.77 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 424.02 861.54 424.02 861.54 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 178.89 363.47 178.89 363.47 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 68.39 138.96 68.39 138.96 Totals 1273.99 2588.53 1061.20 2156.18 Horizontal Horizontal Total Volume Total Surface Volume Surface Area (ft3)

Area (fe)

(ft3)

(ft2)

Inboard (Node 1) 303.89 617.45 125.60 255.20 Penetration (Node 2) 68.85 139.89 68.85 139.89 Outboard (Node 3) 901.25 1831.19 866.75 1761.09 Totals 1273.99 2588.53 1061.20 2156.18 Table 1: Main Steam Line Deposition Parameters - Page 6 of 7

I,

PBAPS Unit 3 Main Steam Line D Inner Diameter (in.)= 23.624 l

Horizontal Horizontal Location Horizontal Volume (ft3)

Surface Area (ft2)

Volume (ft3)

Surface Area (ft2)

Inboard TRUE 4.32 8.78 4.32 8.78 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard FALSE 100.55 204.30 0

0.00 Inboard FALSE 39.52 80.30 0

0.00 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard TRUE 83.98 170.63 83.98 170.63 Inboard TRUE 12.424 25.24 12.424 25.24 Inboard FALSE 46.44 94.36 0

0.00 Inboard TRUE 10.37 21.07 10.37 21.07 Inboard TRUE 19.03 38.67 19.03 38.67 Penetration TRUE 68.85 139.89 68.85 139.89 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard FALSE 34.56 70.22 0

0.00 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 155.29 315.52 155.29 315.52 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 351.12 713.42 351.12 713.42 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 147.44 299.57 147.44 299.57 Outboard TRUE 10.37 21.07 10.37 21.07 Outboard TRUE 56.21 114.21 56.21 114.21 Totals 1202.694 2443.67 981.62 1994.50 Horizontal Horizontal Total Volume Total Surface Volume Surface Area (ft3)

Area (ft2)

(ft3)

(ft2)

Inboard (Node 1) 337.37 685.49 150.86 306.53 Penetration (Node 2) 68.85 139.89 68.85 139.89 Outboard (Node 3) 796.47 1618.29 761.91 1548.07 Totals 1202.69 2443.67 981.62 1994.50 Table 1: Main Steam Line Deposition Parameters - Page 7 of 7

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" Revised TS Pages Markups and TS Markup Inserts pages UNITS 2 & 3 Inserts Page 3.6-16 5.0-13

TS Inserts For PBAPS AST LAR Insert I or Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion,"

1989.

Insert 2 Verify combined MSIV leakage rate for all four main steam lines is 5 150 scfh, and < 75 l

scfh for any one steam line, when tested at 2 25 psig.

t48470I1,423:#*,

9 3-27-03; 9:25A:M PCIMs 3-6.1.3 3

SURVEILLANCE REQUIREHENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.14 et hro eac 4

s In accordance t

psi with the Primary Containment Leakage Rate

/

A 2Testing Program SR 3.6.1.3.15 Verify each 6 inch and 18 1 Mc primary 24 months containment purge valve and each 18 inch primary containment. exhaust valve is blocked to restrict opening greater than the required maximum opening angle.

SR 3.6.1.3.16 Replace the inflatable seal of each 96 months 6 inch and I8 inch primary containment purge valve and each 18 inch primary containment exhaust valve.

I PBAPS UNIT 2 3.6-16 Amendment No.220

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) continued)

b.

Demonstrate for &eh-the jWsyserW that an inplace test of the charcoal a sor er shows a penetration and system bypass < 1.0% when tested in accordance with Regulatory Guide 1.52, Revisi-o-n question 5d, and ASME N510-1989, Sectio 11, at the system flowrate specified below.

ESF Ventilation System Floqate (cfm)

MCREV System 2700 to 3300

c.

Demonstrate for the syste hat a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, Section 6b, shows the methyl iodide penetration less than the value specified below when tested in accordance with the laboratory testing criteria of ASTM D3803-1989 at a temperature of 30 degrees C [86 degrees F], face velocity, F-and the relative humidity specified below.

GT i

M-CREV Svstem Penetration 6

Face Velocity 60 57 (FPM)

Relative Humidity:

95 (X)

(continued)

PBAPS UNIT 2 5.0-13 Amendment No.

237

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.14 J

t hro h ea In accordance en sted/at k 5pyg/

with the Primary Containment i/ 2 Leakage Rate Testing Program SR 3.6.1.3.15 Verify each 6 inch and 18 inch primary 24 months containment purge valve and each 18 inch primary containment exhaust valve is blocked to restrict opening greater than the required maximum opening angle.

SR 3.6.1.3.16 Replace the inflatable seal of each 96 months 6 inch and 18 inch primary containment purge valve and each 18 inch primary containment exhaust valve.

I PBAPS UNIT 3 3.6-16 Amendment No. 223

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTc) continued)

b. Demonstrate for the g sys 9that an inplace test of the charcoal a dorber shows a penetration and system bypass < 1.0% when tested in accordance with Regulatory Guide L.U2 Revisi tion 5d, and ASME N510-1989,

]Secti o

11, at the system flowrate specifie~Ei ESF Ventilation System Flowrate (cfm)

MCREV System 2700 to 3300

c. Demonstrate for the systen that a laboratory test of a sample R

te charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, Section 6b, shows the methyl iodide penetration less than the value specified below when tested in accordance with the laboratory testing criteria of ASTM D3803-1989 at a temperature of 30 degrees C [86 degrees F], face velocity, and the relative humidity specifias below.

Penetration

(%)

Face Velocity (FPM)

MCREV System

.5 57 95 Relative Humidity:

(X,)

(continued)

PBAPS UNIT 3 5.0-13 Amendment No. 240

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" Revised TS Bases Inserts (For information only)

UNITS 2 & 3 Revised TS Bases Inserts B3.6-29

PBAPS Units 2 and 3 Technical Specification Bases Markup Inserts INSERT A {op. B 3.6-291 Total leakage through all four main steam lines must be < 150 scfh, and < 75 scfh for any one steam line, when tested at > 25 psig. The analysis in Reference I 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 {IQ. 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 foi. 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 fpq. 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 {pc. 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

INSERT G {np. 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 {fq. B 3.6-851 These filters are not credited in any DBA analysis.

INSERT I fop. 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 fop. 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 prefilters, 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 fpq. 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 fpq. 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 3.6-741 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

INSERT N {pc 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

INSERT P (pg 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

INSERT Q fpq 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

INSERT R IPG B 3.8-42,43,72,73,74. 94. and 951 involving recently irradiated fuel INSERT S fpq 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

INSERT T {pn 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

INSERT U fpq 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 {pa 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 rDq 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

INSERT X fpa 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

INSERT Y {pn 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 in the calculation are the Boron-10 minimum mass of 162.7 Ibm, and an upper bound Boron-1 0 enrichment of 65%.

INSERT Z {pf 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

INSERT AA {pc 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.13 This SR ensures that in case the non-safety grade instrument air system is unavailable, the SGIG System will perform its design function to supply nitrogen gas at the required pressure for valve operators and valve seals supported by the SGIG System. The 24 month Frequency was developed considering it is prudent that this Surveillance be performed only during a plant outage. Operating experience has shown that these components will usually pass this Surveillance when.performed at the 24 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Verifying the opening of each 6 inch and 18 inch primary containment purge valve and each 18 inch primary containment exhaust valve is restricted by a blocking device t6 less than or equal to the required maximum opening angle specified in the UFSAR (Ref. 4) is required to ensure that the valves can close under DBA conditions within the times in the analysis of Reference 1. If a LOCA occurs, the purge and exhaust valves must close'to maintain primary containment leakage within the values assumed in the accident analysis.

