ML042740524

Input Parameters for Salem AST Dose Calcs

ML042740524
Person / Time
Site:	Salem
Issue date:	01/22/2004
From:	Barkley B Public Service Enterprise Group
To:	Office of Nuclear Reactor Regulation
References
LR-N04-0413 S-C-ZZ-MDC-1987, Rev 1
Download: ML042740524 (15)

Text

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 200T43907 FORM I NC.DE-AP.ZZ-0002 CALC NO.:

S-C-ZZ-MI)C-1987 CALCULATION COVER SHEET Page 1 of 16 REVISION:

I CALC. TITLE:

Input Parameters for Salem AST Dose Cales

1. SHTS (CALC): I 15

1. ATT I # SHTS:

010 I # iDVI50.69 SHTS: 7 413 1 #TOTAL SHTS: 7 22 CHECK ONE:

3 FINAL aZ INTERIM (Proposed Plant Change)

E FINAL (Future Confirmation Req'd)

O VOID SALEM OR HOPE CREEK:

0 Q - LIST 0 IMPORTANT TO SAFETY El NON-SAFETY RELATED HOPE CREEK ONLY:

[IQ DQS i-]Qsh OF a

STATION PROCEDURES IMPACTED, IF SO CONTACT SYSTEM MANAGER 2 CDs INCORPORATED (IF ANY):

DESCRIPTION OF CALCULATION REVISION (IF APPL.):

Revision 1 revised the Containment Spray coverage from 90% to 75%, eliminated Containment Sump Temperaturn, and brought the Recirculation Spray FRow, & ESF leakage discussion up to date. Editorial enhancements were made throughout.

PURPOSE:

The purpose of this calculation is to determine the following 8 Input Parameters for the Salem Alternate Source Term Dose Calculations: 1) MSSV release velocity, 2) MSSV release points for various accidents,

3) N/A, 4) Containment Spray coverage volume & Recirculation Spray flow, 5) Control Room Envelope in-leakage, 6) ESF leakage, 7) Auxiliary Building Charcoal I HEPA assumptions, and 8) Maximum Letdown Flow Rate.

CONCLUSIONS:

For purposes of the AST dose calculations the following values will be used:

1. MSSV release velocity Is 448 m/sec.

2.

MSSV release locations are conservatively specified In Section 3.2.

3. N/A

4. Containment Spray Coverage Is 75% containment free volume and Containment Recirculation Spray Flow is 1900 gpm (sufficient to credit 75% spray coverage during the recirculation phase).

5. The Control Room Unfiltered In-leakage value used will be 150 scfm.

6. ESF leakage value used w'I! be 1.0 gpm.

7. Auxiliary Building (ABV) Exhaust Charcoal and HEPA will not be credited.

8. Maximum Letdown Flow Rate is 165 gpm.

PrInted Name I SIgnature l Date ORIGINATOR/COMPANY NAME:

Barry L. BarkleyI/PSEG (3 tL(pj=l 1-23-2004 PEER REVIEWER/COMPANY NAME:

John A. Rowey I PSEG (Section 3.4.2 only) m 1-1-2004 VERIFIERICOMPANY NAME:

JohnF. Duffy/PSEG /. 3_^,#

1-,23-2004 PSEG SUPERVISOR APPROVAL:

Paul J. Lindsay / PSEG lt f

i-2004 J

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY

'20040907 l

CALCULATION CONTINUATION SHEET l

Page: 2 of 15 CALC. NO.: S-C-ZZ-MDC-1987

REFERENCE:

N/A ORiGINATOR, DATE I REV:

B. Barkley, 1/22/04 Rev. 1 l

REVIEWERNERIFIER, DATE J. Duffy, 1/22/04 REVISION HISTORY Revision Description 0

Original Issue I

Containment Spray coverage was changed from 90% to 75%. Containment Sump Temperature results were eliminated, and Recirculation Spray Flow values were brought up to date. ESF leakage was changed from 0.6 to 1.0 gpm. The content of the Attachments is now completely covered In the text and they are all deleted. Editorial enhancements were made throughout.

TABLE OF CONTENTS Section Sheet No.

Cover Sheet 1

Revision History 2

Table of Contents 2

1.0 Purpose 3

2.0 Scope 3

3.0 Design Bases, Assumptions, & Results 3

3.1 MSSV Release Velocity 3

32 MSSV Release Points 5

3.3 N/A 9

3.4 Containment Spray Coverage & Recirculation Spray Flow 9

3.5 Control Room Envelope In-leakage 13 3.6 ESF Leakage 14 3.7 Auxiliary Bldg Charcoal / HEPA 14 3.8 Maximum Letdown Flow Rate 14 4.0 Conclusions 14 5.0 References 15 6.0 Affected Documents 15 I

I Nuclear Common Revision 9 I

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY

'20040907 CALCULATION CONTINUATION SHE Page: 3 of 15 CALC. NO.: S-C-ZZ-MDC-1987

REFERENCE:

N/A ORIGINATOR, DATE I REV:

B. Barkley, 1/22/04 l Rev. 1 l l

REVIEWEIRNERIFIER, DATE J. Duffy, 1t22/04 1.0 Purpose The purpose of this calculation is to determine the following 8 input parameters for the Salem Alternate Source Term Dose Calculations: 1) MSSV release velocity, 2) MSSV release points for various accidents, 3) N/A, 4)

Recirculation spray flow, 5) Control Room Envelope in-leakage, 6) ESF Leakage, 7) Auxiliary Building Charcoal I JIEPA, and 8) Maximum Letdown Flow Rate 2.0 Scope The Salem Alternate Source Tenn (AST, Reg. Guide 1.183, Ref. 5.4) Phase 2 Project will perform dose calculations for a full scope AST implementation (of Post-LOCA and other accidents). The project will support accomplishing the following objectives:

I. Support Main Control Room (MCR) Habitability issue (unfiltered inleakage) in conjunction with the Generic Letter 2003-01, "Control Room Envelope Habitability" (issued 6-12-03) response to the NRC

2. Allow MCR Tracer Gas Testing (CROD required for the test) to resolve MCR inleakage

3. Increase the allowable ESF leakage outside containment to 1.0 gpm

4. Evaluate the acceptability of new AST doses (Control Room & Offsite) without taking credit for Auxiliary Building Ventilation System (ABVS) charcoal and HIEPA filtration This calculation will assemble information and perform simple calculations to provide some of the input values needed for the various AST dose calculations.

