ML20100G888

Four-Minute Isolation of Postulated Steam Line Breaks at Fort St Vrain Nuclear Generating Station

ML20100G888
Person / Time
Site:	Fort Saint Vrain
Issue date:	09/11/1984
From:	Dvorak K, Marquez S PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20100G882	List: ML20100G880 ML20100G888
References
TAC-42527, NUDOCS 8504080463
Download: ML20100G888 (115)

Text

{{#Wiki_filter:.,. FOUR MINUTE ISOLATION OF POSTULATED STEAM LINE BREAKS AT THE FORT ST. VRAIN NUCLEAR GENERATING STATION by SAMUEL L. MARQUEZ KENNETH C. DVORAK 09/11/84 iW BW eng

O ABSTRACT This study was performed in an effort to validate the steam line rupture curves used in the Fort St. Vrain (FSV) environmental qualification program. These curves assume a four minute isolation 1 time for any major steam line rupture in either the Reactor Building or Turbine Building. All possible locations of a major steam leak were analyzed including the condensate, feedwater, main steam, cold reheat, and hot reheat lines. All pertinent automatic and manual actions were studied for each case. All manual actions were determined using the FSV Emergency Procedures. Times for each manual action-were conservatively estimated to allow for longest reasonable response time. The results of each case are listed in tabular form. All postulated line breaks can be isolated in four minutes or less. Worst case accident in the Reactor Building was determined to be a cold reheat line break, which is isolated in two minutes. Worst case in the Turbine Building was determined to be a hot reheat line break, which is isolated in four minutes. Because of the longer termination time found on Reactor Building temperature curves, they are conservative. Because the hot reheat accident has a lower blowdown rate than that used for the Turbine Building curves, these curves, too, are conservative. Conservatisms found in the steam line rupture curves provides additional margin during testing. All previous test results remain valid and all equipment is qualified. l l 11 ]

1 ., o s TABLE OF CONTENTS SECTION 1... INTRODUCTION......................

p. 1-1 SECTION 2... BACKGROUND........................

p. 2-1 SECTION 3... CASE STUDIES......................

p. 3-1 SECTION 4... CONCLUSION.........................

p. 4-1 SECTION 5... REFERENCES.........................

p. 5-1 ILLUSTRATIONS APPENDIX A-CALCULATIONS

p. A-1 APPENDIX B-FSV EMERGENCY PROCEDURES

p. B-1 iii

ILLUSTRATIONS Fig.1..... Temperature Response of the Environment Near the Rupture for a Reactor Building Cold Reheat Pipe Rupture Fig. 2..... Temperature Response of the Environment Near the Rupture for a Turbine Building Hot Reheat Pipe Rupture Fig. 3.... PPS Valve Isolation Following a Steam Pipe Rupture in the Reactor Building Fig. 4.... Trip Time for High Reactor Building Temperature Switches Fi g. 5.... Alarms and Automatic Actions for Steam Line Breaks on Feedwater, Condensate, and Main Steam Lines in the Turbine Building Fig. 6.... Comparison of Feedwater Break Temperature with Turbine Building Steam Line Rupture Test Curves Fig. 7.... PPS Alanns and Actions Following a Hot Reheat Line Break in the Turbine Building Fig. 8.... Postulated Pipe Rupture Downstream of the Low Pressure Switches on the Hot Reheat Line

a I TABLES TABLE 1.... Sequence of Events Following a Feedwater, Main Steam, or Hot Reheat Line Break in the Reactor Bu11 di ng......................... p. 3-1 TABLE 2.... Sequence of Events Following a Cold Reheat Line Break in the Reactor Building........... p. 3-2 TABLE 3.... Sequence of Events Following a Main Steam Line Break in the Turbine Building Downstream of the Bl ock Va1 ves.......................... p. 3-3 i TABLE 4.... Sequence of Events Following a Low Main Steam Pressure Scram............................... p. 3-3 TABLE' 5.... Sequence of Events Following a Feedwater Line Break Downstream of the Block Valves......... p. 3-4 TABLE 6..... Sequence of Events for a Feedwater Line Break Between the Feedwater Check Valves and BFP Discharge Va1ves............................. p. 3-4 TABLE 7.... Sequence of Events Following a Cold Reheat Line Break in the Turbine Building........... p. 3-5 TABLE 8.... Sequence of Events Following a Hot Reheat Lina Break Upstream of the Stop-Check Valves. p. 3-6 TABLE.9.... Sequence of Events Following a Hot Reheat Line Break Between the Stop-Check Valves and Pressure Switches........................ p. 3-7 TABLE 10.... Sequence of Events Following a Hot Reheat Line Break Downstream of the Pressure Switches..................................... p. 3-7 e v

SECTION 1 INTRODUCTION The environmental qualification program for safe shutdown equipment at FSV is based on two sets of curves which represent the time versus temperature response of the environment in the vicinity of the steam line break (Fig. 1 and Fig. 2). These curves were developed assuming the leak would be terminated in four minutes or less. During the FSV preliminary environmental qualification meeting held with Public Service Company of Colorado (PSC) on February 29,

1984, members of the Nuclear Regulatory Commission's Environmental Qualification Branch (NRC-EQB) questioned the ability of an operator to properly identify a high energy line break and to initiate corrective action within four minutes.

This study was conducted in order to substantiate the assumptions made by Gulf General Atomic (GGA) when generating the steam line rupture curves. A case by case analysis was done for various ruptures occurring on the following high energy lines: condensate, feedwater, main steam, cold reheat, and hot reheat. All appropriate actions, both automatic and manual, are used to determine the termination time. All operator actions are based on the existing FSV Emergency Procedures (EPs). 1-1

SECTION 2 BACKGROUND The original basis for environmental qualification at FSV was presented in GGA Report GA-A12045. This report included a set of steam line rutpure curves which were to be used to establish test chamber temperature conditions in subsequent steam line rupture simulation tests. To generate the steam line rupture curves, the computer code CONTEMPT-G was utilized. This code was a modified version of the CONTEMPT

code, originally formulated for light water reactor containment conditions.

To compensate for the fact that equipment closer to the leak will experience higher temperatures, the analysis expressed heat transfer surface area as a function of distance. At the rupture, the surface area was zero, and increased with distance from the leak. With increasing surface area, more heat is transferred to the heat sinks and less heat is used to increase surrounding air te'1perature. Thermal input to the building was based on the worst case accident. In the Reactor Building, it was determined that a cold reheat line break was the worst case. This was due to the fact that the Steam Pipe Rupture Detection System would automatically isolate a leak on a feedwater, main steam, or hot reheat line. Cold reheat would continue to be fed from the bypass flash tank until the operator acted to close the main steam bypass valves. (See Fig. 3) Based on discussions with plant personnel, it was determined that it would take two minutes to identify and isolate the leaking line, To be conservative, a four minute time limit was established. In the Turbine Building, it was decided that the worst case was the hot reheat line. This was based on the size of the line and the enthalpy of the steam. A four minute time limit was used for this analysis as well. Based on the initial supposition that all safe shutdown equipment was at least twenty feet from the nearest opposite loop steam line, testing began. It soon became apparent that several items were nearer to an opposite loop steam line than twenty feet. GGA then reran the computer program and generated a set of new curves. The results of this analysis were presented in report GA-A14212. Following the conclusion of GGA's test program, all documentation was transmitted to PSC. With the issuance-of IE Bulletin 79-01B, and the environmental qualification rule 10CFR50.49, PSC committed to an audit of the documentation. The audit found several items which did not fully meet environmental qualification requirements. These items were retested. The last test was completed on March 31, 1984 2-1

1 4 SECTION 3 CASE STUDIES The following case studies. analyze the consequences of steam leaks in both the Reactor and Turbine Buildings. All leaks are assumed to be a complete offset shear of the associated steam line, with the plant at-100% power, resulting in the greatest thermal input to the builoing. Included with each analysis are tables which list the sequence of events following the onset of the break. Corrasponding operator acticns are based on existing emergency procedures. Although many of the actions are performed simultaneously, all are considered to be sequential. The time given for each action was estimated following discussions with FSV Training personnel. All times are conservative. (Note: For all tables, TA refers to the time each action takes, TE refers to the total elapsed time from the onset of the break). CASE 1: Steam Pipe Rupture in the Reactor Building ) Any steam leak in the Reactor Building is readily detected by the Steam Pipe Rupture Detection System (SPRDS). Redundant pressure, temperature, and ultrasonic detectors identify the leaking loop, allowing the Plant Protective System (PPS) to shutdown the affected loop. The valves that are closed on a loop shutdown are shown on Fig. 3. If the leak is on a feedwater, main steam, or hot reheat line,-or on a cold reheat line downstream of. the circulator block valves, loop shutdown will isolate the steam source. As shown in Table 1, this occurs in approximately 10 sec. l TA TE' ACTION 5 sec 5 see PPS Alarm identifyina leakino 1o00, 5 see 10 see Leakina 1000 isolated by PPS. TABLE.1: Sequence of Events Following a Feedwater, Main Steam, Hot Reheat or Cold Reheat-line Break Downstream of _ the -Circulator Block Valves in the Reactor Building. CASE 2: Cold Reheat Line Break in the Reactor Building 4 As can be seen in Fig._3, a cold. reheat line break upstream of Lthe circulator block valves cannot be _ : automatically _ terminated 3-1

s - by thc PPS. After the PPS closes the main steam block valve, main steam pressure rises until the main steam bypass valve is automatically opened. Main steam is then bypassed to the flash tank and on to cold reheat. Thus, a cold reheat line break will continue to be fed until the appropriate operator action takes place. These actions are detailed in Emergency Procedure (EP) B-1 and are outlined in Table 2 below. TA TE ACTION 5 sec I 5 see PpS Alarm Identifyina leakinc loco. 5 sec 10 sec Turbine Runback to 50% previous power level. Verified by coerator. 12 sec 12 see Reactor scrammed by PPS on high reactor buildina temoerature. 5 see 17 sec Ocarator inserts manual scram. 5 see 22 sec Ooerator olaces 155' in low cower cosition. l 5 sec 27 sec Ooerator ensures transfer of house cower. 5 see 32 see i Ooerator verifies reduction of turbine oower. ~ 10 sec 42 sec I Ooerator verifies stable core cooline. 30 sec 72 sec Operator reads emergency procedures. Determines course of action. 20 sec 92 sec Operator closes main steam bypass valves. Leax teminated. TABLE 2: Sequence of Events Following a Cold Reheat Line Break in the Reactor Building. It is important to note that the continual thermal input into the Reactor Building from a cold reheat line break results in a reactor scram on high Reactor Building temperature. (The time shown for this scram is based on Reactor Building temperature switch setpoint of 175 degrees + 10 degrees F and assuming they are on the opposite side of the Teactor Building from the break. See Figure 4) Based on this analysis, a cold reheat line break in the Reactor Building will be isolated in less than two minutes. CASE 3: Main Steam Line Break in the Turoine Building I A main steam line break, eithe" in a loop header or in the common _ header is immediately cetected by redundant pressure switches located on the common he:tder. (See Figure 5) Two out of three logic scrams the reactor on low main steam pressure, closes both main steam block valves, and alarms the control room. A break downstream of the block valves is automatically suspended by the PPS in about 10 sec. (See Table 3) 3-2 1