At other times pressurization concerns are not present, thus the purge and exhaust valves can be fully open. The 24 month Frequency is appropriate because the blocking devices may be removed during a refueling outage.

(continued)

PBAPS UNIT 2 B 3.6-29 Revision No. 22

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.13 This SR ensures that in case the non-safety grade instrument air system is unavailable, the SGIG System will perform its design function to supply nitrogen gas at the required pressure for valve operators and valve seals supported by the SGIG System. The 24 month Frequency was developed considering it is prudent that this Surveillance be performed only during a plant outage. Operating experience has shown that these components will usually pass this Surveillance when performed at the 24 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.1.3.15 Verifying the opening of each 6 inch and 18 inch primary containment purge valve and each 18 inch primary containment exhaust valve is restricted by a blocking device to less than or equal to the required maximum opening angle specified in the UFSAR (Ref. 4) is required to ensure that the valves can close under DBA conditions within the times in the analysis of Reference 1. If a LOCA occurs, the purge and exhaust valves must close to maintain primary containment leakage within the values assumed in the accident analysis. At other times pressurization concerns are not present, thus the purge and exhaust valves can be fully open.

The 24 month Frequency is appropriate because the blocking devices may be removed during a refueling outage.

(continuepd)

PBAPS UNIT 3 B 3.6-29 Revision No. 24

ATTACHMENT 4 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" Revised Camera-ready TS pages UNITS 2 & 3 3.6-16 5.0-13

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.14 Verify combined MSIV leakage rate for all In accordance four main steam lines is 5 150 scfh, and with the

75 scfh for any one steam line, when Primary tested at 2 25 psig.

Containment Leakage Rate Testing Program SR 3.6.1.3.15 Verify each 6 inch and 18 inch primary 24 months containment purge valve and each 18 inch primary containment exhaust valve is blocked to restrict opening greater than the required maximum opening angle.

SR 3.6.1.3.16 Replace the inflatable seal of each 96 months 6 inch and 18 inch primary containment purge valve and each 18 inch primary containment exhaust valve.

PBAPS UNIT 2 3.6-16 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP)

(continued)

b.

Demonstrate for the MCREV system that an inplace test of the charcoal adsorber shows a penetration and system bypass

< 1.0% when tested in accordance with Regulatory Guide 1.52, Revision 2, Section 5d, and ASME N510-1989, Section 11, at the system flowrate specified below.

ESF Ventilation System Flowrate (cfm.

MCREV System 2700 to 3300

c.

Demonstrate for the MCREV system that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, Section 6b, shows the methyl iodide penetration less than the value specified below.

when tested in accordance with the laboratory testing criteria of ASTM D3803-1989 at a temperature of 30 degrees C

[86 degrees F), face velocity, and the relative humidity specified below.

MCREV System Penetration 5

(M)

Face Velocity 57 (FPM)

Relative Humidity:

95

(%)

(continued)

PBAPS UNIT 2 5.0 -13 Amendment No.

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUFNCY SR 3.6.1.3.14 Verify combined MSIV leakage rate for all In accordance four main steam lines is

150 scfh, and with the

75 scfh for any one steam line, when Primary tested at 2 25 psig.

Containment Leakage Rate Testing Program SR 3.6.1.3.15 Verify each 6 inch and 18 inch primary 24 months containment purge valve and each 18 inch primary containment exhaust valve is blocked to restrict opening greater than the required maximum opening angle.

SR 3.6.1.3.16 Replace the inflatable seal of each 96 months 6 inch and 18 inch primary containment purge valve and each 18 inch primary containment exhaust valve.

PBAPS UNIT 3 3.6 -16 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP)

(continued)

b.

Demonstrate for the MCREV system that an inplace test of the charcoal adsorber shows a penetration and system bypass < 1.0O when tested in accordance with Regulatory Guide 1.52, Revision 2, Section 5d, and ASME N510-1989, Section 11, at the system flowrate specified below.

ESF Ventilation System Flowrate (cfm)

MCREV System 2700 to 3300

c.

Demonstrate for the MCREV system that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, Section 6b, shows the methyl iodide penetration less than the value specified below when tested in accordance with the laboratory testing criteria of ASTM D3803-1989 at a temperature of 30 degrees C [86 degrees FJ, face velocity, and the relative humidity specified below.

MCREV Systeml Penetration 5

(X)

Face Velocity 57 CFPM)

Relative Humidity:

95 (M)

(continued)

PBAPS UNIT 3 5.0-13 Amendment No.

ATTACHMENT 5 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" Revised Camera-ready TS Bases Pages (For information only)

UNITS 2 & 3 B3.6-29

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.13 This SR ensures that in case the non-safety grade instrument air system is unavailable, the SGIG System will perform its design function to supply nitrogen gas at the required pressure for valve operators and valve seals supported by the SGIG System.

The 24 month Frequency was developed considering it is prudent that this Surveillance be performed only during a plant outage.

Operating experience has shown that these components will usually pass this Surveillance when performed at the 24 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.1.3.14 Total leakage through all four main steam lines must be 5150 scfh, and *75 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.

SR 3.6.1.3.15 Verifying the opening of each 6 inch and 18 inch primary containment purge valve and each 18 inch primary containment exhaust valve is restricted by a blocking device to less than or equal to the required maximum opening angle specified in the UFSAR (Ref. 4) is required to ensure that the valves can close under DBA conditions within the times in the analysis of Reference 1. If a LOCA occurs, the purge and exhaust valves must close to maintain primary containment leakage within the values assumed in the accident analysis.

At other times pressurization concerns are not present, thus the purge and exhaust valves can be fully open. The 24 month Frequency is appropriate because the blocking devices may be removed during a refueling outage.

(continued)

PBAPS UNIT 2 B 3.6-29 Rev i sion No.

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.13 This SR ensures that in case the non-safety grade instrument air system is unavailable, the SGIG System will perform its design function to supply nitrogen gas at the required pressure for valve operators and valve seals supported by the SGIG System.

The 24 month Frequency was developed considering it is prudent that this Surveillance be performed only during a plant outage.

Operating experience has shown that these components will usually pass this Surveillance when performed at the 24 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.1.3.14 Total leakage through all four main steam lines must be *150 scfh, and *75 scfh for any one steam line, when tested at Ž25 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.

SR 3.6.1.3.15 Verifying the opening of each 6 inch and 18 inch primary containment purge valve and each 18 inch primary containment exhaust valve is restricted by a blocking device to less than or equal to the required maximum opening angle specified in the UFSAR (Ref. 4) is required to ensure that the valves can close under DBA conditions within the times in the analysis of Reference 1. If a LOCA occurs, the purge and exhaust valves must close to maintain primary containment leakage within the values assumed in the accident analysis. At other times pressurization concerns are not present, thus the purge and exhaust valves can be fully open.

The 24 month Frequency is appropriate because the blocking devices may be removed during a refueling outage.

(continued)

PBAPS UNIT 3 B 3.6-29 Revision No.

ATTACHMENT 6 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" Regulatory Guide 1.183 Comparison Table Revision

Page 1 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 REGULATORY GUIDE 1.183 COMPARISON Table A: Conformance with Regulatorv Guide (RG) 1.183 Main Sections- :-

'I.- '

-- I RG RGPosition I PBAPS-Comments.

Section j 'Anal sis '

3.1 The inventory of fission products in the reactor core and available Conforms ORIGEN 2.1 based methodology was used to for release to the containment should be based on the maximum determine core inventory. Power level used full power operation of the core with, as a minimum, current was 3514.9 MWt which is approximately the licensed values for fuel enrichment, fuel burnup, and an assumed current licensed reactor thermal power or core power equal to the current licensed rated thermal power times 3514 MWt. These source terms were the ECCS evaluation uncertainty. The period of irradiation should evaluated at end-of-cycle and at beginning of be of sufficient duration to allow the activity of dose-significant cycle (100 effective full power days (EFPD) to radionuclides to reach equilibrium or to reach maximum values.

achieve equilibrium) conditions and worst case The core inventory should be determined using an appropriate inventory used for the selected isotopes.

isotope generation and depletion computer code such as ORIGEN These values were then divided by 3514.9 2 or ORIGEN-ARP. Core inventory factors (Ci/MWt) provided in MWt to obtain activity in units of Ci/MWt.