3.0 Design Bases, Assumptions, & Results 3.1 MSSV Release Velocity MSSV release velocity in units of m/sec is necessary to perform X/Q calculations for secondary side releases.

The releases from the steam generators are provided in terms of total mass (pounds) over a time period of interest by the Nuclear Fuel Section calculation of Reference 5.1. This section determines the appropriate MSSV exit velocity based on the Reference 5.1 information.

3.1.1 Main Steam Valve Operational and Design Characteristics Steam Valves: Each of the 4 steam generators has a single steam line that exits the containment. The steam passes through a flow restrictor before passing through the containment boundary. Then outside containment, the steam line has 5, self-actuated spring-loaded safety valves (SVs, 1070 to 1125 psig), one pneumatically operated Atmospheric Relief Valve (MS-I0), and one MSIV per steam-line (Refs. 4 & 5). There are Steam Dump valves to the condenser, which are isolated by the MSIVs. The MSIO is automatically (1015 psig) and manually operated and the 5 safety valves cannot be manually operated. All our calculations (the offsite dose sub-section of Chapter 15) are done assuming only the fully safety-related components are available. Because of the assumptions in design basis calculations, a Loss of Offsite Power (LOOP) and seismic or "Q" pedigree is required for accident mitigation.

The MS-1Os are not fully "Q" and are not considered in the analysis. Since they are diesel backed and are operated according to the EOPs, they need to be considered if their opening could result in higher offsite or control room dose. On a LOOP, the MSIVs and the Condenser Steam Dump Valves are closed or remain closed and the condenser is unavailable (a conservative assumption for dose calculations).

I I Nuclear Common Revision 9

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20040907 CALCULATION CONTINUATION SHEET Page: 4 of 15 CALC. NO.: S-C-ZZ-MDC-1987

REFERENCE:

N/A ORIGINATOR, DATE I REV:

B. Barkley, 1/22104 Rev. 1 l l

REVIEWERNERIFIER, DATE J. Duffy, I122/04 Valve Release Area: The MS-I Os are 6-inch valves with 18-inch discharge vent pipes (Ref. 5.5,5.6, & 5.7). The safety valves have 8-inch lines leading to them and are 10-inch valves. The MSSVs discharge horizontally into a 10-inch 900 elbow, which turns vertically for a short distance and makes a square edged entry into an 18-inch vent pipe. Both the MS-lOs and each of the MSSVs have 35-foot long 18-inch diameter discharge vent pipes with two 45° elbows. The 10-inch MSSV and the 6-inch MS-10 discharge steam flow will expand to conditions of an 18-inch pipe before being released vertically to the atmosphere. The MSSV and MS-10 discharge plume calculations will use a vertical release diameter of 18-inches (pipe I.D. is 17.25 inches).

MSSV Discharge Flow: The safety valve opens when the upstream setpoint is reached. These MSSVs open at pressures in the range of 1070 to 1125 psig. The Tech Spec total relieving capacity for all 20 MSSVs (TS Bases 3/4.7.1.1) is 16.66 E6 (lbs Air), which is 110.3% (for Unit I) and 110.4% for (Unit 2) of the maximum calculated steam flow of 15.10 E6 lbs/hr (for Unit I) and 15.08 E6 Ibs/hr (for Unit 2). Therefore a nominal value for steam flow for a single MSSV is 16.66 E6 + 20 valves or 833,000 lbs/hr (8.33 E5 Ibs/hr). The MSSV relieving capacity of the lowest setpoint valve (1070 psig) is 806,723 Ibs/hr and the highest setpoint valve (1125 psig) is 847,644 lbs/hr (Reference 5.3). Since the valves are only opened at or above their setpoint pressure and are relieving at a rate based on the actual SG pressure, the relief capacity will always be equal to or greater than their setpoint capacity. The total mass (in Ibs) of steam that is released over the time period of the transient / accident is given in Reference 5.1.

Based on Reference 5.1 input values, the Vertical Exit Velocity in meters per second (r/sec) is calculated for the nominal and minimum mass flow rates. Since the total mass released In a given time period is specified, the MSSVs will be intermittently open based on their upstream pressure. The lowest exit velocity calculated below is the more conservative value for use in determining plume XJQs.

3.1.2 MSSV Release Velocity Calculation The following determines the MSSV release velocity in units of m/sec for the lowest set-point MSSV:

Lowest MSSV pressure setpoint

= 1070 psig = 1085 psia MSSV mass flow rate (Ibn/hour)

= m = 806,723 (lb,,/hr) [from paragraph above) = 224.1 lb,,/second Enthalpy upstream of the MSSV

= 1189.7 Btu/ Ibm [Ref. ASME Steam Tables Fig. 12 5d' Ed.]

Vent pipe is 18-inch diameter, I.D. - 17.25-inches [see Valve Release Area Paragraph & References above]

Vent pipe exit flow area, A

= X d2/ 4 = 1.623 ft2 The exit velocity will be sub-sonic, sonic, or super-sonic. Super-sonic is ruled out by the square edged configuration of the MSSV discharge piping (see Valve Release Area Paragraph & References above). That Is, a diverging nozzle is required for the exit flow to be super-sonic. The exit flow will be sonic based on the following statement:

Exit velocity will be sonic ifpc

= > 14.7 psia [p, is the pressure at the exit of the vent pipe]

Mass velocity pa VE m/A - 224.1 Ibm/s + 1.623 ft2 138.1 lb,/ft2s where v

specific volume p

density v

velocity E

- Subscript for the condition of the process at the vent pipe exit Energy balance hE + vE2t12&J -

1189.7 Btu/ lb.

where ge

= 32.2 ft-lb,,,/s2 lbf, conversion factor J

= 778.0 ft-lbf perBtu, conversionfactor By trial and error select yE 1470 fl/sec [the velocity at the vent pipe exit]

ThenhE 1189.7fBtu/lbex

[14702 /2x32.2x778] - 1146.63Btu/lb, I Nuclear Common Revision 9 1 I NulearCommn Re~s~o 9

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20040907 I

CALCULATION CONTINUATION SHEET I

Page: 5 of 15 CALC. NO.: S-C-ZZ-MDC-1987

IREFERENCE:

N/A ORIGINATOR, DATE IREV:

B. Barkley, 1/22/04 l Rev. 1 l I

l REVIEWERNERIFIER, DATE J. Duffy, 1/22/04 1

At 1470 ft/sec the following conditions exist:

PE = PC va / v0 = 138.1 lbtfts / 1470 ft/sec = 0.09395 lb./ft3 Therefore:

v I

= I/pg = 10.64 ft/lb.