i TA TE ACTION l 5 sec 5 see Reactor Scrammed on Main Steam Pressure low. 5 sec 10 sec PPS closes block valves on both loops. Leak isolated. TABLE 3: Sequence of Events Following a Main Steam Line Break in the Turbine Building Downstream of the Block Valves. A break upstream of the block valves is not isolated by the PPS. The appropriate operator actions following a reactor scram on low main steam pressure are listed in EP B-1. These actions are summarized in Table 4. The leak can be terminated in less than 90 sec. r TA TE ACTION 5 sec 5 see Reactor Scrammed on Mrin Steam Pressure Low. 5 sec' 10 sec PPS Closen Block Valven on Both Locos. 5 see 15 sec Ooerator Enserts Manua' scram. 5 sec 20 sec Ooerator Places ISS in Low Power Position. 5 sec 25 sec Ooerator Insures Transfer of House Power. 5 sec 30 sec Operator Verifies Reduction of Turbine Power. 10 see L0 sec Ooerator Verifies Stable Core Coolino. ) 30 sec 70 sec Operator Reads Emergency Procedures. Determines Course of Action. 5 sec 75 sec Ooerator Starts Auxiliary Boiler. 5 sec 80 sec Operator Closes Feedwater Block Valves. Leak Terminated. TABLE 4: Sequence of Events Following a Low Main Steam Pressure Scram. 1 CASE 4: Feedwater or Condensate Line Break in the Turbine Building j The consequences of a feedwater or condensate line break in the-Turbine Building-are much less severe than those anticipated for i other line breaks. Calculations show that for the worst case I feedwater or condensate break, the maximum steam temperature is 203 degrees F. This temperature is completely enveloped by the { Turbine Building'20 foot curve. (See Fig. 6). i 1 Thus, the operator has an extended period to identify and isolate any break in the feedwater or condensate-system. i Any break in the feedwater or condensate system results in a low main steam pressure scram.- Emergency Procedure B-1 identifies a feed.ater or condensate line break as a possible cause of the i scram and lists appropriate actions. a 3-3 )

r If the -break is downstream of the feedwater block valve, it is terminated in the same manner as the main steam line break summarized in Table 4. A break upstream of the block valve is also detected by redundant pressure switches located on the feedwater common header. Two out of three logic closes the normal boiler feed pump (BFP) discharge valve and opens the bypass valve to emergency feedwater. Since emergency feedwater will also leak out the pipe rupture, lack of back pressure results in high emergency feedwater fl ow. Redundant flow switches then alarm the control room. The operator then isolates the break by closing the emergency feedwater block valves. This lea?. is terminated in less than 90 sec. (See Table 5) TA TE ACTION 5 sec 5 see PPS Senses low Feedwater Pressure. 5 sec 5 sec Reactor Scrammed on Main Steam Pressure Law. 5 sec 10 see BFP Oischarge Valve Closed. Byoass Ooened. 5 sec 15 sec Hiah Emeraency Feedwater Flow A' am. 5 sec 20 sec Ooerator Inserts Manua' Scram. 5 sec 25 sec Ooerator Places ISS in Low Power Position. 5 sec 30 sec Operator Insures Transfer of House Power. 5 sec 35 sec Operator Verifies Reduction -of Turbine Power. 10 sec 45 sec Operator Verifies Stable Core Coolina. 30 sec 75 sec Operator Reads Emergency Procedures. Detemines Course of Action. 5-sec 80 sec Operator Starts Auxiliary Boiler. 5 sec 85 sec Operator Closes Emergency Feedwater l Block Valve. Leak Teminated. 1 TABLE 5: Sequence of Events Following a Feedwater Line Break I Downstream of the Block Valves. A break occurring between the feedwater check valves and the BFP normal discharge valve is isolated imediately by the redundant j pressure switches on the common feedwater header. (See Table 6) 1 i TA - TE ACTION r 5 sec 5 see PPS Senses low Feedwater Pressure. I 5 see 10 sec BFP Discharge Valve Closed. Bypass Opened. Leak Terminated. TABLE 6: Sequence of Events for a Feecwater Line Break Between the Feedwater Check Valves and BFP Discharge Valves. 3-4 l. \\

_. =- 4 i A break occurring between the condensate pump (CP) and BFP does not result in a harsh environment, as the water only reaches a temperature of 110 degrees F at the break. (See calculations). This break is also detected by a pressure switch that alarms the control room. CASE 5: Cold Reheat Line Break in the Turbine Building A break in either a loop cold reheat line or in the common header will result in a depletion of all steam supply to the hot reheat system causing a reactor scram on low hat reheat pressure. It also removes the source of helium circulator motive steam leading to circulator steam drive trips on one loop and complete circulator shutdown on the other loop. l The various operator actions reouired to terminate this leak are j described in EP B-1. Based on the actions listed in Table 7, the leak will be isolated in about 150 sec. TA TE ACTION 10 see 10 sec Reactor Scrammed on Hot Reheat Steam Pressure Low. 5 see 15 sec Ooerator Inserts Manual Scram. 15 sec 15 sec Low Circulator Speed Alarm (Based on Historical Data on Trios from Full Soeed). 5 sec 20 sec Circulator Trios Automatically. 5 sec 20 sec Ooerator Places ISS in Low Power Position. 5 sec 25 sec Ooerator Insures Transfer of House Power. 5 sec 30 sec Ooerator Verifies Reduction of Turbine Power. 10 sec 40 sec Ooerator Verifies Stable Core Coolina. 30 sec 70 sec Operator Reads Emergency Procedures. Determines Course of Action. 5 sec 75 sec Operator Starts Auxiliarv Boiler. 20 sec 95 sec Operator Trips both Circulators in One Loop. Loon Shutdnwn Insured. l' 15 see 110 sec Operator Trips Both Circulator Steam Drives In Other Loop. Closes Hot Reheat Steam Stoo-Check Valves in Loco. 5 see 115 sec Operator Insures Auto Water Turbine Start of tatt Two Circulators Trinned. 10 sec 125 sec Operator Opens Power Operated Main Steam Safetv Valves. 20 sec 145 sec Operator Raises Main Steam Bypass and Start Up Bypass Valves Setpoints to 2800 PSIG. Leak Terminated. TABLE 7: Sequence of Events Following a Cold Reheat Line Break'in the Turbine Building. 3-5

CASE 6: Hot Reheat Line Break in the Turbine Building Hot reheat steam enters the Turbine Building in two separate loops that combine into a common header downstream of the hot reheat stop-check valves (See Fig. 7 and Fig. 8). This common header then divides into two legs which admit hot reheat steam to ~the intermediate stage of the turbine. As in the case' of main steam, redundant pressure switches are used to monitor line breaks on the common header. Two cut of three logic scrams the reactor and alarms the control room. A line break in one of the loop headers will not result in a trip of the pressure switches because the operating loop, prevented from bleeding off through the leak by the stop-check valve, will continue to feed the turbine. However, due to reduced backpressure, the increased cold reheat flow causes excessive helium circulator speed. The result is the circulators trip on overspeed causing loop shutdown. The loop shutdown isolates the leak by closing the circulator block valves. This leak is terminated in about 10 sec. (See Table 8) TA ~ TE ACTION 5 sec' 5 sec Circulator Trios on Oversnood. 5 sec 10 sec PPS Initiates Loop Shutdown. Leak Terminated. TABLE 8: Sequence of Events Following a Hot Reheat Line Break Upstream of the Stop-Check Valves. A line break between the loop isolation valves and the pressure switches will be immediately detected by the switches. Based on EP B-1 and Table 9 below, the leak will be isolated in less than two minutes. 4 M 4 e 3-6

4 3 i TA TE ACTION 5 sec 5 sec .ieactor Scramme'd on Hot Reheat Steam Pressure Law. 5 see 10 sec Operator Inserts Manual Scram. 5 sec 15 sec Operator Places ISS in Low Power Position. 5 sec 20 sec Ocerator Insures Transfer of House Pcwer. 5 sec 25 sec Ooerator Verifies Reduction of Turbine Power. 10 sec 35 sec Operator Verifies Stable Core Coolina. 30 sec 65 sec Operator Reads Emergency Procedures. Determines Course of Action. 5 sec 70 sec Ooerator Starts Auxiliarv Boiler. 20 sec 90 sec Operator Trips Both Circulators in One Loco. Assure Loco shutdown. 15 sec 105 sec Operator Trips Both Circulator Steam Drives in Other tene. 5 sec 110 sec Operator Closes Both Hot Reheat Steam Stop-Check Valves. Leak Isolated. TABLE 9: _ Sequence of Events Following A Hot Reheat Line Break Between the Stop-Check Valves and Pressure Switches. A break downstream of the pressure switches results in a turbine trip on a loss of vacuum.- A turbine trip immediately initiates a programmed feedwater flow reduction to 25% of full load flow at a rate of.5% per second. Calculations show that with a hot reheat line break, the reactor will scram on low hot reheat pressure with feedwater flow at 30%. The sequence of events following this accident are shown in Table 10 below. TA TE ACTION 1 5 sec 5 see Turbine Trip On A Loss Of Vacuum. Feedwater Flow Procram Becins. 5 see 10 sec Ooerator Verifies Transfer of House Power. 5 see 15 sec Ooerator Verifies Reduction of Turbine Power. 140 see 145 sec Reactor Scrammed on Hot Reheat Steam Pressure Low. 5 see 150 see. Ooerator Inserts Manual Scram. 5 see 155 sec Operator Places ISS In Low Power Position. 10 see 165 sec Operator Verifies Stable Core Coolina. 30 see 195 sec Operator Reads Energency Procedures. Determines Course of Action. 5 sec 200 sec Operator Starts Auxiliary Boiler. 20 sec 220 sec Operator Trips Both Circulators In One Loco. Assures loco Shutdown. 15 sec 235 sec Operator Trips Both Circulator Steam Drives In Other loco. 5 sec 240 sec Operator Closes Soth Hot Reheat Stop Check Valves. Leak Terminated. TABLE 10: Sequence of Events Following a Hot Reheat Line Break Downstream'of the Pressure Switches. 3-7 l

i SECTION 4 CONCLUSION 'A feedwater, main steam, or hot reheat line break within the Reactor Building, is immediately detected by the Steam Pipe Rupture Detection System and isolated automatically in a matter of seconds. A cold reheat line break continues to be fed by the bypass flash tank until manual action isolates the leak after two minutes. The cold reheat line break is the most severe in the Reactor Building due to its extended. length. This affirms Gulf General Atomic's initial analysis. Since the cold reheat line break is isolated in two minutes, the four minute curve used for environmental qualification of equipment in the Reactor Building is conservative. The extra time that appears in the four. minute curve provides additional margin for, all test results. All equipment qualified for use 'in the Reactor Building remains qualified. A main steam line break in the Turbine Building is isolated either automatically.in 10 sec or manually in 80 sec. A feedwater line break is isolated in either 10 sec or 85 sec. Cold reheat is isolated in about 150 sec. Although a condensate line break may last longer than four minutes, the consequences are not severe since condensate will not flash to steam. A hot reheat line break can be terminated in as little as 10 see automatically or as long as four minutes manually. Thus, a hot reheat line break is the most severe in the Turbine Building. This confirms Gulf General Atomic's assumption. GGA's steam line rupture curves were generated assuming' a constant blowdown rate following the break and prior to leak termination. However, because of the reduction of feedwater flow following a 1 turbine trip, and subsequent reactor scram, the blowdown rate and thermal input into the Turbine Building will decrease as a function l of time. Thus, the steam line rupture curve for a hot reheat line break in the Turbine Building is conservative. The difference between a constant blowdown rate assumed by GGA and the decreasing blowdown rate occurring after a= turbine trip provides additional margin to that found on all~ test results. All equipment. qualified for use in the Turbine Building remains qualified. i O i 4-1