TID 14844 and used in some analysis computer codes were Accident analyses are based on a 3528 MWt derived for low burnup, low enrichment fuel and should not be used power level that is the current accident with higher burnup and higher enrichment fuels.

analysis design basis allowance for instrument uncertainty.

Source terms are based on a 2 year fuel cycle with a nominal 711 EFPD per cycle.

3.1 For the DBA LOCA, all fuel assemblies in the core are assumed to Conforms Peaking factors of 1.7 are used for DBA be affected and the core average inventory should be used. For events that do not involve the entire core, with DBA events that do not involve the entire core, the fission product fission product inventories for damaged fuel inventory of each of the damaged fuel rods is determined by rods determined by dividing the total core dividing the total core inventory by the number of fuel rods in the inventory by the number of fuel rods in the core. To account for differences in power level across the core, core.

radial peaking factors from the facility's core operating limits report (COLR) or technical specifications should be applied in determining the inventory of the damaged rods.

3.1 No adjustment to the fission product inventory should be made for Conforms No adjustments for less than full power are events postulated to occur during power operations at less than full made in any analyses.

rated power or those postulated to occur at the beginning of core life. For events postulated to occur while the facility is shutdown, e.g., a fuel handling accident, radioactive decay from the time of

I, Page 2of23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004

.TableA: ConformancewithReaulatorv Guide(RG) 1.183MainSections%.:

I.

I" ;,

I I I.

RG RG Position-PBAPS Comment s.:

shutdown may be modeled.

3.2 The core inventory release fractions, by radionuclide groups, for the Conforms The fractions from Regulatory Position 3.1, gap release and early in-vessel damage phases for DBA LOCAs Table I are used.

are listed in Table 1 for BWRs and Table 2 for PWRs. These fractions are applied to the equilibrium core inventory described in Footnote 10 criteria are met.

Regulatory Position 3.1.

Table 1 BWR Core Inventory Fraction Released Into Containment Gap Early Release In-Vessel Group Phase Phase Total Noble Gases 0.05 0.95 1.0 Halogens 0.05 0.25 0.3 Alkali Metals 0.05 0.20 0.25 Tellurium Metals 0.00 0.05 0.05 Ba, Sr 0.00 0.02 0.02 Noble Metals 0.00 0.0025 0.0025 Cerium Group 0.00 0.0005 0.0005 Lanthanides 0.00 0.0002 0.0002 Footnote 10:

The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak rod bumup up to 62,000 MWD/MTU. The data in this section may not be applicable to cores containing mixed oxide (MOX) fuel.

t1e it Page 3 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table A: Conformance with Regul atory Guide (RG) 1 183 Main Section-S_____

RG RG Position PBAPS-,

Comments---

S ec~tion

-2A n~yi 3.2 For non-LOCA events, the fractions of the core inventory assumed Conforms Complies with Note 11 of Table 3.

to be in the gap for the various radionuclides are given in Table 3.

The release fractions from Table 3 are used in conjunction with the Peaking factor of 1.7 used for DBA events that fission product inventory calculated with the maximum core radial do not involve the entire core.

peaking factor.

Table 3 Non-LOCA Fraction of Fission Product Inventory In Gap Group Fraction 1-131 0.08 Kr-85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals 0.12 Footnote 11:

The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak bumup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average powerforrods with bumups that exceed 54 GWD/MTU. As an alternative, fission gas release calculations performed using NRC-approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load. Forthe BWR rod drop accident and the PWR rod ejection accident, the gap fractions are assumed to be 10% for iodines and noble gases.

3.3 Table 4 tabulates the onset and duration of each sequential release Conforms The BWR durations from Table 4 are used.

phase for DBA LOCAs at PWRs and BWRs. The specified onset is LOCA is modeled in a linear fashion.

the time following the initiation of the accident (i.e., time = 0). The Non-LOCA is modeled as an instantaneous early in-vessel phase immediately follows the gap release phase.

release.

The activity released from the core during each release phase should be modeled as increasing in a linear fashion over the

Page 4 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table A: Conformance with Regulatory Guide (RG) 1;183 Main Sections RG RG Position PBAPS Comments Section An' A al ysis duration of the phase. For non-LOCA DBAs, in which fuel damage is projected, the release from the fuel gap and the fuel pellet should be assumed to occur instantaneously with the onset of the projected damage.

Table 4 LOCA Release Phases PWRs BWRs Phase Onset Duration Onset Duration Gap Release 30 sec 0.5 hr 2 min 0.5 hr Early In-Vessel 0.5 hr 1.3 hr 0.5 hr 1.5 hr 3.3 For facilities licensed with leak-before-break methodology, the Not PBAPS does not use leak-before-break onset of the gap release phase may be assumed to be 10 minutes.

Applicable methodology for DBA analyses.

A licensee may propose an alternative time for the onset of the gap release phase, based on facility-specific calculations using suitable analysis codes or on an accepted topical report shown to be applicable for the specific facility. In the absence of approved alternatives, the gap release phase onsets in Table 4 should be used.

3.4 Table 5 lists the elements in each radionuclide group that should be Conforms The nuclides used are the 60 identified as considered in design basis analyses.

being potentially important dose contributors Table 5 to total effective dose equivalent (TEDE) in the Radionuclide Groups RADTRAD code, which encompasses those Group Elements listed in RG 1.183, Table 5.

Noble Gases Xe, Kr Halogens I, Br Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba, Sr Noble Metals Ru, Rh, Pd, Mo, Tc, Co Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am Cerium Ce, Pu, Np 3.5 Of the radioiodine released from the reactor coolant system (RCS)

Conforms This guidance is applied in the analyses.

to the containment in a postulated accident, 95 percent of the iodine II

Al Page 5 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 TableA: ConformancewithRegulatoryGuide (RG),1.183MMainSections

.I

-I RG'ctin RG Position, PBAPS -

comments v.

Sect 2....

Aaysis.--

released should be assumed to be cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide. This includes releases from the gap and the fuel pellets. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form. The same chemical form is assumed in releases from fuel pins in FHAs and from releases from the fuel pins through the RCS in DBAs other than FHAs or LOCAs. However, the transport of these iodine species following release from the fuel may affect these assumed fractions. The accident-specific appendices to this regulatory guide provide additional details.

3.6 The amount of fuel damage caused by non-LOCA design basis Conforms Fuel damage assessment for CRDA and FHA events should be analyzed to determine, for the case resulting in are based on GESTAR standard analyses for the highest radioactivity release, the fraction of the fuel that reaches GE14 fuel.

or exceeds the initiation temperature of fuel melt and the fraction of fuel elements for which the fuel clad is breached. Although the NRC staff has traditionally relied upon the departure from nucleate boiling ratio (DNBR) as a fuel damage criterion, licensees may propose other methods to the NRC staff, such as those based upon enthalpy deposition, for estimating fuel damage for the purpose of establishing radioactivity releases.

4.1.1 The dose calculations should determine the TEDE. TEDE is the Conforms TEDE is calculated, with significant progeny sum of the committed effective dose equivalent (CEDE) from included.

inhalation and the deep dose equivalent (DDE) from external exposure. The calculation of these two components of the TEDE should consider all radionuclides, including progeny from the decay of parent radionuclides that are significant with regard to dose consequences and the released radioactivity.

4.1.2 The exposure-to-CEDE factors for inhalation of radioactive material Conforms Federal Guidance Report 11 dose conversion should be derived from the data provided in ICRP Publication 30, factors (DCFs) are used.