At hF = 1146.6 Btu/ ib. and vE = 10.64 fl?/lbm the following (pi) can be found from the steam tables by an iterative process [Ref. ASME Steam Tables Fig. 12 5 h Ed.]:

P=

38.5 psia, which is > 14.7 psia Therefore the exit velocity will be sonic.

Therefore the minimum MSSV Vertical Velocity

= 1470 ftl/sec

= 448 nrnsec Note:

The MSSV minimum vertical velocity Is only needed to show that it is greater than 5 times the 95k-percentile wind speed. That is the requirement of Reg. Guide 1.194 (Ref. 5.12) to simplify the x/Q value for the buoyant plume rise associated with the energetic releases from steam relief valves. For the Salem X/Q calculation the 95'thpercentile wind speed is 16.5 rn/sec (Ref. 5.13). The factor of 5 (5 x 16.5 rn/see) gives 83 m/sec. Therefore the minimum MSSV vertical release velocity of 448 m/sec is greater than 83 m/sec. Therefore, the other MSSVs that release at higher pressure will be close to the 448 rn/sec value and need not be calculated individually since all values will be much greater than 83 rn/sec.

3.2 MSSV Release Points The UFSAR Chapter 15 Accident Analysis Sections that apply to the Salem AST Phase 2 Dose calculations are discussed below. The AST dose calculations for certain Chapter 15 Accident Analyses address radioactive releases through the MSSVs. These AST dose calculations utilize the total mass released over specified time intervals. The MSSVs open when steam generator pressure reaches their setpoints and then reset when pressure decreases. Table I below shows the specified time intervals and quantifies the fraction of time that an MSSV will be open in the stated time interval. The fraction of time that an MSSV is open is based on the average lift pressure of the five MSSVs per steam line and is for information only. This section defines what steam releases are to be assumed in the AST dose calculations. In all cases the release is assumed through the MSSVs and the reason MS-lOs and steam dump to the condenser is stated.

Transient End Point: The following transients end with the unit proceeding to hot shutdown Mode 4. Getting to the RHR cut-in time of 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />, which ends the steam release to the atmosphere, bounds the steam release transient (Ref 5.1, Page 6 of 25).

3.2.1 UFSAR 15.4.1 LOCA:

This accident is the RCS double-ended cold leg guillotine pipe rupture transient. No MSSV, MS-10, or steam dump operation occurs.

3.2.2 UFSAR 15.4.2 Main Steam Line Break Accident (MSL Break):

For the dose calculation sub-set of this UFSAR accident, the plant is operating at full power with the steam line break occurring outside containment and upstream of the MSIV.

MSSV Operation: Assume the steam line fails inside the Auxiliary Building. This is upstream of the MSIV and outside of Containment and represents the worst case for offsite and CR dose. Assume the failed steam line is the one that is on the Auxiliary Building side of containment such that the blow out panel that relieves I Nuclear Common Revision 9 l I Nuclear Common RevIsIon 9 I

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20040907 CALCULATION CONTINUATION SHEET Pago: 6 of 15 CALC. NO.: S-C-ZZ-MDC-1987

REFERENCE:

N/A ORIGINATOR, DATE I REV:

B. Barkley, In2/04 I Rev. 1 l l

REVIEWERNERIFIER, DATE J. Duffy, 1/22/04 the steam is closest to the CR Outside Air Intake. The SG with the failed steam line will completely boil dry within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (Ref 5.1). For the dose calculation a conservative assumption should be made that half of the Intact SG MSSV release is through the I of 3 sets of MSSVs nearest to the Auxiliary Building.

LOOP & MS-I 0s: For the dose calculation a LOOP is assumed. If a LOOP was not assumed, the operators would be operating the steam dumps to the condenser, which would not be limiting for offsite and CR dose.

If the other three SG MS-lOs were operable, the operators would open them equally to keep the SGs and RC Loops operating uniformly and to get the reactor to cold shutdown. This would minimize MSSV operation and move 2/3rds of the release further from the CR ventilation intakes. As the transient progresses with lower SG pressures, the MS-I Os would create plumes with less exit velocity, however, it is reasonably conservative to analyze the MSL Break Accident without MS-I 0 operation.

3.2.3 UFSAR 15.4.4 SteamGenerator TubeRuptureAccident(SGTR):

This accident is the complete severance of a single tube in one of the four steam generators.

MSSV Operation: Assume the failed SG is the one with MSSVs closest to the CR Outside Air Intake. The operators lower RCS pressure rapidly to get below the lowest MSSV setpoint (1070 psig) of the failed SG (200F subcooling). The 3 Intact SGs MSSVs would operate uniformly. For the dose calculation a conservative assumption should be made that half of the intact SG MSSV release is through the I of 3 sets of MSSVs nearest to the Auxiliary Building.

LOOP & MS-l0s: For the dose calculation a LOOP is assumed. If a LOOP was not assumed, the operators would be operating the steam dumps to the condenser, which would not be limiting for offsite and CR dose.

If the other three SG MS-I Os were operable, the operators would open them equally to keep the SGs and RC Loops operating uniformly and to get the reactor to cold shutdown. This would minimize MSSV operation and move 2/3rds of the release further from the CR ventilation intakes. The operators would still try to minimize MS-10 operation since the core is assumed to have failed fuel and there is some RCS leakage assumed into the intact SG (SG tube leakage) and this radioactivity would be released to the atmosphere. As the transient progresses with lower SG pressures, the MS-10s would create plumes with less exit velocity; however, it is reasonably conservative to analyze the SGTR Accident without MS-10 operation.