SECTION 5 REFRENCES GGA Report GA-A12045 " Qualification of Fort St. Vrain Safe Shutdown Equipment for Steam Environments Resulting from Pipe Ruptures." GGA Report GA-A14212 " Environmental Temperatures in the Vicinity of the Rupture Point of Steam Lines for Fort St. Vrain Equipment Qualification." Fort St. Vrain Final Safety Analysis Report FSV-50-22-1 " System

Description:

Secondary Coolant System" FSV-50-51 " System

Description:

Turbine-Generator and Auxiliaries." FSV-50-93-1 " System

Description:

Controls and Instrumentation." FSV-50-93-2 " System

Description:

Overall Plant Control and Plant Protective System." FSV-50-93-5 " System

Description:

" Steam Pipe Rupture Detection System:"

FSV-EP-B-1 "Emer ency Procedure: Reactor Scram (Without Two Loop Trouble ". FSV-EP-B-2 " Emergency Procedure: Two Loop Trouble Scram, With a Trouble Alarm in the Operating Loop." FSV-EP-C " Emergency Procedure: Loop Shutdown." s a l 5-1

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APPENDIX A CALCULATIONS The calculations' contained in this appendix were performed.to-analyze certain steam line ruptures in the turbine building. The calculations show that any main steam line break is detectable within 5 seconds by the low main steam pressure switch. The calculations also show that for a condensate or feedwater line break, the maximum release is a steam water mixture of approximately 200 degrees F. The calculations also show that following a turbine trip with automatic feedwater flow reduction the existing low pressure switches on the hot reheat line will trip within 140 seconds. Crane Technical Paper No. 410, Flow of Fluids through Valves, Fittings and Pipe, was used as a reference throughout. i 4 l A-1 i

PIPE RUPTURE IN THE TURBINE BUILDING ON THE CONDENSATE OR FEE 0 WATER LINES Problem: Find the temperature of the flashing steam escaping from the rupture. CONDENSATE Location - discharge of condensate pumps T=110 degrees F P=310 psia, h=78.78 Btu /lb Conditions flashes to atmosphere w/ constant enthalpy fluid: Therefore 9 P"12.3 psia, h=78.78 Btu /lb T=110 degrees F w/ no steam flash. FEEDWATER Location - discharge of boiler _ feedpumps T=316 degrees F, P=3300 psia, h=292 Btu /lb Conditions flashes to atmosr.1ere 9 P=12.3 psia, h=292 Btu /lb, Therefore: T=203 degrees F w/ 12% steam Location - disch'arge of feedwater heater #6 T=403 degrees F, P=3182 psia, h=382 Btu /lb Conditions Flashes to atmosphere 9 P=12.3 psia, h=382 Btu /lb, Therefore: T=203 degrees F w/ 22% steam. MAIN STEAM RUPTURE AT THE TURBINE The steam will reach sonic velocity at the exit for maximum choked flow. For compressible sonic flow at the break: 6 W = vda p

• W = flow (lb/hr) = 2.25 x 10 lb/hr

".THiOV y = velocity (ft/sec) d = inside pipe diameter = 10.4 in Substituting: p = density 6 2.25 x 10 lb/hr>= v x(10.4 in)2 xp .0509

• Blowdown rate for steam by GA Report GA12045-Solving: p = 1059 Eqn 1 v

Sonic velocity is reached at the exit and is equal to: p v= velocity (ft/sec)sonig v= kxg x 144 x P9 g= gravity = 32.2 ft/sec P = exit pressure 2 A-2

Substituting: p = density (lb/ft3) k = specific heat ratio = 1.3 for steam y= 1.3 x 32.2 ft/sec 2 x 144 in2/ft2 x P 9xv 1059 Solving: v = 5.7 x P 9 Eon. 2 Solving Eqn. 1 and Eqn 2 simultaneously yields: -outlet Temp = 1000 degrees F Outlet pressure P 2 = 400 psia same as normal Outlet velocity v = 2280 ft/sec flow for worse case Outlet density p =.46 lb/ft 3

== Conclusion:== The setpoint for the low main steam pressure scram is 1500 psig. The switches are located approximately 40 ft from the break where the exit pressure with choked flow is about 400 psia. Therefore, the pressure switches would trip in less than 5 seconds. l A-3 -v w ~ ,m

MAIN STEAM RUPTURE UPSTREAM OF THE LOOP ISOLATION VALVES CONDITIONS: P = 2437 psia T = 1002 degrees F h = 1460 Btu /lb y =.316 ft 3 /lb [$5 N 158?NE SITUATION: Line 320 y v lb/sec 20.75" Line a \\ RUPTyRE N ib/sec 0 25 Line A rupture in a single loop line will allow all of the steam in loop 2 to escape to the atmosphere. However, the check valve will not allow the steam from loop 1 to escape. Therefore, flow still exists at the pressure switch which is set 9 1500 psig. SOLUTION: Loop Lines: 0 = 16.25 in d = 10.82 in A = 91.9 i n 2 =.6385 ft2 4 Combined Line: 0 = 20.75 in d = 14.758 in A = 171.06 in e = 1.188 ft* Velocity of steam under normal conditions: 158 ft/sec in each loop V = 320 lb/sec x.316 ft3 /lb = .6385 f t 2 170 ft/sec in combined line V = 640 lb/sec x.316 ft 3 /lb = 1.188 ft 2 After one loop is lost, approximately the same pressure drop, thus the same velocity, will exist in the combined line. The velocity will increase some but this will result in a lower pressure, thus using 170 ft/sec is conservative. Therefore, after one loop is lost, the flow in loop one and the combined loop is equal to 320 lb/sec. Therefore: Loop 1 flow in lb/sec = Combined Line flow in lb/sec I l 158 f t/sec x.6385 f t 2 = 170 ft/sec x 1.188 f t 2 l .316 ft 3/10 v' ft3 /lb l A-4, ~

f 1 i At the pressure switches in the T mbined line: l '_ V =.633 ft3/lb and T = 1000 degrees F Therefore: P = 1293 psia @ PSL i CONCLUSION: The pressure at the switches reaches about 1295 psia instantly, therefore the switches would trip with a setpoint of 1500 psig. 4 i l 1 I 4 ( k l i I ) 4 A-5

HOT REHEAT STEAM RUPTURE NEAR THE TURBINE Find percent of flow to reach the set point of 35 psig. Resistance of pipe: K=f L/D - 30 ft. of 34" pipe 4 K=f L/DB - 40 ft of 22" pipe K =.5(1-82 )/sinf - sudden contraction 84 ~ K = 1/B 4 - exit where B =_,d1 20 in .65 = d2 31 in f =.012 for turbulent flow + = 180* 34" pipe K =.012 X 30 ft. X 12 .127 34 in. 22" pipe K =.012 X 40 ft. X,12, 1.47 22 in. X.65 contraction K =.5 (1 .652) X 1 , 1.62 .654 4 exit K = 1/.65 = 5.6 K TOT = 8.82 47.3 psia A+

PSL, P1 = 35 psig + 12.3

= Atmospheric P2 = 12.3 psia 35 =.74 for AP =.74 and K = 8.8 AP = f{ 47.3 T{ - Y =. 71 w/ subsonic flow A-6

Therefore: W = 1891 X Y X d2 X aP W = flow (lb/hr) Y =.71 KXY1 d = 31 in. AP = 35 psia K = 8.82 it=14ft3/lb@35psig h = 1360 Btu /lb W = 1891 X.71 X 31 in2X 35 psia 8.82 X 14 ft3/lb W = 686, 924 lb/hr 31% Percent normal flow = 686, 924 lb/hr =.31 = 2.25 X 10" lb/hr 4 After a turbine trip, the feedwater flow is automatically reduced to 25% flow by h% per second. Therefore it would take: += 70% = 140 seconds .5%/sec. t A-7

APPENDIX B This appendix contains the emergency procedures that are used by the reactor operators to respond to a steam line rupture. Depending on the type of steam leak, the operator will receive certain alarms or indications that direct him to the appropriate procedure where the correct imediate and followup actions are found. -8 1-

PUBLIC SERVICE COMPANY OF COLORADO g; 5-1 FORT ST. VRAIN NUCLEAR GEhERATING STATION 55Ue 31 4 age. o : / TITLE: REACTOR SCRAM (WITHOUT T40 LOOP TRCU3LE) ISSUANCE 0 ? _ 0 ( Q]J AUTHORIZED BY *'"' c. 21 - sz [TE "g,"b ponc 486 0Ci141982, y I i r o e somu 372 22 3e42

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PUBLIC SERVICE COMPANY OF COLORADO 73al 3 1 FORT ST. VRAIN NUCLEAR GENERATING STATION !ssue 51 Page 1 of 1 0 Table 3-1 PLANT CCNDITIONS REQUIRING MANUAL SCRAM 1) Abnormal reactivity changes beyond the caoability of tne regulating rod to compensate. (Emergency Procedure E) 2) Failure of a circulator helium outlet flapper valve to shut after circulator trip (Etergency Procedure C or D.) l 3) PCRV Relief valve opens (Emergency Procedure H-2) 4) Steam leakage that is not isolated by a loop shutdown (Emergency Procedure C.) 5) Loss of all hydraulic pressure in one hydraulic power system ( (Emergency Procedure M) 6) Turbine shutdown following a loss of outside power without a turtine trip. 7) Loss of outside power with turbine trip. (Emergency Procedure l F-3) l 8) Loss of condenser vacuum (Emergency Procedure F-2) 9) Fire in, or affecting, a control panel, instrument cabinet, or I load center. (Emergency Procedure I)

10) Fire in the general plant area affecting continued safe l

operation. (Emergency Procedure I)

11) Loss of access to the control room (Scram from I-49) (Emergency Procedure R)

12) Loss of instrument air Headar (Emegency Procedure L)

13) Permanent loss of instrument Bus 1 or 2 (Emergency Procedure N)

14) Loss of interruptible Bus 3.

(Emergency Procecure N)

15) wnen any automatic scram is called for.

L Pomw 372 32 3443

PUBLIC SERVICE COMPANY OF COLORADO gp App g_; FORT ST. VRAIN NUCLEAR GENERATING STATION Issue 31 Page 1 of la INTRODUCTICN Reactor scram is the ultimate defense against any circumstance or condition that threatens to damage the reactor core and release radioactive fission products. Reactor scram is initiated automatically by the Plant Protective System (PPS) in a number of situations described in the following discussion of symptoms. Reactor scram is also initiated manually in a numcer of situations in which the control of the plant is threatened, and a rapid shutdown is called for. The reactor scram results in a raaf d shutdown of the nuclear chain reaction and the attendant power production from fi s sion.

However, immediately after reactor scram there remains a significant amount of heat generation due to the radioactive decay of tne fission products.