"Limits for Intakes of Radionuclides by Workers" (Ref. 19). Table 2.1 of Federal Guidance Report 11, "Limiting Values of

Page 6 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table A: Conformance with Reaulatorv Guide (RG) 1.183 Main Sections._. -.

RG RG7Position PBAPS; Comments':

Section-Analsis Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20),

provides tables of conversion factors acceptable to the NRC staff.

The factors in the column headed "effective" yield doses corresponding to the CEDE.

4.1.3 For the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the breathing rate of persons offsite should be Conforms The analysis uses values to three significant assumed to be 3.5 x 104 cubic meters per second. From 8 to 24 figures that correspond to the rounded values hours following the accident, the breathing rate should be assumed in Section 4.1.3 of RG 1.183.

to be 1.8 x 104 cubic meters per second. After that and until the end of the accident, the rate should be assumed to be 2.3 x 104 cubic meters per second.

4.1.4 The DDE should be calculated assuming submergence in semi-Conforms Federal Guidance Report 12 conversion infinite cloud assumptions with appropriate credit for attenuation by factors are used.

body tissue. The DDE is nominally equivalent to the effective dose equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE. Table 111.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 21), provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.

4.1.5 The TEDE should be determined for the most limiting person at the Conforms The maximum two-hour LOCA EAB dose is as EAB. The maximum EAB TEDE for any two-hour period following follows:

the start of the radioactivity release should be determined and used in determining compliance with the dose criteria in 10 CFR 50.67.

PC Leakage: 0.0 to 2.0 hours0 days 0 hours 0 weeks 0 months (6.907 Rem The maximum two-hour TEDE should be determined by calculating TEDE) due to the 15-minute unfiltered, the postulated dose for a series of small time increments and ground-level SC drawdown time.

performing a "sliding" sum over the increments for successive two-MSIV Leakage: 11.8 to 13.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (1.155 hour0.00179 days 0.0431 hours 2.562831e-4 weeks 5.89775e-5 months periods. The maximum TEDE obtained is submitted. The time Rem TEDE).

increments should appropriately reflect the progression of the ECCS Leakage: 2.0 to 4.0 hours0 days 0 hours 0 weeks 0 months (0.104 Rem accident to capture the peak dose interval between the start of the TEDE)

Page 7 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table A: Conformance with Regulatory Guide (RG) 1.183 Main Sections:: -

RG RG Position.

'PBAPS-Comments Section-Aralysis event and the end of radioactivity release (see also Table 6).

Conservatively, the maximum 2-hour period Footnote 14:

dose was determined by adding the maximum With regard to the EAB TEDE, the maximum two-hour value is the 2-hour dose for each of the components listed basis forscreening and evaluation under 10 CFR 50.59. Changes above even though they do not occur to doses outside of the two-hour window are only considered in the simultaneously. This yields: 6.907 + 1.155 +

context of their impact on the maximum two-hour EAB TEDE.

0.104 = 8.166 Rem TEDE (Rounded up to 8.2 Rem TEDE).

4.1.6 TEDE should be determined for the most limiting receptor at the Conforms This guidance is applied in the analyses.

outer boundary of the low population zone (LPZ) and should be used in determining compliance with the dose criteria in 10 CFR 50.67.

4.1.7 No correction should be made for depletion of the effluent plume by Conforms No such corrections made in the analyses.

deposition on the ground.

4.2.1 The TEDE analysis should consider all sources of radiation that will Conforms The principal source of dose within the control cause exposure to control room personnel. The applicable sources room is due to airborne activity.

will vary from facility to facility, but typically will include:

Contamination of the control room atmosphere by the intake or Calculations of doses from reactor building infiltration of the radioactive material contained in the radioactive airborne activity have been recalculated with plume released from the facility, AST source term assumptions, no credit for Contamination of the control room atmosphere by the intake or contained structures except floors and exterior infiltration of airborne radioactive material from areas and structures walls, and with a relatively detailed geometry adjacent to the control room envelope, treatment.

Radiation shine from the external radioactive plume released from the facility, SGTS and MCREV filters are well away and/or Radiation shine from radioactive material in the reactor shielded from the Control Room and have not containment, historically been considered a source for Radiation shine from radioactive material in systems and operator doses. This historical conclusion components inside or external to the control room envelope, e.g.,

continues to apply as discussed below.

radioactive material buildup in recirculation filters.

Gamma shine from reactor building:

This component includes shine from the unshielded refuel floor and from the shielded

I l Page 8 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table A: Conformance with Regulatory Guide (RG) 1.183 Main Sections

_-__,-_-:_i_

_-_i RG RG Position PBAPS Comments Section tis volume below the refuel floor. Reanalysis with AST source terms indicate a dose contribution of only 0.05 Rem for the duration of the accident.

SGTS filter shine:

This effect due to this source is negligible because the SGTS filter assembly is located on the 91'6" elevation of the radwaste building..

The Control Room is located on the 165' elevation within the turbine building, well away from the filters.

MCREV filter shine:

The effect due to the MCREV filters is negligible because the filters are 30' away from the control room air space with an intervening 2' concrete shield. This conclusion is based on experience from other Exelon units with similar geometry.

Primary containment shine:

The 2' reactor building wall plus the 5' containment wall provides ample shielding for the control room.

External cloud:

The control room in an interior space, surrounded by its own 2' thick wall and ceiling concrete shielding. Therefore, doses due to the external cloud is negligible.

4.2.2 The radioactive material releases and radiation levels used in the Conforms The source term, transport, and release control room dose analysis should be determined using the same methodology is the same for both the control

Page 9 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table A: Conformance with Regulatory Guide (RG) 1.183 Main Sections:

RG-RG Position PBAPS

-Comments.

Section t.

s a,______________,_____;,.________________.,_

Analysis Hi-source term, transport, and release assumptions used for room and offsite locations.

determining the EAB and the LPZ TEDE values, unless these assumptions would result in non-conservative results for the control room.

4.2.3 The models used to transport radioactive material into and through Conforms This guidance is applied in the analyses.

the control room, and the shielding models used to determine radiation dose rates from external sources, should be structured to provide suitably conservative estimates of the exposure to control room personnel.

4.2.4 Credit for engineered safety features that mitigate airborne Conforms Control Room pressurization and intake radioactive material within the control room may be assumed. Such filtration are credited in the LOCA accident features may include control room isolation or pressurization, or analysis. No credit is taken in the FHA, MSLB intake or recirculation filtration. Refer to Section 6.5.1, "ESF and CRDA accident analyses.

Atmospheric Cleanup System," of the SRP (Ref. 3) and Regulatory No credit is taken for SGTS HEPA or charcoal Guide 1.52, "Design, Testing, and Maintenance Criteria for Post-adsorber filtration in any accident.

accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants" (Ref. 25), for guidance.

4.2.5 Credit should generally not be taken for the use of personal Conforms Such credits are not taken.

protective equipment or prophylactic drugs. Deviations may be considered on a case-by-case basis.

4.2.6 The dose receptor for these analyses is the hypothetical maximum Conforms The cited occupancy factors and breathing exposed individual who is present in the control room for 100% of rate are used. An unrounded breathing rate of the time during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event, 60% of the time 3.47E-04 m3/sec is used.

between 1 and 4 days, and 40% of the time from 4 days to 30 days.

For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 10 4 cubic meters per second.

AI, Page 10 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table A: Conformance with Regulatory Guide (RG) 1 83 Main Sections RG RG Position PBAPS

Comments Section i

- : : E - : : t ~ : ; :

A nalysis.

5

_ s 4.2.7 Control room doses should be calculated using dose conversion Conforms The equation given is utilized for finite cloud factors identified in Regulatory Position 4.1 above for use in offsite correction when calculating external doses dose analyses. The DDE from photons may be corrected for the due to the airborne activity inside the control difference between finite cloud geometry in the control room and room.

the semi-infinite cloud assumption used in calculating the dose conversion factors. The following expression may be used to correct the semi-infinite cloud dose, DDE., to a finite cloud dose, DDEflt,, where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room (Ref. 22).