3.2.4 UFSAR 15.4.5 Reactor Coolant Pump Locked Rotor Accident (RCP-LR):

This event is the postulated instantaneous seizure of a reactor coolant pump rotor with flow through the affected reactor coolant loop rapidly reduced. The UFSAR Accident Analysis (Section 1 5A.5.3) assumes some fuel clad damage occurs. In the subsequent transient all four steam generators equally release the flow provided in Reference 5.1.

MSSV Operation: For the dose calculation a conservative assumption should be made that 2/3rds of the SG MSSV release is through the 2 SG MSSVs closest to the CR Outside Air Intake.

LOOP & MS-I Os: For the dose calculation a LOOP is assumed. If a LOOP was not assumed, the operators would be operating the steam dumps to the condenser, which would not be limiting for offsite and CR dose.

If the other three SG MS-I Os were operable, the operators would open them equally to keep the SGs and RC Loops operating uniformly and to get the reactor to cold shutdown. This would minimize MSSV operation and move 2/3rds of the release further from the CR ventilation intakes. The operators would still try to minimize MS-1O operation since some fuel failure occurs and there is some RCS leakage assumed into the intact SG (SG tube leakage) and this radioactivity would be released to the atmosphere. As the transient progresses with lower SG pressures, the MS-lOs would create plumes with less exit velocity; however, it is reasonably conservative to analyze the RCP-LR Accident without MS-10 operation.

I Nuclear Common Revision 9 I

I Nuclear Common RevisIon 9 I

I

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY

• 20040907 l

CALCULATION CONTINUATION SHEET Page: 7 of 1 CALC. NO.: S-C-ZZ-MDC-1987

IREFERENCE:

N/A ORIGINATOR, DATE I REV:

B. Barkley, 1/22/04 l Rev. 1 l

REVIEWERNERIFIER, DATE J. Duffy, 1/22/04 3.2.5 UFSAR 15.4.7 Control Rod Ejection Accident (CRE):

This event is the postulated mechanical failure of a control rod mechanism pressure housing resulting in the ejection of a RCCA and drive shaft. In the subsequent transient all 5 of each of the four steam generator MSSVs (20 total) lift to handle the power excursion transient and some fuel damage occurs. The entire contents of the 4 SGs is assumed to be discharged.

MSSV Operation: All 5 MSSVs per loop open for the short time duration of the MSSV steam release transient. For the dose calculation a conservative assumption should be made that the 2 SGs with MSSVs closest to the CR Outside Air Intake blow down 2/3rds of the flow.

LOOP & MS-lOs: This SG release transient is over quickly with the steam release only 103 seconds (Unit 1) and 110 seconds (Unit2) induration(Reference 5.1,Page 10 of 25). Forthe dose calculation a LOOP is assumed (no steam release to the condenser) and no MS-10 operation is assumed during the release.

3.2.6 UFSAR 15.3.1 Instrument Line Break Accident:

This event is the postulated small line break of an instrument line connected to the RCS. No MSSVs operate in this accident.

3.2.7 UFSAR 15.3.6 VCT & WGDT Rupture Accident:

This event is the postulated rupture of the Volume Control Tank (VCT) and the Waste Gas Decay Tank (WGDT). No MSSVs operate in this accident.

5 l Nuclear Common Revision 9 lI N

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY

• 20040907 CALCULATION CONTINUATION SHEET Page: 8of15 CALC. NO.: S-C-ZZ-MDC-1987

REFERENCE:

N/A ORIGINATOR, DATE REV:

B. Barkley, 1/22/04 Rev. I l I

REVIEWER/VERIFIER, DATE

3. Duffy, 1/22/04 Table 3.2 Post-Accident Main Steam Safety Valve Releases Note: The purpose of this table is to address MSSV steam release that is depicted differently in References 5.1 and 5.2 for the input to Radiological Dose Analysis. This table shows that the MSSVs will be cycling (assuming the average lift pressure of the 5 MSSVs per steam line) and quantifies the fraction of time they will be open in the stated time interval.

Total Mass Nubro lwPrFraction of the Numbr of FlowPer Time Interval Accident Scenario Released by Active Valve that an MSSV MSSVs MSSVs (k-lbs/hr) is Open Accident Time Interval (k-lbs)

(Fr RcfS.,

(From Discussion (mass I number, flow, (From Ref5.1)

Table4)

In 3.1.1 Abovc)

& the timc Interval)

LOCA Loss of Coolant Accident l

None N/A J

N/A N/A Loss of Non-Emergency AC Power 4 SGs: 0-2 hours (2hrs) 655 4

833 0.098 4 SG: 2-8 hours (6 hrs) 540 4

833 0.027 4 SG: 8 - 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> (24 hrs) 2400*

4 833 0.030 Main Steam Line Break Accident One Faulted SG: 0-2 hours (2 hrs) 128 None: Rupture Panel N/A N/A 3 Intact SGs: 0-2 hours (2 hrs) 500 3

833 1

0.100 3 Intact SGs: 2-8 hours (6 hrs) 452 3

833 0.030 3 Intact SGs: 8 - 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> (24 hrs) 2008**

3 833 J

0.033 Steam Generator Tube Rupture Release One Faulted SG: 0-30 min (0.5 hrs) 56.5 1

833 0.136 3 Intact SGs: 0-2 hours (2 hrs) 465 3

833 0.093 3 Intact SGs: 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (6 hrs) 1055 3

833 0.070 3 Intact SGs: 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (16 hrs) 1503 3

833 0.038 3 Intact SGs: 24 - 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> (6 hrs) 477 3

833 0.032 3 Intact SGs: 30 - 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (6 hrs) 451 3

833 0.030 RCP Locked Rotor Accident 4 SGs: 0-2 hours (2 hrs) 655 4

833 0.098 4 SGs: 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (6 hrs) 1 540 4

833 0.027 4 SGs: 8 - 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> (24 hrs) 2400*

4 833 0.030 Control Rod Ejection Accident 4 SGs: 0 - 110 seconds (0.0306 hrs) 5,120 20 833 1.00 OtherAccidents Letdown Line Break Accident None I

N/A I

N/A I

N/A Waste Gas Tank Rupture Accident None N/A N/A N/A

• 2400 k-lbs - 3602 k-lbs x (32 hrs - 8 hrs)/ (48 hrs - 8 hrs) x 1.1106 [Ref. 5.1, Page 6 of 25]

• • 2008 k-lbs = 3013 k-lbs x (32 hrs-8 hirs) /(48 hrs-8 hrs) x 1.1106 [Ref. 5.1, Page 8 of25]

Nuclear Common Revision 9

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY

• 20040907 CALCULATION CONTINUATION SHEET Page: 9 of 1!