Tnere is also a large amount of thermal energy storea in the primary components, particularly the graphite core anc reflector. Therefore, the control and reduction of primary system temperature requires a continuation of heat removal through the steam generators. The control system is designed to acco"plish these necessary heat removal activities automatically following the scram. The primary function of the immediate operator actions of this Emergency Procedure are to backup the Control System to insure adequate removal of decay heat f from the core, k The specific ?PS actions that are initiated by reactoa scrar* are as follows: 1. The controi rod brakes are de-energiced. 2. Power to #1 and .2 Red Drive Motor Control Centers are interrupted (K49-51 and K48-50 green lights on - Board I-10). 3. First-in scram annunciator and indicating light are actuated, indicating which logic channels have tripped, A, B, or C. 4. Turcine runback is initiated (Scram auxiliary relay rec lights on). 5. Turbine is tripped 120 seconds after runcack is initiated. (TR54-55 green lights on - Board I-10). NOTE: Turbine trip is initiated immeciately ucon receiving reactor pressure low (programmed) signal or signal indicating trip of all four helium circulators. i 8CRM 372 22 3843 .g g,

PUBLIC SERVICE COMPANY OF COLORADO gp ;;p g. FORT ST. VRAIN NUCLEAR GENERATING STATION Issue 51 a Page 2 of 13 6. The feecwater flow program is ini-iatec (scram auxfitary relay red lignts on). NOTE: The feedwater program is inhibited on reheat steam temperature high scram. 7. Circulator trip on programmed low feedwater flow (caused by scram signal) is prevented. DISCUSSION OF SYMPTCMS SYMPTCMS 1.1 Startuo Count Rate Hich (RMS Fuel Leadine) I-038 4/4. 5/4. 6/4 This scram is actuated when either source range nuclear cnannel I equals or exceeds a neutron countrate of 1.0E-5 counts /second with tne Reactor Mode Switch in the FUEL LOAOING position. This scram is provided for use during fuel loacing and preoperational testing or other low power operations. The scram action is initiated by 1 of 2 logic trip by the Nuclear Start-up Channels I or II. I 1.2 Neutron Flux Rate of Change Hich (ISS Start-Vol I-038 4/3. 5/3. \\ $!1 This scram is actuated as a result of Wide Rance Nuclear 1 Channels III, IV or V equaling or exceeding a neutron flux rise of 5 OpM. This scram is utilized during plant start-up and results in additional protection and better scram response than the " Neutron Flux-Hign" scram in case of accidental control rod withdrawal when operating with the l',5 in the start-uc position. The setpoint is selected to be above the usual operating rate of flux change. 1.3 Neut on Flux High I-033 1/1.1/2, 2/1, 2/2, 3/1, 3/2. l This scram is actuated as a result of the correct two of six nuclear channels (III, IV, V) or (VI, VII, VIII), equaling or exceeding 140'.' of full power flux. Channels III, IV, and V are l combination power and wide range nuclear detectors. Channels VI, VII, and VIII are power range nuclear detectors. High neutron flux levels with the associated excessive heat generation requires a scram to prevent damaging core temperature increases. now ass.ss sess -B 9-

PUBLIC SERVICE COMPANY OF COLORADO Ep App 5-1 + FORT ST. VRAIN NUCLEAR GENERATING STATION Issue 51 page 3 of 13 / 4 Renea: Steam Temaerature Hich I-035 1/3. 2/3. 3/3 When the hot reheat steam temperature, as measured by the hign selected average of two thermoccuples from eacn loop par scram channel, ecuals or exceeds the 1075"F setcoint, ths. channel trips. Tripping of two of the three channels causes scram action. Hot reheat steam temperature high indicates an increase in core power generation or a decrease in primary coolant flow or reheat steam flow. Hot reheat temperature is used in lieu of reactor primary coolant outlet temperature because of the difficulty in providing redundant gross primary coolant temperature measurements. The design of the steam generator is sucn that enanges in primary coolant temperature first affects the hot eeneat steam temperature. Tnis scram serves as a cackup to the neutron flux high scram. The trip level is chosen to be just above expected transients in order to minimi:e the scram response time and corresponding temoerature overshoots on rod withdrawal accidents not terminated by the neutron flux high scram. To further reduce the temperature transient, the feedwater flow f program is initially defeated by clamping the cutout of the rate (' measured reheat temperature is reduced to 975"F, the feedwater limiters in the feedwater flow setpoint circuit. When the flow program will occur. Maintaining the feedwater flow fixed for a reheat steam temperature high scram from 100% load will require about 23% of the steam to be bypassed by the electromatic relief valves. This condition will last for Jhout nine minutes to reduce the reheat steam temperature to 9758F. 1.5 Reactor Pressure Hich (Procrammed) I-038 1/4 2/4 3/4 The reactor pressure high protective action is a backup to the High Moisture protective action. The reactor pressure high scram action is to scram the reactor and shut down and dump tne loop that is preselected by a hand switen. Preselection is necessary because the reactor pressure high action has no basis for identifying the leaking loop. It cepends solely on the physical fact that suf ficient moisture inleakage will raise the reactor pressure above the high pressure setpoint (7 1/2% above normal) wnich is programmed with circulator inlet helium temperature. The reactor pressure hign action also automatically depressurizes the operating steam generator loop to reduce the pressure differential across tne lean in case tne wrong 1000 has been dumped or both loops are leaking. 1.6 Two Looo Trouble I-039 4/1. 5/1. 6/1 Each of the two primary-secondary coolant loocs is providec with a series of PPS actions tnat shutcown the loop for a variety of ~ reasons. These loop shutdown actions are coverec :y Emergency i Pomu aft-23 3643 -B 10-

PUBLIC SERVICE COMPANY OF COLORADO p app 5-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue 51 Page 4 of 13 ? Procedure "C" and explained in Appendix C. Wita one loop shutdown, the second loop must remain in operation at all times to provide active cooling of the reactor core. Thus, with one loop shutdown, the automatic shutdown of tne second loop is inhibited and a reactor scram called "Two Loop Trouble" is substituted when shutdown of the second loop is called for by the PPS. Two Loop Trouble scram is inhibited with the RMS switch in fueling loading. 1.7 Loss of plant power I-038 4/2. 5/2. 6/2 Undervoltage detectors sense the voltage on all three phases of I both essential buses 1A and IC. Detection of voltage loss l persisting for 30 seconds by 2 of the 3 detectors on each bus l produces 2 of 3 scram channel trips, causing scram. The accident of concern is the loss of outside power, coincident with turbine generator trip and failure of one diesel generator to start. A scram is required to reduce heat generation to allow for heat removal with less than a normal complement of plant equipment. The immediate actions specified are designated to allow for orcerly start-up and/or loading of the diesel generators in the event that they both haven't automatically / supplied power to the essential loads. The follow-up actions ( are details for the manual starting and loading, as well as procedures for continued shutdown cooling. 1.8 Reactor Buildino Temoerature Hich I-038 4/6. 5/6, 6/6 High reactor building temperature (greater than or equal to 1758F.) indicates steam or helium leakage into the building. If PPS action to shutdown a loop did not isolate tne leak, immediate actions are required to determine the source of the leak and to take actions to isolate it. The temperature and pressure recorders on I-09, actuated by a high pressure or temperature at any of the 14 sensor locations, should help in determining which loop is leaking. The opening of the main steam power operated relief valves (22% capacity) with 25% loop flow would not result in a pressure decay unless a leak of significant si:e (greater than or equal to 3%) was present in the loop main steam piping. If no pressure decay takes place, and a significant si:e leak has occurred, the isolation of all reheat piping in the affected loop, or the isolation of all common cold reheat piping should. isolate the leak. 1.9 Reactor pressure Low (proccammed) (ISS - Power) I-0331/5. 2/5, 3/5 Pressure elements sense the pressure in the circulator cisenarge plenum and operate pressure switenes in the PPS. Actuation of two of the three switenes trips their respective scram cnannels and causes scram, k ponha 372 22 3643 -8 11-

PUBLIC SERVICE COMPANY OF COLORADO gp App 5 ; FORT ST. VRAIN NUCLEAR GENERATING STATION Issue il a page 5 of 18 a Low primary coolant pressure is an indication of gross primary coolant leakage from the system. A scram is initiated because the reactor is in danger of being inacequately cooled, which would increase the ha:ard associated with activity release from the PCRV. The low oressure trip setpoint is 50 psi below normal and is programmed with helium circulater inlet temperature to reduce the resoonse time. The reduction of response time is desirable to minimi:e steam generator tube stresses caused by an increasing primary coolant temperature in comoination with a constant (controlled) main steam temperature. The control system tends to aggravate the situation until a scram overrides the control system. A turbine generator trip is initiated simultaneously with the scram to anticipate the ensuing crop in main steam temperature. The turcine should trip immediately in tnis scram insteac of running back and tripping after 120 seconds as in most scrams. Immediate trip is required bacause the low pressure helium is not capable of transporting enough heat from the core to the steam generators to prevent steam generator flooding and water washing of the turoine if the normal runback-trip sequence is followed. The operator should manually trip the turbine as quickly as possible if automatic PPS actions does not. ( This scram protects the reactor in the inconceivable event of the complete failure of both closures of a PCRV penetration. Any concet table leak of primary coolant results in the reduction of the primary coolant system pressure' to atmospheric over a period of from several hours to several days. The source of any leak should be determined and the leak isolated in this length of time. If the scurce of leakage cannot be located or isolated, the reactor should be manually scrammed and the PCRV depressuri:ed (por 50P 24). ~ 1.10 Main Steam pressure Low (ISS-oewer) I-023 1/6, 2/6, 3/6. If the main steam pressure drops below 1500 psig, each of three pressure switches located on the common main steam line in the y turbine building will cause a scram logic channel trip. Two of the three channel trips will cause a scram. Low main steam pressure is an indication of either main steam line rupture or of gross failure of the feedwater system. Immediate shatdown of the reactor is appropriate in eitner situation. In addition, both superheater outlet stop check valves are automatically closed to reroute main steam to tne flash tank through the individual loop cypass valves and desuperheaters. This is necessary for the centinued coeration of the helium circulators on steam. The scra5' trip point is selected to be below ' normal.coerating levels and system transients. The main turoine lead is run back by tne ini,ial pressure regulator as pressure drops below :ne 2t.00 psig normal 6 non sn.ss sus 4 gg,

PUBLIC SERVICE COMPANY OF COLORADO 5p ;; 3_; FORT ST. VRAIN NUCLEAR GENERATING STATION biue 51 4 paga 6 of 15 operating cressure. The turoine is snutcown oy tne turoine protective system at - 2200 psig. If the cause of the low main steam pressure was rupture of the main steam line, the automatic closing of the suoerneater stoo check valves would isolate the leak and cooling would continue with both loops on steam drives of the helium circula Srs. Continuation of core cooling would be accomplisned by one of the shutdown cooling modes. The loss of feecwater considered is the complete less of tne use of all three boiler feec pumps due to failure of either the main condensate line on deaerator side of LCV-3175, tne ceaerator, the boiler feed pump suction line, the cumos themselves, or both main and emergency feedwater lines. If the cause of the low main steam pressure was failure of the feedwater system, all four helium circulators are automatically shutdown due to loss of feedwater fl.w and one loop will be shutdown cue to both circulators being tripped. In this case, further actions covered by abnormal operating procedures will be required to establish core cooling. 1.11 Reheat Steam pressure Low (ISS - power) I-038 1/7, 2/7, 3/7. ( Oetection of low pressure by pressure switches located on the . common hot reheat steam line in the turbine building causes a scram logic channel trip. Tripping two of three channels causes a scram. The low pressure trip setpoint is 35 psig. Hot reheat steam pressure low is an indication of a ructure in either a cold reheat line or a het reheat line in the section common to both locos. Loss of the cold reheat steam line results in loss of the' steam supply to all circulators and boiler feed pump turbines. The direct scram in this case precedes a scram resulting from Two Loop Troucle. The trip point is selected to be below normal operating and transient pressures which vary over a wide range. The Pelton Wheel drives on the helium circulators are automatically started by Ahutting down all circulator steam l drives. The AUTO WATER NRBINE START will come in for the last [ operating loop on steam drives. The hot reheat stop valve is closed automatically in the second loop to aid in leak isolation. 1.12 Manual I-038 4/5. 5/5. 6/5 A manual scram is. inserted when called for in Tacle 3-1 and following an automatic scram as a backuo to the DDS to insure a l full scram. A manual scram can be inserted by coerating tne l scram handswitch to tne scram cosition, by cooressing two of I three pusn tuttons on I-49, or y placing the RMS (Reactor voce [ $ witch),.r.o the off position. Also there are several mechanical i rom m.n nu -B 13-