DDE DDEV0.

33 8

~1 1 7 3_

4.3 The guidance provided in Regulatory Positions 4.1 and 4.2 should Conforms The Technical Support Center at PBAPS is in be used, as applicable, in re-assessing the radiological analyses the Unit I Control Room. A review of the identified in Regulatory Position 1.3.1, such as those in NUREG-current TID-14844 based analysis indicates 0737 (Ref. 2). Design envelope source terms provided in NUREG-that it is unnecessary to reanalyze doses 0737 should be updated for consistency with the AST. In general, therein to assure accessibility.

radiation exposures to plant personnel identified in Regulatory For other areas requiring plant personnel Position 1.3.1 should be expressed in terms of TEDE. Integrated access, a qualitative assessment of the radiation exposure of plant equipment should be determined using regulatory positions on source terms indicates the guidance of Appendix I of this guide.

that, with no new operator actions required, radiation exposures are bounded by those previously analyzed.

5.1.1 The evaluations required by 10 CFR 50.67 are re-analyses of the Conforms These analyses were prepared as specified in design basis safety analyses and evaluations required by 10 CFR the guidance.

50.34; they are considered to be a significant input to the evaluations required by 10 CFR 50.92 or 10 CFR 50.59. These analyses should be prepared, reviewed, and maintained in accordance with quality assurance programs that comply with Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.

5.1.2 Credit may be taken for accident mitigation features that are Conforms The analyses take credit for SLC System classified as safety-related, are required to be operable by technical I operation. The SLC System is safety-related,

Page 11 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix l.

April 2004 Table A: Conformance with Renilatorv Guide (RG) 1.183 Main'Sections^:.- :* *-

I RG -;

RG Position.",'

PBAPS Comments,,

Section

.Anal sis specifications, are powered by emergency power sources, and are required to be operable by Technical either automatically actuated or, in limited cases, have actuation Specifications, and supplied with emergency requirements explicitly addressed in emergency operating power. The SLC System is manually initiated procedures. The single active component failure that results in the from the main control room, as directed by the most limiting radiological consequences should be assumed.

emergency operating procedures. There are Assumptions regarding the occurrence and timing of a loss of four proceduralized injection methods for SLC offsite power should be selected with the objective of maximizing with at least one alternate method for SLC the postulated radiological consequences.

injection that does not require personnel access into the secondary containment.

5.1.3 The numeric values that are chosen as inputs to the analyses Conforms Conservative assumptions are used based on required by 10 CFR 50.67 should be selected with the objective of nominal values, as per prior plant analysis determining a conservative postulated dose. In some instances, a practice.

particular parameter may be conservative in one portion of an analysis but be non-conservative in another portion of the same analysis.

5.1.4 Licensees should ensure that analysis assumptions and methods Conforms Analysis assumptions and methods were are compatible with the AST and the TEDE criteria.

made per this guidance.

5.3 Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the Conforms New atmospheric dispersion values (X/Q) for control room that were approved by the staff during initial facility the EAB, the LPZ, and the control room were licensing or in subsequent licensing proceedings may be used in developed, using meteorological data for the performing the radiological analyses identified by this guide.

years 1984-1988. ARCON96 and PAVAN Methodologies that have been used for determining X/Q values are were used with these data to'determine documented in Regulatory Guides 1.3 and 1.4, Regulatory Guide control room and EAB/LPZ atmospheric 1.145, "Atmospheric Dispersion Models for Potential Accident dispersion values. Control room X/Qs from Consequence Assessments at Nuclear Power Plants," and the releases from the Main Stack were developed paper, "Nuclear Power Plant Control Room Ventilation System in conformance with RG -1.194.

Design for Meeting General Criterion 19".

The NRC computer code PAVAN implements Regulatory Guide 1.145 and its use is acceptable to the NRC staff. The methodology of the NRC computer code ARCON96 is generally acceptable to the NRC staff for use in determining control room X/Q values.

Page 12 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix April 2004 I

JTable B: Conformance withRGAA83 Appendix A (Loss-of-Coolant Accident) tomzip

___________-a-*

RG RG Position PBAPS ;

Comment.

Section -

7 Analysis Acceptable assumptions regarding core inventory and the release of Conforms Fission Product Inventory: Core radionuclides from the fuel are provided in Regulatory Position 3 of this source terms are developed using guide.

ORIGEN-2.1 based methodology.

Release Fractions: Release fractions are per Table 1 of RG 1.183, and are implemented by RADTRAD.

Timing of Release Phases: Release Phases are per Table 4 of RG 1.183, and are implemented by RADTRAD.

Radionuclide Composition:

Radionuclide grouping is per Table 5 of RG 1.183, as implemented in RADTRAD.

Chemical Form: Treatment of release chemical form is per RG 1.183, Section_3.5.

2 If the sump or suppression pool pH is controlled at values of 7 or Conforms The stated distributions of iodine greater, the chemical form of radioiodine released to the containment chemical forms are used.

should be assumed to be 95% cesium iodide (Csl), 4.85 percent The post-LOCA suppression pool pH elemental iodine, and 0.15 percent organic iodide. Iodine species, has been evaluated, including including those from iodine re-evolution, for sump or suppression pool consideration of the effects of acids pH values less than 7 will be evaluated on a case-by-case basis.

and bases created during the LOCA Evaluations of pH should consider the effect of acids and bases created event, the effects of key fission during the LOCA event, e.g., radiolysis products. With the exception of product releases, and the impact of elemental and organic iodine and noble gases, fission products should SLC injection. Suppression pool pH be assumed to be in particulate form.

remains above 7 for at least 30 days.

3.1 The radioactivity released from the fuel should be assumed to mix Conforms The radioactivity release from the instantaneously and homogeneously throughout the free air volume of fuel is assumed to homogeneously the primary containment in PWRs or the drywell in BWRs as it is mix only in the drywell free volume released. This distribution should be adjusted if there are internal during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the compartments that have limited ventilation exchange. The suppression assumed LOCA, and in the pool free air volume may be included provided there is a mechanism to combined drywell free air volume and

Page 13 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table B: Conformance with RG 1.183 Appendix A (Loss-of-Coolant Accident)

RG RG Position

- PBAPS Comments Section:

-Analysis ensure mixing between the drywell to the wetwell. The release into the suppression chamber air space for containment or drywell should be assumed to terminate at the end of the the 2 to 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months period. Mixing is early in-vessel phase.

caused by steam flashing and flow from the drywell through the suppression pool to the suppression chamber air space, after core reflood.

3.2 Reduction in airborne radioactivity in the containment by natural Conforms Credit is taken for natural deposition deposition within the containment may be credited. Acceptable models per the methodology of NUREG/CR-for removal of iodine and aerosols are described in Chapter 6.5.2, 6189, as implemented in RADTRAD.

"Containment Spray as a Fission Product Cleanup System," of the No deterministically assumed initial Standard Review Plan (SRP), NUREG-0800 (Ref. A-1) and in plateout is credited.

NUREGICR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments" (Ref. A-2). The latter model is incorporated into the analysis code RADTRAD (Ref. A-3).

3.3 Reduction in airborne radioactivity in the containment by containment Not While containment sprays are a spray systems that have been designed and are maintained in Applicable design feature that is available at accordance with Chapter 6.5.2 of the SRP (Ref. A-1) may be credited.

PBAPS, no credit is taken for aerosol Acceptable models for the removal of iodine and aerosols are described removal by them in the LOCA AST in Chapter 6.5.2 of the SRP and NUREG/CR-5966, "A Simplified Model reanalysis.

of Aerosol Removal by Containment Sprays"1 (Ref. A-4). This simplified model is incorporated into the analysis code RADTRAD (Refs. A-I to A-3).