CALC. NO.: S-C-ZZ-MDC-1987

REFERENCE:

N/A ORIGINATOR, DATE I REV:

D. Barkley, 1/22/04 I Rev. 1 R REVIEWERIVERIFIER, DATE J. Duffy, 1/22/04 3.3 N/A 3.4 Containment Spray Covcragc & Recirculation Spray Flow 3.4.1 Reference Value for Containment Spray Coverage 3.4.1.1 The Original 90% Spray Covcrage: The Westinghouse Report, WCAP-7952, 'Iodine Removal by Spray in the Salem Containment," R.M. Kemper, August 1972, (Reference 5.9) provides the original basis for Containment Spray Coverage. The 'estinghouse mathematical model of WCAP-7952 considered various aspects of the spray droplet interaction with iodine vapor and steam condensation and compared results with experimental data. Phone conversations were held on April 1, 2003 and during the November 18 through 21, 2003 time frame with Westinghouse Containment & Radiation Analysis Branch personnel. The phone conversation discussed the present validity and conservatism of the old WCAP-7952 Report for use of Salem Containment Spray lambda (exponential iodine removal coefficient) and spray coverage. The calculated containment spray coverage is presented in Table 4-2 of the WCAP as 90% (unsprayed volume of 10%/0). The value of 90% could not be easily verified by a search of the WCAP background information by Westinghouse. If the volumes (computed from the areas and fall heights) were summed in Table 4-1, the spray coverage would be 77.9% (based on a containment free volume of 2.5 x 106 cu ft). If the volumc above the average spray header elevation (the 4 sub-volumes from Ref. 5.10 - 6.5%) were added, the spray coverage would be 84.4% still short of 90%. Therefore the original WCAP-7952 spray coverage of 90%

cannot be easily validated.

3.4.1.2 The Conservative 60% Spray Coverage: The Stone & Webster Calculation VTD 321038 (Reference

5. 10) was commissioned in 1996 in order to derive the containment spray coverage. The S&W calc used a conservative technique of doing a volumetric depiction of spray coverage in the containment and determined that spray coverage was 60%. The spray pattern was taken from manufacturer's data at "ambient" atmospheric conditions. The coverage was then based on a proprietary Stone & Webster trajectory compression multiplier (M) at the containment temperature and pressure condition between 90 seconds and 48 minutes after the LOCA.

3.4.1.3 The 60% Spray Coverage is Unnecessarily Conservative:

Geometry vs. Iodine Removal: The S&W Calc discussed above uses a conservative volumetric depiction of spray coverage in the calculation. This technique does not recognize the extensive turbulence in the sprayed and unsprayed air interaction process. The WCAP discussed above specifically considers various aspects of the spray droplet interaction with iodine vapor. The WCAP also included the effects of drop size by considering the complete spectrum of drop sizes emitted by the containment spray nozzles, condensation of steam on the spray drops, coalescence of drops, and the mass transfer resistance in the liquid phase of the drops. These considerations lead to the iodine (fission product of primary concern) removal effectiveness.

The WCAP spray coverage was considered geometrically to be 100% coverage in the spray drop zone regions where there were no equipment interferences. The S&W calculation determined this zone to be less than 100% and arrived at 60% spray coverage using a purely geometric technique without crediting the assumptions of WCAP-7952.

Timing of Release WRT Containment Response: The 60% coverage was based on a proprietary Stone &

Webster trajectory compression multiplier (M) at the time-weighted average containment temperature and pressure condition between 90 seconds and 48 minutes after the LOCA. With the new AST methodology the timing of the release of fission products occurs later after the LOCA compared to the TID-14844 (Ref. 5.17) methodology where the release is at time zero and the spray effective time ends at 48 minutes. With AST the gap release phase (about 5% of the core inventory) begins at 30 sec. and lasts 0.5 hr. and the early in-vessel 5

l Nuclear Common Revision 9 l

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20040907 CALCULATION CONTINUATION SHEET Page: 10 of 15 CALC. NO.: S-C-ZZ-MDC-1987

REFERENCE:

N/A ORIGINATOR, DATE REV:

B. Barkley, 1t22/04 Rev. 1 REVIEWERNERIFIER, DATE

1. Duffy, 1122/04 release phase (the remainder of the release) begins at 0.5 hr. and lasts 1.3 firs. Therefore the iodine release is not complete until 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident. With this timing change from TID-14844 to AST the decreasing containment temperature and pressure would increase the spray effectiveness areas. If the S&W Cale was revised to reflect the timing of the AST release the spray coverage would increase.

Treatment of Space in the Dome: The space in the top of the containment dome (above the Spray Headers) is considered unsprayed by the S&W Calc for two reasons. First the spray trajectory above the spray headers in the dome area is unnecessarily compressed due to the timing assumptions discussed above. Secondly, the sequence of events is not considered, which is more pronounced with AST methodology. Phone conversations were held on April 1, 2003 and during the November 18 through 21, 2003-time frame with Westinghouse Containment & Radiation Analysis Branch personnel and on April 3, 2003 with Westinghouse Fluids System Branch personnel. The phone conversation discussed the fact that it is reasonable to assume that the fission products must travel through sprayed volumes to reach the top of the containment (see next paragraph below). After the rapid initial containment pressure transient has peaked, the only flows into the dome volume are gases replacing steam condensing in the dome or gases transferred there due to turbulence between the dome and lower volumes. Regardless, when the significant fission products begin to enter the lower region of containment from the LOCA the containment pressure has already peaked and the sprays are already operating.

Timing of Release WRT Containment Spray: The conservative timing specified by Reg. Guide 1. 183, Table 4 (Rcef. 5.4) has the large break LOCA release beginning at 30 seconds (onset of the gap release phase that releases -5% of the core halogen inventory) and at 30 minutes (onset of the Early In-Vessel Phase that releases the major fraction of core inventory). The actuation time of containment spray is 35 to 85 seconds (UFSAR Table 15.4-3 & S-C-VAR-NZZ-0020 Page 95 of 2 10). Therefore a portion of the release could begin after the pressure transient but before the start of containment spray; however, this would be very small.