PUBLIC SERVICE COMPANY OF COLORADO Ep ;;p 3-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue 31 Dage 7 of 13 ways in wnich brake power may be cisruptec.nere oy allowing rods to fall (pull brake fuses, cut power cables, etc.). DISCUSSICN OF IMMEDIATE ACTION 2.1 Insert Manual Scram. Manual scram is inserted when called for (See Table B-1), and, following automatic scram as a backup to the PDS to insure a full scram. Inward red motion and decreasing flux are coserved to verify that the scram is having the desired effect. 2.2 Place ISS in Low powe-Position The ISS may be placed in the low power position after plant conditions have stabili:ed and reactor power is < 3 0*.". Stable plant conditions are insured when:

1) Main turoine generator has tripped; 2) Boiler feed pumps and feedwater flow has stabili:ed;

3) Helium circulator sceeds are normal; 4) Main and reheat steam temperatures are decreasing at a maximum rate of 2*F/ minute.

When the ISS is placed in the low power position the maximun / cooldown rate of 2* F/mi nute must be controlled manually by ( adjusting feedwater flow and circulator speed. Also, during plant cooldown the operator must be aware of the relationship between feedwater flow and circulator speed. If l feedwater flow is lost, circulator speed must immediately be reduced to :ero to prevent damage to steam generators and/or l helium circulators. When restoring core cooling, feedwater flow should always be established before primary coolant flow is established. l 2.3 Ensure Transfer of House Power Following all scrams at power (except the reactor pressure low scram and a scram resulting from all four circulators tripped) an immediate acti.on of the operator is to manually transfer the 4160V buses to the reserve auxiliary transformer, within the 120 seconds before the autcmatic turoine trip takes place for tne following reasons. Any of the following PPS actions cause the main generator auxiliary tripping relay 86G1 to be energi:ed: 1) Any scram, with a 120 sec. time delay. 2) Reactor pressure low scram (programmec), no time celay. 3) All four circulators tricoed, no time celay, comu at

22. u43

.g 14

PUBLIC SERVICE COMPANY OF COLORADO Ep App g-1 FORT ST. VRAIN NUCLEAR GENERATING STATION I$su* 51 Page 3 of 15 l / Energi:ing auxiliary ripping relay 56G1 procuces tne fc11cwing actions: 1) Opens CCS 5301 (230 KV generator breaker) 2) Opens GCS 5300(230 KV generator breaker) 3) Energi:es 41G1, which trips the field breaker. 4) Energi:es 86GT1, generator and transformer auxiliary tripping relay which in turn: a. Opens ACB 152ATIA (4160 volt feed to 1A bus) b. Opens ACS 152ATIC (4160 volt feed to 1C bus) c. Starts all emergency diesel engines. 5) Energi:es the electro-hydraulic master relay, whfen in turn: a. Oe energi:es two pilot solenoids of the master trip solenoid valve (24VOC) (Trip requires 2 of 2). t ( b. Energi:es mechanical trip solenoid (125V OC). c. Either of these trip signals will trip the turbine SV's and ISV's. 6) Produces light and alarm. The following acti6ns also take place, as appropriate: 1) Low voltage on IB bus causes 152RT1B to close. 2) If 152ATIA is open and 152RT1B is closed, bus tie STAS closes. 3) If 152ATIC is open and 152RT1B is closed, bus sie 373C closes. As outlined above, it is seen that a momentary loss of power to the 4160 VAC buses, and hence to the 480 volt buses, is required for the automatic switching to the reserve auxiliary transformer to occur. However, all motor control center breakers are coerated and held closed by holding coils energt:ed from the high side of each breaker. Thus, a momentary interruption of power to the motor control center buses will cause the associated closed breakers to open. Ecutoment must snen ce restarted tnrough tne auto start provisions of the various systems or througn opera *cr action. ponu 373 22 J443 -B 15-

PUBLIC SERVICE COMPANY OF COLORADO EP App 3-1 Issue C FORT ST. VRAIN NUCLEAR GENERATING STATION + page 9 of 13 2.4 Ensure Turcine Trio This is a backup to the PPS action required only when the turaire is in operation. The turoine is tripped in anticipation of the drop in steam pressure resulting from the scram. Witnaut the turoine trio, control of steam pressure and temperature would be difficult and the coolcown rate would ce excessive. Furtner, the time period after scram in wnich the circulators and feed pumos could continue to be operated on reactor steam would be rapidly decreased by continued turbine operation. Most of the scrams result in turbine runback at 1% per second to 10*. load followed 120 seconds after scram by a turbine trip. This provides a less severe turoine shutdown transient, insures a high rate of heat removal immediately following the scram to absoro energy that may have been produced by events leading to scram, and gives the operator the option to transfer tne in-house power supply to the reserve auxiliary transformer. Two scrams, reactor pressure low and all four circulators tripped (i.e., two loop trouble), result in immediate turoine trip without runback. In both these cases, continued turoine operation only aggrevates the situation, cc immediate trip is 7 appropriate. ( If the PPS does not runback and/or trip the turbine as recuired, the turbine should be tripped manually. l 2.5 Ensure or Establish Stable Core Cooling Conditions Verify that the transients associated with the scram have not caused circulator trips, and if they have, verify that the water turbine drives have started automatically. The circulators should continue to operate on reactor steam for some time after scram. If the circulators are not operating, then the operator must take immediate action to restart at least one circulator on water or steam drive, as appropriate. Procedures for supplying circulators with water or steam under a variety of circumstances are provided in the 50P's, in Abnormal Operating Procedures and in the " Safe Shutdown Cooling with Highly Degraded Conditions," document. Because of the initially high decay heat load and the stored energy in the core, steam conditions snould remain near the normal operating setpoints for some time following the scram. The operator should ve ri fy snat sne main and reheat steam temperatures anc pressure controllers have handled tne scram transient and continue to maintain proper se point concitions. If not, the operator should take manual control of tne eraant controller. I f amM 373 22 3443 -B 16-

PUBLIC SERVICE COMPANY OF COLORADO g,: Apo 3-1 FORT ST. VRAIN NUCLEAR GENERATING STAT!CN Issue 31 Page 10 of 13 The rapid power cecrease of a scram, turoine runcack anc trip requires pre-programmed feedwater control system action to mitigate the thermal shock to tne steam generators. The ooerator should verify that the feecwater flow program is initiated (inhibited on Hot Reheat steam temoerature high scram until Hot Reheat temperature is below 975'F.). If not, the operator should take manual control of the errant controller. ( ( l r, l i 80mW 373 33 3443- -B 17-

PUBLIC SERVICE COMPANY OF COLORADO g; App g-; FORT ST. VRAIN NUCLEAR GENERATING STATION Issue 51 Page 11 of IS B OISCUSSICN OF FCLLOWUP ACTIONS l 3.1 Ensure Reactor Internal Maintenance Te rminated and Coenines Closec This scram occurs only when the Reactor Mode Switch is in :ne " FUEL LOADING" position. The coerator should acvise any personnel involved in operations from the refueling floor that a scram has occurred and that further operations snould be terminated until it is determined to be safe to proceed. 3.2 Start Auxiliary Boiler As reactor steam pressure drops off following ne scram, an auxiliary boiler should be started to supply the 150 psig steam heacer. The time after scram that the auxiliary boiler will be required varies with a numcer of factors including the power conditions at which the scram occurred, the decay heat rate (influenced by core power history), the rate of cooldown and equipment operability conditions in the plant (for example, auxiliary boiler steam may be desired as a backuo pcwer source .for the circulators if the feedwater system has problems). / 3.3 Establish Two Circulators at - 7000 rpm ( Two circulators (on steam driven or on Pelton' Wheel) at 7000 RPM will provide sufficient primary coolant flow to prevent core over heating from decay heat. The two circulators may be one in each loop or two in one loop and still provide sufficient primary coolant flow. If any core recion outlet temoerature exceeds 2200*F, increase circulator soeec to 8000 rem to octain maximum availaole crimary coolant flow. 3.4 Isolate Primary Coolant Leak. Deoressuri:e Throuch Purification Train if Possible Pumping down the PCRV through the Helium Purification System will minimice the leakage of radioactive helium into the reactor building, release to the atmosphere and recuce subsequent cleanup problems. 3.5 Isolate Secondary coolant Leak or Establish Feecsater Flow i In conjunction with symptom 1 (1.8 Reactor Building Temperature Higr.' Scram.) 1) If temoerature and pressure records on I-09 clearly indicate the leaking loop, trip both circulators 'n

nat loco.

pomu 372 22 3443 -8 18-

PUBLIC SERVICE COMPANY OF COLORADO Ep A p s-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue il + Page 12 of 13 Tripping both circulators results in loop snutcown wnica shuts the feedwater, main steam and reheat steam stoo valves as well as the circulater steam inlet and outlet block valves. This action isolates the steam leak unless it is in the cold reheat line or the hot reheat piping downstream of the reheat stop-check valve. 2) If I-09 recorcers do not indicate the leaking loop: a) When feedwater flow has been reduced to 25%, open the power operated main steam safety valves in both

loops, b)

If main steam pressure decreases in either loop, trip both circulators in the loop. If the leak is in the 5-10% of steam flow range, it may be difficult to locate by the I-09 instruments and the rate is too small to make a significant difference in loop steam pressures. In this case opening the power operated main steam relief valves will dump about 22% of the 25% steam flow so that a smaller leak should make a significant difference in the steam pressure in the leaking loop. k 3) If loop steam pressure decrease coes not indicate the leaking loop, the leak must be in the reheat steam piping. In this event: a) Trip all four circulators and shut both circulator bypass block valves. b) Start water turbine drives in one loop. c) If leak is still not isolated it must be in the cold reheat piping. In this event, raise setpoint on the main steam bypass valves until all steam flow is through the Main Steam Relief Valves. These steps assume that the leak is in the reheat pioing and cannot be isolated by shutting down a loop. Therefore, steam flow to all circulators is secured and water turbines are started to maintain active core cooling. If the leak turns out to be in the ecmmon cold reheat piping between the turbine and the circulators, it is necessary to raise the main steam bypass valve setpoints until they are shut and all steam flow is through the relief valves. This action is necessary because the bypass valves acmit steam to the flash tank which cannot be isolated from the cold reheat piping. pomu 372 22. aso