The evaluation of the containment sprays should address areas within the primary containment that are not covered by the spray drops. The mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment building, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour, unless other rates are justified. The containment building atmosphere may be considered a single, well-mixed volume if the spray covers at least 90% of the volume and if adequate mixing of unsprayed compartments can be shown.

The SRP sets forth a maximum decontamination factor (DF) for

Page 14 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table B: Conformance with RG 1.183 AppendixA (Loss-of-CoolantAccident)

[I

.>. A ;

I

. I..

RGP RG osition:

P;:

~

BAPS ents Section

?

-I vA n, Analysis.

'__-A'

__i'..'

elemental iodine based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination. The SRP also states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass. There is no specified maximum DF for aerosol removal by sprays. The maximum activity to be used in determining the DF is defined as the iodine activity in the columns labeled "Total" in Tables 1 and 2 of this guide multiplied by 0.05 for elemental iodine and by 0.95 for particulate iodine (i.e., aerosol treated as particulate in SRP methodology).

3.4 Reduction in airborne radioactivity in the containment by in-containment Not No in-containment recirculation filter recirculation filter systems may be credited if these systems meet the Applicable systems exist at PBAPS.

guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. A-5 and A-6). The filter media loading caused by the increased aerosol release associated with the revised source term should be addressed.

3.5 Reduction in airborne radioactivity in the containment by suppression Conforms No credit is taken for suppression pool scrubbing in BWRs should generally not be credited. However, the pool scrubbing in the LOCA AST staff may consider such reduction on an individual case basis. The reanalysis. Analyses have been evaluation should consider the relative timing of the blowdown and the performed that determined that the fission product release from the fuel, the force driving the release suppression pool liquid pH is through the pool, and the potential for any bypass of the suppression maintained greater than 7, and that, pool (Ref. 7). Analyses should consider iodine re-evolution if the therefore, iodine re-evolution is not suppression pool liquid pH is not maintained greater than 7.

expected.

3.6 Reduction in airborne radioactivity in the containment by retention in ice Not PBAPS does not have ice condensers, or other engineering safety features not addressed above, Applicable condensers. No other removal should be evaluated on an individual case basis. See Section 6.5.4 of mechanisms are credited other than the SRP (Ref. A-1).

natural deposition.

3.7 The primary containment (i.e., drywell for Mark I and 11 containment Conforms Primary containment leakage is designs) should be assumed to leak at the peak pressure technical assumed to be at the 0.7% of specification leak rate for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . For PWRs, the leak rate containment mass per day for 24 may be reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 50% of the technical hours, 0.392% from 24 to 38 hours4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months ,

'I Page 15 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table B: Conformance with RG 1.183 Appendix A (Loss-of-Coolant Accident)

'L- __I

`

RG Position PBAPS Comments S ec tio n A n aly sis7 specification leak rate. For BWRs, leakage may be reduced after the and 0.35% per day from 38 to 720 first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if supported by plant configuration and analyses, to a hours. This is based on the results of value not less than 50% of the technical specification leak rate.

the leak characteristic methodology Leakage from subatmospheric containments is assumed to terminate evaluation performed (turbulent flow).

when the containment is brought to and maintained at a subatmospheric The Darcy's Formula evaluation condition as defined by technical specifications.

methodology is considered the most For BWRs with Mark Ill containments, the leakage from the drywell into conservative approach for the the primary containment should be based on the steaming rate of the evaluation of the primary heated reactor core, with no credit for core debris relocation. This containment leak rate. The large leakage should be assumed during the two-hour period between the break LOCA was found to be the initial blowdown and termination of the fuel radioactivity release (gap bounding containment pressurization and early in-vessel release phases). After two hours, the radioactivity is event. Even if a LOCA were to occur assumed to be uniformly distributed throughout the drywell and the during purging, isolation valve primary containment.

closure would occur within a small fraction of the time before start of the gap release. Dose due to this purge would be negligible as compared to other dose contributors.

PBAPS uses a Mark I containment.

3.8 If the primary containment is routinely purged during power operations, Conforms The PBAPS primary containment is releases via the purge system prior to containment isolation should be not routinely purged during power analyzed and the resulting doses summed with the postulated doses operation. Purging is limited to from other release paths. The purge release evaluation should assume inerting, de-inerting and occasional that 100% of the radionuclide inventory in the reactor coolant system short pressure control activities.

liquid is released to the containment at the initiation of the LOCA. This inventory should be based on the technical specification reactor coolant system equilibrium activity. Iodine spikes need not be considered. If the purge system is not isolated before the onset of the gap release phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable.

4.1 Leakage from the primary containment should be considered to be Conforms Secondary Containment elevated collected, processed by engineered safety feature (ESF) filters, if any, release via the Main Stack credit is and released to the environment via the secondary containment exhaust taken at 15 minutes after the start of

9.

Page 16 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table B: Conformance-with RG-1.183 Appendix A (Loss-of-Coolant Accident),-:,I RG RG Position PBAPS Comments S e c t i o n !

_ _ _ A n a l y s i s ' '

_ _ A a y i '

system during periods in which the secondary containment has a gap release. Gap release begins at negative pressure as defined in technical specifications. Credit for an

- 2 minutes after LOCA initiation.

elevated release should be assumed only if the point of physical release For EAB and LPZ doses, ground is more than two and one-half times the height of any adjacent structure.

level releases are assumed. For Control Room doses, releases are based on zero-velocity RB/TB vent stack release assumptions, yielding ground level release equivalent dispersion factors.

4.2 Leakage from the primary containment is assumed to be released Conforms For EAB and LPZ doses, ground directly to the environment as a ground-level release during any period level releases are assumed. For in which the secondary containment does not have a negative pressure Control Room doses, releases are as defined in technical specifications.

based on zero-velocity RB/TB vent stack release assumptions.

4.3 The effect of high wind speeds on the ability of the secondary Conforms The wind speed exceeded only 5%

containment to maintain a negative pressure should be evaluated on an of the time at PBAPS in the individual case basis. The wind speed to be assumed is the 1-hour secondary containment vicinity is average value that is exceeded only 5% of the total number of hours in approximately 11 mph. It has been the data set. Ambient temperatures used in these assessments should determined that a 23 mph wind be the 1-hour average value that is exceeded only 5% or 95% of the speed would be required before the total numbers of hours in the data set, whichever is conservative for the secondary containment pressures intended use (e.g., if high temperatures are limiting, use those exceeded would be positive relative to outside only 5%).

air pressures at the downwind side of the reactor enclosure.

4.4 Credit for dilution in the secondary containment may be allowed when Conforms No credit is taken for dilution/mixing adequate means to cause mixing can be demonstrated. Otherwise, the in secondary containment. An leakage from the primary containment should be assumed to be artificially small secondary transported directly to exhaust systems without mixing. Credit for containment volume is assumed in mixing, if found to be appropriate, should generally be limited to 50%.

the RADTRAD analysis in This evaluation should consider the magnitude of the containment conjunction with a large SGTS flow

I Page 17 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table B: Conformance with RG 1.183 Appendix A (Loss-of-Coolant Accident).

_____.,,,,,,_,,_x,>_,.

RG RGPosition----.

PBAPS -,Comments Section Analysis.

leakage in relation to contiguous building volume or exhaust rate, the rate to ensure mixing is not an issue.

location of exhaust plenums relative to projected release locations, the recirculation ventilation systems, and internal walls and floors that impede stream flow between the release and the exhaust.