This small amount of halogens would be negligible compared to the total halogen release. It is reasonably conservative to assume the fission products do not reach the dome region without passing through spray.

Therefore the dome region is included in the sprayed volume given below.

Note:

The transition from the Injection Spray Phase to the Recirculation Spray Phase is to be done by procedure without spray interruption.

3.4.1.4 The Reconstituted 75% Spray Coverage: The calculation below uses the volumes calculated in the S&W calc except that a substituted volume is used for the sprayed volume. The substituted sprayed volume credits the spray coverage as complete (like was done in WCAP-7952, Table 4-1) in the open sprayed sub-volumes below the spray headers and above the operating floor. It also credits the sub-volumes above the reach of the upward directed sprays at the top of the containment as indicated in WCAP-7952, Table 4-2 (10% Unsprayed Volume, i.e., 90% coverage). As discussed above the sub-volumes In the dome area can be credited consistent with the assumption that the fission products must pass through sprayed volumes to reach the top of the containment. The spray coverage then is as follows:

Gross sub-volume above the operating floor (El 130')

2,073,031 n3 (Table I Page 20, Ref. 5.10)

Minus total occupied volume above the operating floor 43,537 ft (Step 5, Page 26, Ref. 5.10)

Plus sprayed volume below the operating floor

+ 59.387 e (Step 6, Page 29, Ref. 5.10)

Gross volume that is sprayed 2,088,881 ft' The containment net free volume 2,605,746 ft' (Step 8, Page 32, Ref. 5.10)

Calculated Spray Coverage 80.2%

Recommended Spray Coverage Value 75%

(For added conservatism)

I Nuclear Common Revision 9 l I Nuclear Common RevIsbn 9 I

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY

-20040907 l

CALCULATION CONTINUATION SHEET Page: 11 of 15 CALC. NO.: S-C-ZZ-MDC-1987 I

REFERENCE:

N/A ORIGINATOR, DATE I REV:

B. Barkley, 1/22/04 l Rev. I l l

REVIEWERNERIFIER, DATE J. Duffy, 1/22104 3.4.2 Reference Value for Recirculation Spray Flow:

3.4.2.1 Westinghouse Containment Analyses: The Westinghouse Containment Analysis done on March 6,2002, for the Phase 2 Containment Capability Study (Reference 5.8) and the CFCU Project letters (References 5.14 &

5.15) provide the basis for the minimum value of Recirculation Spray Flow. These values for recirculation spray flow arc the most accurate available for present Salem Units 1 & 2 independent of the CFCU Project. These will result in a formal WCAP with final values in early 2004 (Ref. 5.16: WCAP-16193, "Salem Unit I and Unit 2 Containment Response to LOCA and MSLB for Containment Fan Cooler/Service Water System Enhancement Project"). The letter, Ref 5.15 Attachment 3,4, gives the Containment Spray Flowrate during the recirculation phase with minimum safeguards (1974.8 gpm) and maximum safeguards (1181.7 gpm). The minimum recirculation spray values of 874.7 gpm for Unit I and 817.5 gpm for Unit 2 that are shown in Reference 5.8 are recirculation spray flow cases where there is I RIIR Pump that fails to restart and there are no other safeguards failures. That is, all safeguards pumps start and the operators align safety injection on that basis. Then when the RHR Pumps are shut down and restarted for the recirculation phase, one fails to start and the operators do not re-align the injection flow path. If containment pressure was still high at this point, the operators would be procedurally driven into addressing containment integrity and would establish higher containment spray flow by realigning the injection flow as directed by the EOPs. These cases are summarized as follows:

I Rolrultln pryMin Recirc ApplIcability to AST Recirculation Spray Spray Flow Calc with Catastrophic Reference Case Fuel Failure Min Safeguards 1974.8 gpm Applicable 5.15 (EDG failure)

Max Safeguards 1181.7 gpm Not Applicable 5.15 Max Safeguards with failure 874.7 gpm (UI) of one RHR Pump to restart 817.5 gpm (U2)

NotAppicable 5.8 3.4.2.2 Minimum Recirc Sprav Flow for AST: For purposes of the AST dose calculations, the minimum containment recirculation spray flow is >1900 gpm and is sufficient to credit the function of sprays for fission product removal from the containment air space. When maximum safeguards are available (Si & ECCS Charging Pumps) less RHR Pump flow is available for the recirculation spray. The AST dose calculation needs a basis for considering that the Containment Spray Coverage is 75% during the Recirculation Spray phase of the accident and the value of recirculation spray flow. Phone conversations were held on April 1, 2003 and between November 18 and 21,2003 with Westinghouse Containment & Radiation Analysis Branch personnel and on April 3,2003 with Westinghouse Fluids System Branch personnel. The discussions were based on the fact that recirculation spray is now more important in dose analysis due to the later release timing of AST compared to TID and supported the following conclusions. Westinghouse Containment & Radiation Analysis Branch personnel stated that a Salem recirculation spray flow value greater than 1200 gpm should provide adequate spray coverage. The minimum flow value given in Reference 5.15 of 1974.8 gpm is the minimum recirculation spray Diesel Failure flow that is applicable to the Post-LOCA AST dose evaluations. The maximum safeguards values shown above are not considered to be applicable to the AST analysis. They represent recirculation spray flow cases where there are no safeguard pump failures or where I RHR Pump fails to restart after all safeguards pumps had initially started. These maximum safeguards cases result in less flow available for containment spray. It is not credible to have a catastrophic corefailure (the basis of ASTcalculatlons) if all ECCS (other than I RIHR Pumpfailing to restart) is operating normally. Therefore, the Diesel Failure case is considered a reasonably conservative minimum containment recirculation spray flow value.

Nuclear Common Revision 9 I

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20040907 l

CALCULATION CONTINUATION SHEET Page: 12 of 15 CALC. NO.: S-C-ZZ-MDC-1987

REFERENCE:

N/A ORIGINATOR, DATE l REV:

3B.