PUBLIC - SERVICE COMPANY OF COLORADO E; App 3-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue 31 + Dage 13 of 13 e 3.5 Isolate Seconcary Coolant Leak or Estaoitsn reecwa:er Flow In conjunction with symptom (1.10 Main Steam Pressure Low Scram) 1) Main Steam Pressure Low Caused by Line Ruoture. I a) Ensare rupture is isolated. Low main steam pressure would be excectec to result from a rupture in the main. steam line, the loop steam lines, or main or emergency feedwater lines. The rupture should be isolated by automatic closing of the loop isolation valves on loop shutdown, closing of the loop main. steam stop checks or closing of the boiler feed pump discharge valves when low pressure is sensed in the main feecwater line. Oserator action would be required to isolate an emergency feedwater header rupture. b) Maintain shutdown cooling with available helium circulators and steam generators. ( With the line rupture isolated, shutdown cooling may be maintained with steam or water driven helium circulators and normal or emergency feedwater supply to the operable steam generator depending on the plant conditions. 2) Main Steam Pressure Low Caused by Loss of Main and Emergency Feed Water Supply. a) Depressurize at least one steam generator loop and establish condensate supply of steam generators via the emergency condensate line. The loss of main and emergency feedwater supoly could result from failure of the main condensata line, the deareator, the boiler feecouco suction line, the pumps themselves or both the main anc emergency feedwater line. The loss of feedwater supply will cause loop shutdown of one loop and trip of the remaining two circulators on low speed. Both loco mainsteam stop-check valves are closed. The decressuri:ation of the steam generators is required prior to estaclishing concensate flow anc is accomolished with the electromatic relief valves. A concensate flow path is estaclished with the emergency concensate line suoply and disenarge tnrougn the main steam generator bycass valves to the flash tanks. k somu 372 22 34&3 -8 20-

PUBLIC SERVICE COMPANY OF COLORADO Ep r,pp 5-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue il Pace 14 of 13 / b) Operate at least one nelium circulater in an operable loop on concensate to match helium flow with condensate supply to the steam genera:ce. After condensate sucoly to one steam operator nas been established one or two helium circulators, as necessary, are started on condensate drive anc circulator speed increased until the desired nelium flow is obtained. The helium flow should be adjusted to maintain the steam generator in tne sub-boiling conditions. 3.5 Isolate Secondary Coolant Leak or Establish Feedwater Flow In conjunction with symptom (1.11 Reheat steam pressure low scram) 1) If circulators are still operating on steam: a) Trip both circulator steam drives in one loop and insure appropriate loop shutdown. b) Trip both circulator steam drives and shut het ( reheat stop-check valves in 'the other loop. c) Insure automatic water turbine start of last two circulators tripped or manually start circulators on water turbines. This scram indicates a major steam leak in the reheat steam piping'. Such a leak will probably affect steam flow to all circulators, therefore, all circulators are shut down manually if they haven't already tripoed automatically. In this situation, the water turbines in one loop should start automatically to maintain active core cooling. 2) Open power operated main steam safety valves 3) Raise main steam and startup bypass valve setpoint to 2300 psig. These steps provide a path for steam flow so that core cooling is maintained and then shut off steam flow to the flash tank. This is necessary in the event of a steam leak in the reheat piping because the flash tank cannot be isolated from the cold reneat line. Continue cooldown with this flow cath until temperatures are low enough to switen to condensate. The secondary coolant temperature and pressure will be low enougn that a flow path may se re-established through the bycass flash tank. k form 372 22 3443 ,g

PUBLIC SERVICE COMPANY OF COLORADO

g. App 3-1 a

FORT ST. VRAIN NUCLEAR GENERATING STATION !ssue 51 page 15 of 18 l 4) Ensure motor driven ootier feed pume 1,= operating anc :nen shut down steam driven boiler feed pumps. As discussed above, all reactor steam is being dummed to atmosphere through the safety valve. Therefore, steam flow to the deaerator and turbine driven feed pumas is lost, at least until the auxiliary boiler can be brought into service. l 5) Ensure hetwell makeup valves are open. As discussed above, no steam is being returned to the condenser, so hotwell level must be maintained by makeuo from the condensate storage tank. 6) Reduce feedwater flow to maintain main steam cressure below the setting of tne main steam safety valves. This step assumes that the scram occurs with feecwater ~ flow greater than 25*.. In this event, the feedwater flow must be reduced to the capacity of the open power operated relief valve (about 22*.). Thereafter, the feedwater flow should be controlled to maintain and/or reduce the steam temperature and pressure depending on whether the plant is ( to be cooled down. The steam pressure should not be permitted to increase above the setpoint of the spring loaded safety valves. 3.6 If Reactor Pressure Hioh Scram or Two Loce Trouble Scram Due to Moisture Inleakace. Proceed to Emercency Procecure "A" Emergency procedure "A" contains the appropriate operator action in the event of moisture inleakage. 3.7 Start Emergency Diesel Sets and Enerei:e 480 Volt Essential Busses In conjunction with symptom (1.7 Loss of plant power scram.) I 1) Ensure all standby diesel generators have started. If not, start them from the Control Room. If none will start from the Control Room, start them from their control panel. In this situation the primary concern is maintaining enough electrical power to continue active core cooling. Only one diesel generator is required to produce this minimum amount of power. Thus if the turoine generator and all outside power are lost, at least one ciesel-generator must be started. The second diesel generator provices backup to tne first and produces accitional cower 4 nom an.11.aus _g gg_

PUBLIC SERVICE COMPANY OF COLORADO Ep App 3-1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue P. page 16 of 13 to facilitate sucsequent operations. Ectn ciesel generators should start automatically when the turoine generator trips and should pick up essential loads when power is los to the 4KV busses. In the event of malfunction in the automatic control s,

ne diesel generators can be started and/or loaded manually f-em the Control Room.

If manual control from the Control Room is lost, the diesel generators can be started from their local control panels. Whatever the operator action

required, it is essential that at least one diesel generator operates when all other power sources are lost.

The following are suggested troubleshooting steps to be followed when the diesel generator runs, but will not pick up the electrical load as it should. The auto-start prohibit rel,ays prevent the breaker between the diesel generator and the 480 volt bus from closing until the 480 volt bus is de-energi:ed. Because the 480 volt bus normally carries more than the essential leads, load shedding must occur and the bus must be isolated before the supply breaker can be closed without overloading the diesel-generator. When these conditions are satisfied, the diesel generator feed breakers must close and the 480 volt bus tie between the energi:ed bus (1A or IC) and bus i IB must close to supply the essential loads. When power is available to the 480 volt bus, the automatic load sequencer should operate to start essential equipment in the proper order. If any of these normally automatic actions fails to occur, the operator should take the appropriate manual action. 2) Ensure that the diesel generator auto-start prohibit relays (HS-9221/5 and HS-9225/5) are tripped. [ 3) Ensure that 480 Volt bus load shedding has occurred. 4) Ensure isolation of de-energized 480 volt busses. I 5) Ensure diesel generator feed breakers are closed. l 6) Ensure the 480 volt 18 bus tie is closed to the bus of the first functioning diesel generator. l 7) Ensure that the automatic load sequencer is operating to start essential equipment. If.not, manually start loads I according to the program list. 8) If power cannot be re-established to the security system start the ACM diesel generator and transfer security system loads to the ACM system. ~ 8oMM 372 22 3443 ,g

PUBLIC SERVICE COMPANY OF COLORADO g; Ace 5 1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue 51 Page 17 of 18 The security system backup power is supplied from the ACM system and requires operator action to transfer from normal power to backup power. 9) If power cannot be re-established to sufficient equipment to mair.tain active core cooling proceed with Emergency Procedure G, " Extended Loss of Active Core Cooling." If the cause of the loss of plant power was of such nature as to prevent the use of normal methods of removal of decay heat the procedures of Emergency Procedure G provide alternate methods of heat removal and preservation of PCRV integrity. 10) When condenser vacuum improves and condensate pressure is adequate, saut power operated hot reheat relief valves. Condenser vacuum is affected by the reduction of circulating water flow. When the diesel generator is operating, one circulating water pump is restarted and vacuum should improve permitting steam to be dumped to the condenser instead of vented to atmosphere through the power operated hot reheat relief valves. Returning the / condenser to service as quickly as possible is important s to conserving the feedwater supply. 11) When power is available to the small condensate pumps, re-establish makeup to the deaerator. This is necessary to maintain feedpump suction pressure. 12) Initiate shutdown cooling. The operator selects the appropriate mode of cooling to suit the specific circumstances he faces. l 3.8 Ensure Feedwater Flow Remains at Pre-Scram level ~Until Reheat Temeerature at 975*F. This action is backup to PPS. The operator should take manual control of feedwater flow if the automatic system does not respond as programmed. REPORTING / ACTIVATION 4.1 The event as written should be classified as a "Sionificant Event" Unless the listed symptoms ceteriorate further, this event should be reported within one hour to the NRC in accordance with the Significant Event Reporting Procedure. I. somu 372 22-3s43

PUBLIC SERVICE COMPANY OF COLORADO gp App 5 1 FORT ST. VRAIN NUCLEAR GENERATING STATION Issue 31 Page 18 of 18 I l 4.2 The event as wri ten, is a "Sienificant Event" If there is a failure of ciesels to start anc loac. it may orceress to an ALERT emercency class. If diesels start and load, report this incicent within one hour to the NRC in accordance with the Significant Event Reporting Procedure. If the diesels do not start and leid, classify tne event as an ALERT, and implement RERP imolementing procedure CR-ALERT. 4.3 The event. as written, is a "Stenificant Event", orovidine that tne leaKace was seconcary coolant. NOTIFICATION OF UNUSUAL EVENT. or nioner, emercency class if leakace was c rima ry coolant. If the leakage was secondary coolant, report snis incicent within one hour to the NRC in accordance with the Significant Event Reporting Procedure. If the leakage was primary coolant, classify the event as a NOTIFICATION OF UNUSUAL EVENT, and implement CR-UE RERP imp'lementing procedure. 4.4 The event would be, as a minimum NOTIFICATICN OF UNUSUAL EVENT, cecendine ucon the maanituce of the release of crimary coolant. (. Table 1.2-4 in EP-CLASS to determine if the event is more Review Item 3, Table 1.2-2; Item 8, Table 1.2-3; and Item 1, serious than a NOTIFICATION OF UNUSUAL EVENT. Implement CR-UE for UNUSUAL EVENT class, or CR-ALERT for an ALERT or higher class. 4.5 This item must be considered on a case by case basis decendine uoon tne reason for Manual Scram. See Item / of Tacle 1.1-1 of EP-Class for details. Table 1.1-1 of EP-CLASS in Item 7 lists cases whereby a manual scram does not constitute a "Significant Event" If the event does not meet these exceptions, report the incident within one hour to the NRC in accordance with the Significant Event Reporting Procedure. i ., 25

PUBLIC SERVICE COMPANY OF COLORADO EP 5-2 FORT ST. VRAIN NUCLEAR GENERATING STATION I55'Je 50 Page ' of 3 TITLE: TWO LOOP TRCUELE SCRAM. WITH A TROUELE ALARM IN CPERATING LOOP ISSUANCE AUTHORIZED / j BY P0RC PORc 4 6 4 MAY 131982 [Ff7,EC"05'5 81 9gv g ( l l -B 26-