Page 18 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 TableB: ConformancewithRG-1.183ADppendixA(Loss-of-CoolantAccident)-.'t U

I I

RG IRGPosition

A.,..-

PBAPS Comments a-;

S e c tio n n a. ysis

'Anl

_si 4.5 Primary containment leakage that bypasses the secondary containment should be evaluated at the bypass leak rate incorporated in the technical specifications. If the bypass leakage is through water, e.g., via a filled piping run that is maintained full, credit for retention of iodine and aerosols may be considered on a case-by-case basis. Similarly, deposition of aerosol radioactivity in gas-filled lines may be considered on a case-by-case basis.

Conforms' No primary containment leakage, with the exception of MSIV leakage, has been identified which bypasses the secondary containment. Only the MSIV pathway leak rates are incorporated into the Technical Specifications.

The AST analysis is based on an MSIV leakage limit of 150 scfh total leakage with not more that 75 scfh per line when tested at 2 25 psig.

The dose consequences for releases through this pathway (with piping

'deposition credit) are separately calculated. Therefore, MSIV leakage can continue to be excluded from Type B and C leakage total evaluated against the revised L. of 0.7% per day.

Piping deposition credit is determined using the AEB 98-03 well-mixed method. Delay in transit through these piping system is not credited. The credited piping is that which has previously been seismically qualified and is from the reactor vessel to the Turbine Stop Valves. However, consistent with an assumption of a LOCA in a main steam line inside containment,'the most beneficial line for deposition is assumed to have the break and also have its inboard MSIV failed. In consideration of possible turbulence

(I )

Page 19 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table B: Conformance with RG 1.183 Appendix A (Loss-of-Coolant Accident)

RG RG Position PBAPS -

'Comments '

Section Analysis.- }

in containment in the failed MSIV vicinity, the first two pipe diameters of penetration piping are not credited for this line. The balance of penetration piping is treated as inboard piping.

4.6 Reduction in the amount of radioactive material released from the Conforms SGTS HEPA and charcoal adsorber secondary containment because of ESF filter systems may be taken into filters are not credited in the account provided that these systems meet the guidance of Regulatory evaluation of analyzed accidents Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).

onsite and offsite dose consequences.

5.1 With the exception of noble gases, all the fission products released from Conforms With the exception of noble gases, all the fuel to the containment (as defined in Tables 1 and 2 of this guide) the fission products released from should be assumed to instantaneously and homogeneously mix in the the fuel to the containment are primary containment sump water (in PWRs) or suppression pool (in assumed to instantaneously and BWRs) at the time of release from the core. In lieu of this deterministic homogeneously mix only in the approach, suitably conservative mechanistic models for the transport of drywell during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after airborne activity in containment to the sump water may be used. Note the assumed LOCA, and in the that many of the parameters that make spray and deposition models combined drywell and suppression conservative with regard to containment airborne leakage are non-chamber free volume for the 2 to 720 conservative with regard to the buildup of sump activity.

hour period.

5.2 The leakage should be taken as two times the sum of the simultaneous Conforms The 5 gpm leak rate is assumed to leakage from all components in the ESF recirculation systems above be two times the sum of the which the technical specifications, or licensee commitments to item simultaneous leakage from all ECCS III.D.1.1 of NUREG-0737 (Ref. A-8), would require declaring such components as discussed in the systems inoperable. The leakage should be assumed to start at the dose calculations. ECCS leakage is earliest time the recirculation flow occurs in these systems and end at minimized at PBAPS through the latest time the releases from these systems are terminated.

implementation of the Program Consideration should also be given to design leakage through valves committed to in T.S. 5.5.2 'Primary isolating ESF recirculation systems from tanks vented to atmosphere, Coolant Sources Outside e.g., emergency core cooling system (ECCS) pump miniflow return to Containment".

the refueling water storage tank.

Since certain ECCS systems take suction immediately from the

Page 20 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table, B: -Conformance with RG1.183 Appendix A (Loss-;of-Coolant Accident)

RGS RG Position

'-.AP Coments; Section I__

nlsis suppression pool, this leak path is assumed to start at time 0.

Leakage to atmospheric tanks is credible only for lines connecting from ECCS pump discharges to such a tank, because of relative elevations. The sole leakage paths to a tank vented to atmosphere meeting this condition are the High Pressure Coolant Injection / Reactor Core Isolation Cooling test lines that discharge to the Condensate Storage Tank (CST). These lines are isolated by two normally closed valves. Since the CST contents are demineralized water, ECCS leakage would quickly turn the water basic. Therefore, minimal elemental iodine is expected, and as a result, negligible iodine volatilization.

5.3 With the exception of iodine, all radioactive materials in the recirculating Conforms With the exception of iodine, all liquid should be assumed to be retained in the liquid phase.

radioactive materials in ECCS liquids are assumed to be retained in the liquid phase.

5.4 If the temperature of the leakage exceeds 2120F, the fraction of total Not The temperature of the leakage does iodine in the liquid that becomes airborne should be assumed equal to Applicable not exceed 212'F.

the fraction of the leakage that flashes to vapor. This flash fraction, FF, should be determined using a constant enthalpy, h, process, based on the maximum time-dependent temperature of the sump water circulating outside the containment:

FF hr. -hf hfg Where: hf. is the enthalpy of liquid at system design temperature and

Page 21 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table R:,Conformance with RG 1.183 ADoendixA (Loss-of-CoolantAccidenfl

: I:'.,.

v --

RG RG Position.

.,AP Comments -

Section

-~Analysis--,

pressure; ht2 is the enthalpy of liquid at saturation conditions (14.7 psia, 212'F); and hr, is the heat of vaporization at 212'F.

5.5 If the temperature of the leakage is less than 2120F or the calculated Conforms An airborne release fraction of 1.41%

flash fraction is less than 10%, the amount of iodine that becomes is used. Suppression Pool water pH airborne should be assumed to be 10% of the total iodine activity in the is maintained above 7 for the entire leaked fluid, unless a smaller amount can be justified based on the 30 days of the accident dose actual sump pH history and area ventilation rates.

assessment period. Under these conditions virtually none of the iodine will be in elemental form, and organic iodine formation will be inhibited.

Because of the subcooled condition no flashing is expected. Neverthe-less, this value, derived based on ORNL-TM-2412 methodology for iodine partition factor determination, is used.

5.6 The radioiodine that is postulated to be available for release to the Conforms The credited Control Room intake environment is assumed to be 97% elemental and 3% organic.

charcoal and HEPA filters meet the Reduction in release activity by dilution or holdup within buildings, or by requirements of RG 1.52 and ESF ventilation filtration systems, may be credited where applicable.

Generic Letter 99-02. These are Filter systems used in these applications should be evaluated against credited at 90% efficiency for the guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-elemental and organic iodines.

02 (Ref. A-6).

Aerosol removal efficiencies are assumed to be 99% based on the HEPA/charcoal combination.

6.1 For the purpose of this analysis, the activity available for release via Conforms The radioactivity release from the MSIV leakage should be assumed to be that activity determined to be in fuel is assumed to homogeneously the drywell for evaluating containment leakage (see Regulatory Position mix only in the drywell free volume 3). No credit should be assumed for activity reduction by the steam during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the separators or by iodine partitioning in the reactor vessel.

assumed LOCA, and in the combined drywell free air volume and suppression chamber air space for the 2 to 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months period. Mixing is

. *4 i

Page 22 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table B:: Conformance with RG 1.183 Appendix A (Loss-of-Coolant Accident),

RG RG Position P

.BAPS Comments' -.-

Section Analysis~-

caused by steam flashing and flow from the drywell through the suppression pool to the suppression chamber air space, after core reflood.

6.2 All the MSIVs should be assumed to leak at the maximum leak rate Conforms MSIV leakage assumed in this above which the technical specifications would require declaring the accident analysis is 150 scfh for all MSIVs inoperable. The leakage should be assumed to continue for the steam lines and 75scflh for any one duration of the accident. Postulated leakage may be reduced after the line when tested at 2 25 psig.

first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if supported by site-specific analyses, to a value not less Reduction in leakage rates after 24 than 50% of the maximum leak rate.

hours are, as previously discussed, based on calculated post-accident containment pressures. No credit is taken for leakage rate reductions l below 50% of the MSIV leakage limit.