Barkley, 1/22/04 l Rev. I l l

REVIEWERNERIFIER, DATE J. Duffy, 1/22104 Note: The Chapter 15 LOCA thermal-hydraulic analysis (when 10 CFR 50.46 acceptance criteria must be met) results in limited fuel damage, which will always be less than that assumed for the AST DBA-LOCA. The AST DBA-LOCA is the radiological consequences analysis that assumes catastrophic fuel damage (100% of the total core noble gases released, 40% of the total core halogens released, et.al.).

Note: The containment pressure and temperature decreases from the post-LOCA maximum value over the time period that Containment Spray is operating. When Recirculation Spray begins the containment temperature and pressure has already decreased from the maximum value and therefore the spray pattern characteristics will be enhanced. This is due to better nozzle AP and less trajectory compression due to the smaller elevated temperature and pressure condition during recirculation spray (Ref. 5.10, Page 4).

3.4.3 Containment Spray Coverage Results: For purposes of the AST dose calculations, the Containment Post-Accident Spray Coverage is as developed above. The Sprayed Volume is 75% and the Unsprayed Volume is 25%.

3.4.4 Reclrculatlon Spray Flow Results: For purposes of the AST dose calculations, the Containment Recirculation Spray Flow is >1900 gpm from the end of Containment Spray through the critical 3 to 4 (or 5) hours of Recirculation Spray after a DBA-LOCA. This flow rate (if greater than about 1200 gpm) is sufficient to credit the function of sprays for fission product removal from the containment air space.

I Nuclear Common Rovialon 9 I N

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY

-20040907 CALCULATION CONTINUATION SHEET Page: 13 of 15 CALC. NO.: S-C-ZZ-MDC-1987

REFERENCE:

NIA ORIGINATOR, DATE I REV:

B. Barkley, 1/22/04 I Rev. 1 I

I I

REVIEWERIVERIFIER, DATE J. Duffy, 1/22/04 l

3.5 Control Room Envelope In-leakage 3.5.1 Control Room Habitability: TheNRCGenericLetter2003-01, "Control Room Habitability,"was issued June 12,2003 (in conjunction with 4 associated Reg. Guides). The completed CR-Tracer Gas Test (CR-TGT) combined with LCR S03-005 underway as part of the Salem AST Phase 2 Project addresses significant requirements for this regulatory change. Both hlope Creek and Salem are among the third of the plants that have addressed this issue before GL 2003-01 was issued. The Control Room Tracer Gas Test (CR-TGT) was successfully completed and obtained the favorable results shown here (Ref. 5.1 1):

Table 3.5.1 Salem CR-TGT Results SCREACS Measured In-Maximum Value eCREACS leakage and 95th by Applying the Mode system Percentile Positive Alignment Uncertainty Uncertainty Maint. Press. Mode Unit 1 (with Unit 55 scfm 116 scfm (Stairwell Doors Closed) 2 Intake air)

(+/-61)

Maint. Press. Mode Unit 1 (with Unit 47 scfm 107 scfm (Stairwell Doors Open) 2 Intake air)

(+/- 60)

Maint. Press. Mode Unit 2 (with Unit 95 Scfrn 116 scfm (Stairwell Doors Closed) 1 intake air)

(+/- 21)

Dual Accident Both Unit 1 &

96 scfm187 f

Press. Mode 2 IS (wh Unit I (i 91)

SC intake air)

Fire Outside the Both Unit 1 &

542 acfm 559 acfm CR (Isolation) 2 IIS

(+/- 17) 3.5.2 Conclusion-Use 150 scfm: The values for the Maintenance Mode are 55 scfm (Unit 1) and 95 scfm (Unit 2). Sincethese2 values are less than 100 cfm, theoptionofReg. Guide 1.197,Paragraph C.1.4,"TestResults and Uncertainty" that discusses the difficulty of uncertainty measurement when in-leakage is low can be used. There is an option of not adding the uncertainty value to the CR-TGT results. Nevertheless, 150 scfm will be used in the AST design basis calcs and this conservatively bounds the uncertainty in the Unit I and 2 (single - non-dual filtration) measurements. The dual pressurization data was 96 scefm with 187 scfm at maximum uncertainty and does not cause difficulty in choosing 150 scfin since (as has been shown by sensitivity calcs) double the filtration (2 CREACS charcoal filters) will not be a conservative / limiting dose case. Again the issue of uncertainty at low leakage rates (< 100 scfm) is given recognition in RG 1. 197. (Note: The CR in leakage for the "Fire Outside the CR" Mode of CREACS (up to 559 acfm) does not enter into the radiological dose calcs.)

3.5.3 CROD: CR-TGT results for our present proceduralized Maintenance and Normal Modes are 46 scfm (U-

1) and 102 scfm (dual CREACS) compared to 60 scfm, which is the design input value (cf., UFSAR Section 15.4.1.9 & Table 6.4-3). This makes the Control Room Envelope (CRE) operable but non-conforming as dispositioned by a CROD & CRFA (SAP Order 70031962). This CROD will remain in effect until the TS Amendment is approved and issued by the NRC. The inleakage values of 46 scfm and 102 scfm in the CROD discussion do not match the values shown in Table 3.5.1. The CROD/CRFA used the preliminary test results whereas Table 3.5.1 shows the final QA reviewed values. Since the CROD/CRFA was written so that it is applicable as long as the final values remain below 800 scfln, it is valid as written.

I Nuclear Common Revision 9I-I Nu l a o m nR vso

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20040907 l

CALCULATION CONTINUATION SHEET Page: 14 of 15 CALC. NO.: S-C-ZZ-MDC-1987

REFERENCE:

N/A ORIGINATOR, DATE I REV:

B. Barkley, 1/22/04 l Rev. 1 l l

REVIEWERNERIFIER, DATE J. Duffy, 1/2/04 3.6 ESF Leakage ESF Leakage 1.0 gpm: The amount of leakage from systems outside Containment that could contain highly radioactive fluids after an accident (e.g., ECCS) is referred to as Engineered Safeguards Feature (ESF) Leakage.