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PUBLIC SERVICE COMPANY OF COLORADO E? APP 5-2 Issue 50 FORT ST. VRAIN NUCLEAR GENERATING STATION Page,. o, 4 INTRODUCTION This emergency procedure is an extension of Emergency Procedure E-1 " Reactor Scram". The action item steos presume one coolant loco has tripped and been isolated. The reason for the first loop trio could I be for any number of causes and it is futher assumed tha: the loco is not available for active core cooling. With trouole in :ne remaining loop and Two Loop Trouble Scram, Emergency Procedure E-2 is a guice for the operator to insure that active core ecoling is ei:ner restored or continued on the remaining loop. OISCUSSION OF SYMPTCMS Also see discussions of symptoms in Emergency Procedure "C" 1.1 Helium Circulator Trioced A & 5 105A 4-5 OR C & 0 1050 4-5 i This condition could result from a numoer of causes due to individual circulator trips and loop trip or fixed feedwater flow low. In the latter case, trip of both circulators in the remaining loop is not inhibited. 1.2 Steam Generator Penetration Pressure Hion. Loco 1 105A 2-6 OR Loco 2 T050 2-6 1.3 Reheat Header Activity Hich Loco 1 105A 2-5 OR Loco 2 "T050 2-5 1.4 Main Steam Temeerature Low Loco 1 105A 4-6 OR Leco 2 T050 4-6 1.5 Pioe Ruoture Loco 1 105A 3-6 OR Loot 2 T050 3-6 These conditions in the remaining loco will result in Two Loop Trouble Scram but the loop continues in coeration to orovide active core cooling. now an 12 sen -B 29-

PUBLIC SERVICE COMPANY OF COLORADO EP 3: 5-2 Issue 50 FORT ST. VRAIN NUCLEAR GENERATING STATION 3 age : Of a DISCUSSION OF IMME0! ATE ACTION 2.1 Insure immediate actions fer reactor scram (E: 3-11 comolete. These actions provide backup for the PPS automatic s: ram and insure active shutdown core c: cling. 2.2 Restore active core :: cline. With all four circulators trisced, active core :: cling must be restarted. It will be necessary to clear at least one circulator trip and possibly correct the condition tnat caused the trip. I DISCUSSION OF FOLLOWUP ACTICN 3.1 Reduce feedwater cressure to about 350 osie and continue coolcown on coeratinc loco. l With a Steam Generator Penetration Pressure High alarm in the operating loop, this action will reduce the criving ( force behind the leak. The reduction in feecwater pressure is acecmolished by recucing the setpoint on I PC-22129 or PC-22130, or by opening relief valves PV-22167 or PV-22168. 3.2 If oossible, continue coolcown on affected loco until recion outlet temoeratures are 900 cecrees Fanrennett or i less. Otnarwi se, croceec to Steo 3.5. l l l PCmM 372 22 3643 -B 30-

PUBLIC SERVICE COMPANY OF COLORADO EP AP: E-2 FORT ST. VRAIN NUCLEAR GENERATING STATION {ssue30 f -age s c7 3.3 If leak is in *eheater. isolate reneaters and sta*: circulators en celton wneel. Continue coolcown usinc EES section of steam canera:ce. I With Hot Reheat Header Activity High alarm in :ne operating loop and radioactive gases being released from l the plant, it is cestrable to both maintain active core cooling anc to isolate One leak. Not isola-ing :ne l affectec reneater until core region outle; temoeratures I are 900 degrees Fahrenheit or less will assure :na: :ne i reheater remains well within its design :emoeratures. 3.a If leak is in Main Steam cicine. establish flow throuch power coeratec relief valve (s). This will recuce the amount of steam releasec directly into plant buildings and help crotect plant ecuipmen; from damage. 3.5 If core cooline cannot be maintained or restored. croceed to EP-G. Extenced Loss of Active Core Cooline. EP-G references procecures " Cooling Using Abnormal Procedures" and " Safe Shutdown and Cooling with Hignly k l Degraced Conditions" which provide many alternative means of obtaining active core cooling. I REPORTING / ACTIVATION l 4.1 The event would be recortable as a "Sienificant Event" as I written. l Any manual or automatic scram from >2*; power is recortable l as a "Significant Event", and should be reported to :ne l NRC within one hour in accorcance with the Significan: l Event Reporting procedure. l. 4.2 The event would be recortable as a "Sicnifican: Event" if l core cooline coulc ne maintained or cuicKly restorec. If l

ne event cecraces to a LOFC for creater : nan two nours 1

( from 10C'. oower) :ne event ceccmes an ALERT. or niener. l emeroency class. l If circulation is maintainec/cuickly restored, repor: :ne l event within one hour to the NRC in accorcance with the l Significant Event Reporting procedure. Otherwise, l implement RERP imolementing procecure CR-ALERT. PonM 372-22 3643 -B 31-

PUBLIC SERVICE COMPANY OF COLORADO EP AFF 5-2 Issue 50 FORT ST. VRAIN NUCLEAR GENERATING STATION

age i of 4 l

4.3 The even: would resul: in an unolanned -elease. and weu1: l be. as a minimum, a NOTIFICATION GF UNU$UAL EVENT I Deoending upon the magnitude of the release, :ne even: l will ce as a mimimum, an UNUSUAL EVENT. Refer to Tables I 1.2-2, Item No. 4; 1,2-3, Item No. 8; or 1.2-4, I em No. 1 I of EP-CLASS for further cetails. Imolemen RERP l imolementing procecure CR-UE for an Ut! USUAL EVENT, I CR-ALERT for an ALERT or higher emergency classification. C.. 4 0 4 = c a u 272 22. "" -8 32~

i PUBLIC SERVICE COMPANY OF COLORADO 5: c FORT ST. VRAIN NUCLEAR GENERATING STATION !$5ue 50

age ; of 3 I

TITLE: LCC: SHUTCCWN ISSUANCE AUTHORIZED BY PCRC EFFECTIVE PORc 4 6 4 MAY 131982 OATE 5-t 4 - 8L Review ( l l I i t 86mu 372 22 3e42 l -B 33-

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PUBLIC SERVICE COMPANY OF COLORADO EP APD C FORT ST. VR AIN NUCLEAR GENERATING STATION {ssue50, : rage. 0, i INTRODUCTION There are a number of potential preolems and : ncitions involving tne steam generators, tne circulators, the steam oiping and/or :ne control system -hat recuire shu;cown of major sections of the plant. Because of the two loop cesign, many of these problims anc :enci: ices can be hancied without taking the plant completely out of service by snutting down only one loop of :ne Primary anc Se::ncary COclant Systems. Loop shutdown may be initiated manually by the Operatcr er automatically by :ne PPS. Regardless of how they are initia:ac, all icop shutdowns curing power coeration result in the fellowing actions ceing automatically performed by the Plan-Pectective System: 1. The het reheat s:cp check valve is closed. (This causes trios of bcth helium circulator steam drives !SS in power.) 2. The raciation moni or sample line valve is closed. 3. Control signals for the het rehea: temperature centroller and the feedwater valve differential controller are generated, as des:ribed in cetail below. 4 The loop feecwater control valve is closed. 5. The loco feecwater stop valve is closed. 6. Both helium circulator steam crives are tripped. 7. The helium circula:Or byoass block valve is closed. l The trip of both circulator steam crives, in turn, crocuces signals to: 1. Close coth circula cr steam drive speed centrol valves. 2. Cicse both circulator steam crive outlet block valves. 3. Close soth wa er turmine outlet steam trap isola-ion valves. 4 Close tne loop main steam stop valve. 5. Close ne loop eenea er attemperator flow ::ntrol valve. 6. Close the loop reneater attempera:Or feedwater :le:k valve. 80mu 372 22 364 -8 36-

PUBLIC SERVICE COMPANY OF COLORADO EP AP: C FORT ST. VRAIN NUCLEAR GENERATING STATION {ssue50,, -age c o : ] 7. Reduce uroine lead to 50*4 of the inittai value..wni:n produces a signal tnrougn the first stage pressure to :ne flux controller to reduce reactor power. 8. Adjust main steam cressure controls as described below. 9. Adju'st feedwater controls as described below. The remaining loop continues to coerate a: accroximately its ini-ial power level because :ne load setting on :ne :urcine governor is automatically reduced to 50*; of its initial power level at IC'; mer second. This balances the turbine steam flow wi n the operating I loop's steam flow at its initial level. At the same time, the gain is doubled on the cutout of the main steam pressure controller, so tha: the feedwater control valve will responc twice as fas: to i pressure changes, making its response compatible with the recucec steam delive y capcacity. 1 If, at the time of loop shutdown or curing subsequent operation. the remote setpoint for the coerating feecwater controller becomes 50*; or less of rated loco flow. the setpoint will be locked and furtner reduction of feecwater flow can only be accomplished by placing the feedwater controller in local se: point or manual. This action is [ taken to prevent the total steam flow from dropping celow 25'; of i l rated, the minimum steam flow required to keen the main turoine in operation. The feecwater dif.ferential pressure signal across the snut:cwn loco feedwater control valve is elimina:ec so that the boiler feed cumo steam turoine governce; responc only to the differential pressure across the coerating festwater control valve. As the main turoine governor valves close towarc 50*J of the ini.ial power, the feed forwarc signal from the first stage cressure to :ne flux controller will cause the automatic control roc, and probacly the six pre-selected runback rods, to be inser.ed to reduce One reactor power to meet main turbine steam flow requirements. After the transient nas passed, the coerator mus manually adjus: the snim rods, so as to keep the automatic control roc in a position of control and to allow witnerawal of the runcack rocs. l The het reheat steam temoerature inout to the renea steam temperature controller-from the shutdown loop is automatically

ancelled.

The setecint of the not reheat steam emperature

en: roller nen becomes the average of the six mocules in :ne operating loco.

som sn s2 seas -B 37-

PUBLIC SERVICE COMPANY OF COLORADO EP AFP Issue 50 FORT ST. VRAIN NUCLEAR GENERATING STATION c ge 3 of 3 a DISCUSSIC.l 0F SYMFTCMS SYMPTOMS l 1.1 Helium Circulater Tricoed A & 5 I-05A. 4-5 C & ) I-050, 4-5 Be: ruse active core ecoling must be mair. ainec, only one loco is permittec to be shu cown. ~his is ac:cmolfsnec by "first in with lockout" circuitry wnich icen-iff es -he I loos na: had -he firs: trip signal and prevents snut own of the remaining loop. I 1.2 Steam Gene-ator Dene ration D-essure Hien Loco 1 I-05A 2-6 Loco 2 I-050. 2-6 This trip is caused by a pipe rupture wi:nin the steam generator PCRV penetration. In ei-her loop, a::uation of I at least-two of three pressure swit:hes se: a 757 psig will :ause loco snutdown. To minimize the amount of steam / water inleakage to the penetration, a steam / water dump is initiated. In addition, the penetration inerscace ( l clock valve (HV-11151 or HV-11152) to the penetration is closed to prevent moisture backflow to the purified helium system. The pene ration is prote::ed against over-pressuri:ation by relief valves. The rip point is set above no rmal coera-ing pressures bu celow penetration relief valve setting. l 1.3 Reheater Heacer Activity Hich Loco 1 '-05A 2-5 Loco 2 I-050, 2-5 Each loop has three dete: ors monitoring radia-ton from the hot reheat neader. If two of the :nree de:e:: ors l sense radiation levels in excess of 5 mR/hr, a loco shutdown is initiated. The high radiation level is :aused by a reheater :::e rupture resulting in leakage of primary :colant inte :ne reheat steam system. Isolation of -he helium leak requires shutting he rehester outlet stop : heck valve anc a loom shutdown to minimi:e the amount of a::iv':y introcuced into ne steam system.

ollowing loco snu cwn and ecuali:ation of cressure in :ne renea ar w':n -ne primary coolant, =cisture may diffuse,into :ne crima y ecolant sys;em.