6.3 Reduction of the amount of released radioactivity by deposition and Conforms Modeling is per AEB 98-03 well-plateout on steam system piping upstream of the outboard MSIVs may mixed approach, with no transport be credited, but the amount of reduction in concentration allowed will be delay credit.

evaluated on an individual case basis. Generally, the model should be based on the assumption of well-mixed volumes, but other models such as slug flow may be used if justified.

6.4 In the absence of collection and treatment of releases by ESFs such as Conforms Releases are assumed to be from the MSIV leakage control system, or as described in paragraph 6.5 the RB/TB vent stacks, without credit below, the MSIV leakage should be assumed to be released to the for holdup or dilution in the environment as an unprocessed, ground-level release. Holdup and condenser or turbine building. The dilution in the turbine building should not be assumed.

zero velocity RB/TB vent stacks release assumption is effectively a ground level release assumption.

6.5 A reduction in MSIV releases that is due to holdup and deposition in Conforms Non-faulted main steam piping that is main steam piping downstream of the MSIVs and in the main capable of performing its safety condenser, including the treatment of air ejector effluent by offgas function during and following an SSE systems, may be credited if the components and piping systems used in is credited. No credit is taken for the release path are capable of performing their safety function during holdup and deposition in piping

Page 23 of 23 PBAPS AST LAR RG 1.183 (UPDATED) Compliance Matrix I April 2004 Table B: Conformance with RG 1.183 Appendix A (Loss-of-Coolant Accident)

RG:

RG Position PBAPS Comments '.

Sctio

-Analysis and following a safe shutdown earthquake (SSE). The amount of downstream of the qualified main reduction allowed will be evaluated on an individual case basis.

steam piping, or in the condenser.

References A-9 and A-10 provide guidance on acceptable models.

The modeling is per the AEB 98-03 well-mixed approach..

7.0 The radiological consequences from post-LOCA primary containment Conforms Containment purging as a purging as a combustible gas or pressure control measure should be combustible gas or pressure control analyzed. If the installed containment purging capabilities are measure is not required nor credited maintained for purposes of severe accident management and are not in any design basis analysis for 30 credited in any design basis analysis, radiological consequences need days following a design basis LOCA not be evaluated. If the primary containment purging is required within at PBAPS.

30 days of the LOCA, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. Reduction in the amount of radioactive material released via ESF filter systems may be taken into account provided that these systems meet the guidance in Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).

ATTACHMENT 7 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" Post-Accident Vital Area Access Considerations Table Revision

4 k Table 2: Post-Accident Vital Area Access Considerations (not including occupancy for the Control Room and Technical Support Center)

Activity/Access Route Current Design Bases Bases for Post-AST Accessibility Projected Doses to Individuals Accessing Vital Based on exterior dose rates and travel times, Doses have been re-analyzed with AST source Areas Requiring Continuous Occupancy such typically from Guard House to TSC or Control terms based on worst 4-hour release rates and as the control room and TSC Room. Bounding dose is 0.295 whole body.

include cloud doses and unshielded refuel floor shine and are between 0.75 and 0.82 rem.

Inhalation doses were not evaluated for this SCBA use is assumed (i.e., no inhalation

pathway, dose).

Projected Doses to Individual for Necessary Based on Travel time of 10 minutes and 8 A review of identified functions indicated that Access to and Infrequent Occupancy of Vital hours of continuous occupancy the currently 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is an excessive allowance. This Areas within Turbine Hall/Radwaste Building analyzed whole body dose is 1.675 rem.

occupancy assumption has been reduced to I Complex (Chem Lab/Counting Room, PASS, hour except for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in the shielded Chem Radwaste Control Room, and Cable Spread Lab/Counting Room. Dose assessments Room) include unshielded cloud shine (no geometry factor credit) with resulting doses at a maximum of 0.295 rem for access and 1.77 rem maximum for occupancy.

Projected Total Whole Body Dose to Currently analyzed dose is 0.433 rem whole RADTRAD analyses, using control room X/Qs Individuals for Necessary Access to and body, with doses dominated by 16 minute to conservatively simulate the site in general, Occupancy of Diesel-Generator Building travel time due to DG Building Shielding.

yield a peak TEDE dose rate of 1.8 rem/hr.

For the subject 13 minute travel time the resulting dose would be 0.576 rem including 30 minute occupancy inside the shielded D/G building.

Projected Total Whole Body Dose to See to the right Access to the CAD building is no longer Individuals for Necessary Access to and required since the former Containment Occupancy of the CAD Building outside Atmosphere Dilution System post LOCA Reactor Building.

combustible gas control function is no longer required.

Table 2: Post-Accident Vital Area Access Considerations (not including occupancy for the Control Room and Technical Support Center)

Activity/Access Routc Current Design Bases Bases for Post-AST Accessibility Projected Total Whole Body Dose to Currently analyzed dose is 4.981 rem whole The location of this monitor has previously Individuals for Necessary Access to and body, with doses dominated by shine from been moved to the Turbine Building El. 195 Occupancy of the Cartridge Exchange at the refuel floor airborne activity and with a from the original location on the Reactor RAD Effluent Stack Monitor.

significant contribution from equipment/piping Building El. 234.

shine.

Additionally, a more direct pathway starting from the control room, to the monitor, and then to the Sample Analysis Chem Lab on Turbine Building El. 116 has been selected, significantly reducing operator dose. The calculated dose is 1.84 rem.

.t I

0 11 iI I

iI

e -.

I Table 2: Post-Accident Vital Area Access Considerations (not including occupancy for the Control Room and Technical Support Center)

Activity/Access Route I Current Design Bases I Bases for Post-AST Accessibility Projected Total Whole Body Dose to Individuals for Necessary Access to and Occupancy of the Makeup Water to Spent Fuel Pools at the RAD Effluent Stack Monitor.

Currently analyzed dose is 4.952 rem whole body, with doses dominated by shine from refuel floor airborne activity and with a significant contribution from equipment/piping shine.

The original analysis was performed based on provision of spent fuel pool makeup on conditions associated with a LOCA at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the event. The need for the action at that time was not addressed, nor was refuel floor accessibility.

This issue has been reassessed and accessibility is no longer considered necessary based on the following:

If spent fuel cooling is lost due to loss of offsite power, immediately after a refueling outage, at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months would be required before the onset of spent fuel pool boiling.

On the order of an additional 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months would be required to lose 10 feet of water coverage over the stored spent fuel. This would leave on the order of 13 feet of water coverage for shielding over the fuel.

Regulatory Guide 1.155 supporting documentation such as NUREG-I 109 and NSAC-103 indicated that the median loss of offsite power duration is 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , with restoration within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 90% of the time, and the maximum observed duration of 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months . More recent, and especially severe, loss of offsite power durations such as the wide-spread August 14, 2003 grid disturbance had offisite power restoration (Fermi) within 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months .

Ample time is thus available for off-site power restoration for the important but non-safety related function of spent fuel pool level control and cooling.

I I

I

ATTACHMENT 8 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" LOCA Radiological Consequences Analysis Revision

Table 9: LOCA adiol&6iceal Consequencei

_Aalysis

-Regulatory.1Limit' Location.

Duratin TEDE (rem)" T Control Room 30 days 4.59*

5 EAB Maximum, 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 8.17 25 LPZ 30 days 4.99 25 The doses here include the direct shine and inhalation doses from radioactivity drawn into the control room as well as the dose from external sources. Dose is based on an assumed MSIV total leakage of 150 scfh, which contributes 3.922 rem to the total.

The other contributions are 0.606 from Primary Containment leakage, 0.009 from ECCS leakage in Secondary Containment, and 0.050 rem from gamma shine.

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