The ESF leakage presently allowed is 3500 cc/hr (per SC.SA-AP.ZZ-005 1), which is 92% of the nominal value of I gallon per hour (gph). The value used in the LOCA dose analysis is doubled with a 1.5% margin (7680 cc/hr per UFSAR Table 15.4-SB). The nominal ESF leakage value to be used in the AST design/licensing basis will be changed from -1 gph to 1.0 gpm (from -0.017 to 1.0 gpm). This increase in allowable ESF leakage (which provides a more reasonable operating margin and enables a better operational strategy for Salem Units I & 2) still maintains dose consequences within the regulatory limits.

3.7 Auxiliary Bldg Charcoal / HEPA ABV Exhaust Charcoal & HEPA Removed from TS: The AST Phase-2 Project objective is to eliminate the Auxiliary Building (ABV) Exhaust Charcoal and HEPA filtration from all AST dose analysis and the Tech Specs.

3.8 Maximum Letdown Flow Rate For purposes of the AST dose calculations, the maximum CVCS letdown flow is needed to calculate the Iodine Appearance Rate based on the Technical Specification LCO of Dose Equivalent 1-131 specific activity in the reactor coolant. The value of 165 gpm is recommended as the maximum flow for letdown in the AST analyses. This value bounds any likely actual maximum letdown flow value. Procedurally, letdown flow is presently limited to 13D gpm (indicated). The nominal value is 120 gpm. Westinghouse stated that this value had a variance of 10% (112 to 132 gpm). NUCR 70030162, Activity 0060, states that Engineering is recommending that the maximum assumed letdown flow value be 165 gpm. The conservative value of 165 gpm will be used for the letdown flow in the AST analysis.

4.0 Conclusions For purposes of the AST dose calculations the following values will be used:

I. MSSV release velocity is 448 m/sec

2.

MSSV release locations are conservatively specified

3. N/A

4.

Containment Spray Coverage is 75% containment free volume and Containment Recirculation Spray Flow is 1900 gpm (sufficient to credit 75% spray coverage during the recirculation phase).

5.

The Control Room Unfiltered In-leakage is 150 scfm

6. ESF leakage value used will be 1.0 gpm

7. Auxiliary Building (ABV) Exhaust Charcoal and HEPA will not be credited

8. Maximum Letdown Flow Rate is 165 gpm INuclear Common RevIslon 9 I I Nuclear Common RevIsion 9 I

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY

-20040907 CALCULATION CONTINUATION SHEET Page: 15 of 15 CALC. NO.: S-C-ZZ-MDC-1987

REFERENCE:

N/A ORIGINATOR, DATE I REV:

B. Barkley, 1/22/04 Rev. I 1

REVIEWERIVERIFIER, DATE J. Duffy, 1/22/04 5.0 References 5.1 DSI.6-0453 Nuclear Fuel Section Calculation File, "Determination of Steam Release Flows for Input to the Radiological Dose Analysis," Dated 4-29-03 5.2 VTD 321035 MES Co., "Accident X/Q Values...", Rev. 3, Dated 6-16-98 5.3 S-C-1969-DSP-6422, Sheet 001, "Steam Generator Safety Relief Valves - Main Steam, 'Rev. 3, Dated 7-18-97 5.4 NRC Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 200 5.5 205203, Sh. 1-6 Unit I (& 205303, Sh. 1-6, Unit 2) Main, Reheat, and Turbine Bypass Steam, P&ID 5.6 218249 No. I Unit Containment Main Steam Vent Valve Safety Valve Piping, Rev. 8, Mechanical Arrangement Drawing 5.7 218299 No.2 Unit Containment Main Steam Vent Valve Safety Valve Piping, Rev. 3, Mechanical Arrangement Drawing 5.8 Westinghouse Letter, PSE-02-19, Summary Report for Phase 2 of Containment Capability Study for Salem Unit I and Unit 2, March 6,2002, from M.P. Osborne, Westinghouse to Rob DeNight PSEG, Attachment to PSE-02-19 5.9 WCAP-7952 Iodine Removal by Spray in the Salem Containment, R.M. Kemper, August 1972 5.10 VTD 321038 Stone & Webster-Vendor Technical Document, Containment Spray Coverage Following a LOCA, Rev. 3,6-16-98 5.11 VTD 326043 NCS Corporation / Lagus Applied Technology, Inc., "Control Room Envelope Inleakage Testing, SNGS, 2003" 5.12 NRC Regulatory Guide 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, June 2003 5.13 S-C-ZZ-MDC-1959, ARCON 96 X/Qs for Non-LOCA Releases 5.14 PSEG Letter EA-CFCU-03-004, PSEG Nuclear Response to Westinghouse Input Request for CFCU Project Containment Mass and Energy Release Analyses, July 10, 2003, from Ashok Moudgill, CFCU Project Mgr. PSEG, to Jerold Kusky, Westinghouse, Page 24 of 40 5.15 PSEG Letter EA-CFCU-03-005, Additional clarification for CFCU Project Containment Analyses Input Parameters, October 20,2003, from Ashok Moudgill, CFCU Project Mgr. PSEG, to Jerold Kusky, Westinghouse, Attachment 3 5.16 WCAP-16193 Final Draft Document: "Salem Unit I and Unit 2 Containment Response to LOCA and MSLB for Containment Fan Cooler/Service Water System Enhancement Project" February / March 2004 (Proprietary) 5.17 TID-14844 USAEC TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites,"

J.J. DiNunno et al., U.S. Atomic Energy Commission (now USNRC), 1962 6.0 Affected Documents S-C-ZZ-MDC-1945, Rev. 01R0, Post - LOCA EAB, LPZ, & CR Doses - AST, In Approval Stage S-C-ZZ-MDC-1948, Rev.OIRO, EAB, LPZ, & CR Doses - Rod Ejection Accident (EA) - AST,9-5-03 S-C-ZZ-MDC-1 949, Rev. 01R0, EAB, LPZ, & CR Doses - Steam Generator Tube Rupture (SGTR) Accident - AST, 9-5-03, (Rev. 01RI, In Approval Stage)

S-C-ZZ-MDC-1950, Rev. OIRO, EAB, LPZ, & CR Doses - Main Steam Line Break (MSLB) Accident - AST, 9-5-03 S-C-ZZ-MDC-19511 Rev.OIRO, EAB, LPZ, & CR Doses - RCP Locked RotorAccident (LRA) -AST,9-5-03 Nuclear Common Revision 9