The-efo-e. an innib': is int-o:uce: *n the moisture monito 'n; sys em ucon ce e: 'on of " 'gn Reheat Leacer AC-ivi y" c creven; an unne: essa y " Steam /wa er cume" a som 272 22 3sas -8 38-

PUBLIC SERVICE COMPANY OF COLORADO E: AP: C fag"e 83 I 0 FORT ST. VR AIN NUCLEAR GENERATING STATION a eo If a small leak in the reheater develooed over a cer',oc of l time, activity would be alarmed on I-05, either canel "B" ( Loop I) or panel "C" ( Leop II) a: = 1 mr/hr acove background by the snine monitors or by :ne reheat samole monitor. Reheat sample monitor activity is recor ed on l the I-05 (RR-2263/2264) panel. Activity snould also snow I uo at the SJAE air disenarge and will be inci a ac l (RI-31193) and alarmed on I-05 and recorced on I-14 l 1.4 Main Steam Temeerature Low Lees 1 I-05A, a-6 Loco 2 I-050. 4-6 Thermocouples in the main steam heacer on each icop sense l the temperature in :nat loco anc cause loco snu:Cown on j the low tempera ure loco, if :ne :emoerature is less snan 800*F, and ne difference between loco temoeratures exceecs 50*F en a leas: two of the three thermoccuoles. Low main steam header temperature in a loop is incicative either of a feecwater valve or controller failure which results in excessive loop feedwater flow or a deficiency of primary coolant flow. As temperatures decrease celow 800*F, this situation eculd result iimeisture carryover ( with resulting mechanical damage to the turbine. 1.5 Pice Ruoture Loco 1 I-05A. 3-6 Loco 2 1-050. 3-6 [ This situation requires racid a::fon to isolate the leak, so as to minimi:e the pressure and temperature builcuo within -he PCRV succor ring area. A co=cination of ultrasonic noise dete : ors with pressure or temperature instruments is used to initiate PPS action to shut cown the leaking loop. The ultrasonic instruments detect nign frequen:y noise generated oy the turculence associatec with es acing nigh Dressure steam from leaks. Each channel contains we l identical ultrasonic probes an: amolifiers, one for sacr loco, energi:ec by a common power supoly, i som an 22 3*'8 -B 39-

PUBLIC SERVICE COMPANY OF COLORADO EF A?: 0 l FORT ST. VRAIN NUCLEAR GENERATING STATION {ssue50,, -age : c: There are three ultrasonic dete:: r :nannels uncer -he PCRV supcor: ring. The sensing probes are moun ac on opposite sices of a flat reflector surfa:e locatec in a vertical plane that divides the loops. A steam leak in the piping of one loop should cause One ultrasonic cetec ces on that loop's side of tne refle:: r surface o trip. The rio se: point is an ultrasonic noise level of twice background. The ultrasonic tric alone c:es n:: cause loco shu:cown until it is ac::meaniec by trips of either the pressure or tem:erature ce:ectors. Loco shutdown o::urs wnen two of three ultrasonic cetecters agree on wnich loop is leaking anc two of the :nree l pressure cetec ors or two of the -hree temoerature I cetec:ces trip, confirming the presence of a major leak. The pressure dete:: ors trip when tne pressure insica ne PCRV suoport ring increases by Zh inches of wa er a:ove the reactor builcing reference pressure. The emaerature cete::ces trip at 130*F. When a loop shu cown due to steam leakage uncer the PCRV l oc urs, the operator should verify that :ne leak has been successfully isolated. Verification that steam leakage has been arrested is aided by the pressure and temoerature recorders located on the I-09 panel. These re:Orcers are ( started when the pressure at any sensor location exceecs 2 inches of water or the temperature at any sensor location exceeds 12C'F. For steam pipe ruptures outsice of :ne PCRV, :ne I cetectors, se::oints, and trip logic are identical :: those usec under the PCRV. Because of the si e of :ne PCRV, two separate noise detector channels are used, one on the north sice and one on the scutn. As in the case of the steam pipe rupture uncer the PCRV, it is im:ortant that the coerator verify tna the PPS action has isolated

• 09 the leak. The cressure and temperature recorders on l

aic in this verifica-ion. l somu sts. 22. see -B 40-

PUBLIC SERVICE COMPANY OF COLORADO EP Apo C !ssue'50 FORT ST. VRAIN NUCLEAR GENERATING ETATION age 6 of 5 1.6

eeewa er :iow Low Looe ~. I-055, 1-1 Loco 2 I-05C. 1-3 Feecwater flow is monitored for each loco by the 705.

l With ne Interlock Sequence Switch (ISS) in :ne ?CWER position and an incication of less tnan 20% feecwater flow, ne loop will be triocec and isciatec to orotect the I affected steam generator. Also, :ne automatic feecwater flow control is programmed to main ain feecwater flow acove 25%. Should a controller or valve malfun:: ion cause l a flow less than 20% of rated, ne loep is tripoec and isolated. This prote::s the affected steam generator f-em either an over emeerature transien: or from :nermal snock cue to over-cooling. DISCUSSION OF IMMEDIATE ACTIONS 2.1 Insure turbine runback and reac or cower dscreases to acoreximately one-nalf :ne crevious cower ievel. The operator acts as a backup to PPS action that redu:es steam demand to the level that was being procuced by one ( loop. 2.2 Insure circulators in shutcown loco stco anc :neir nelium outlet valves close. These actions backup the PPS anc insure primary coolant I flow in :ne snutdown loop is isolated. See:ifically, co:n I circulator speed control valves, steam cutle; anc helium l outlet valves, and common bypass block valves should ce closed. 2.3 If the helium cutlet valves in the shut:cwn loco do not I close. scram reac cr oer Emeroency Procecure B-1. A stuck open helium outle valve would ce cetected Oy I higner sceed inan the normal self turoining sceec of aceu i 300 to 400 rom as a result of reverse flow. Some of :ne l primary coolant helium flow would be dive *:e: from :ne operating loop as baenflow througn :ne snuscown loco. Therefore, tne procer action is to scram ne rea::or to recuce neat procc:-ion to cecay neat sta us. acau sta. 22. sus .B 41-

PUBLIC SERVICE COMPANY OF COLORADO E.D AP: C FORT ST. VRAIN NUCLEAR GENERATING STATION .ssuej0 -age e o: 2.4 Insure se:encary flow in snute:wn 100: *solated. l This action is a cackuo to the DDS and prevents :nermai shock to the steam generators. The feedwa.er inlet block, control, and reneat attem:eration block valves shoul ce closed. The main steam st:p theck, reneat stem stop :he:k and sammle valves should be closed. 2.5 Insure at leas: one cume valve oesn. This is acolicable to the Steam Genera:ce Pene rati:n l Pressure Hign condition, only where tne steam / water cum: is part of :ne PDS automatic a:: ion. The loop cum: limits the s,eam inleakage into the pene, ration cavity nus limiting :ne pressure buileu:. DISCUSSION OF :0 LLC'dVP ACTION 3.1 Insure that steam oenerator eenetration interscace valve is closec in :ne snutcown looe. The operator acts as backup to PDS action nus preventing steam from flowing back into :ne purifec helium system. ( 3.2 If dumo tank eressure exceeds 250 : sic or dume tank activity is incicatec or wnen cume is :cmciete - manually close cume vaive. l Dump ,ank pressure and/or activity increase incicates :ne presence of helium from ne primary ::olan system. Tne dump valves are shut to minimize ne s: read of a::ivity. The duma valves shoule also be shu wnen :ne steam water dump i s ::mplete. l 3.3 Reduce clant cower te recever shutcowe looe. It is necessary to recu:e plant :ower to re:over :ne shutdown loop. An orderly plant snute:wn is prefer-ed to a reactor scram wnenever possible. Loo: recovery I conditions anc cro:ecures are soe ified in CPCD V**I.

ce l

the case of Reneat Heace* Ac-ivity Hign, it is necessary l to carry out the notifica-ions in Steo 4.3 prior l resuming routine operations. l 3.4 Insure stacle c eratice of crimary : elan: system. With loop tric due to main steam tem:erature or fee:wa er flow preslems, :areful a: ention snoulc be giver to sta:'e operation of :ne primary an: seconca y :: elan: sys ams. somu 372 22 3443 .B 42-

\\l PUBLIC SERVICE COMPANY OF COLORADO EP r,co : FORT ST. VRAIN NUCLEAR GENERATING STATION .ssue SC

age 5 cf S l

3.5 !? steam ieak nas not been isciatec. manually s:-a.- :9e l reac or oer Emeroency :-oce:ure 5-1. The PPS action snould have isola.ed one lear,ing loco. However, if the steam ieak was not isola:ec, .ne -ea :or should oe scrammed to minimi:e ne amoun-of nyn temperature steam leakage and consecuent camage to ciant equipment. l REPCRTING/ACT!VATICN l 4.1 The event. as written is a "Sienifican: Event." i ihis event is reportable as a "Significan. Even:" as a j result of PDS a : ions resulting in a loco shu::own. an l should be reported to :ne NRC witnin one nour, in l accordance with the Significan: Event Reporting cro:ecure. I 4.2 The event, as written, is a "Sienifican: Event" mu may [ eesult in oroolems for crimary cooiant/fue'. temoeratu-es. l anc snouic ce evaluatec as recu1 rec. l This event is reportable as a "Significan Event" as a I result of PPS actions resulting in a loco shutcown, and t, I should be recorted to the NRC witnin one nour in I accorcance wi:n the Significant Even: Reporting pro:ecure. l I em Numcar 5 on Tacle 1.2-1 of EP-CLASS ciscusses I possible effe::s of this symotom, wnich could even: Wally I require elevation to a NOI!:ICATION M NUSUAL EVENT l emergency class. l 4.3 The event would result in an uncianned radioloc'cai I reiease. anc woulc ce. as a minimum, a NOTIFICATION C l UNUSUAL EVENT emeroency class. l Evaluate the magnituce of the radiological -elease l utili:ing work snee s containec in RERP imolemen-19; l procedure CR-UE, as recuired, to assist in assessment of I release magnituce.

urtner discussion of alternate i

emergency classifications is contained in Item 2 on co n i Taole 1.2-2 an: Table 1.2-3 of EP-CLASS. If :nis even: 's I an UNUSUAL EVENT, implement RERP imolementing cro:ecure 1 CR-UE. If ne event is an ALERT or nigner, imolemen-l CR-ALERT. 8cmu 272 22 2s43 .B 43-

ENCLOSORE D March 28, 1985 Search Technology, Inc. 25b Technology Park / Atlanta Norcross, Georgia 30092 Public Service Company of Colorado 16805 Weld County Rd. 19 1/2 Platteville, Colorado 80651 Attention: Mr. M. E. Niehoff

SUBJECT:

Walkthrough Validation of Response Times for Steam / Feed Pipe Ruptures

Dear Mr. Niehoff:

- In response to NRC questions regarding the four-minute basis for environmental qualification of vital equipment, NED attempted to validate the four minute isolation time. The result of this NED analysis is a report (NDG-84-0631) that describes a total of ten design basis pipe breaks, along with the plant and operator actions that would result from those breaks. For each plant or operator action, the NED report lists an approximate elapsed time that is based on a combination of engineering analysis and interviews with site operator training instructors. In order to determine whether the time estimates in the NED report are reasonable from an operational perspective, walkthroughs of the pipe break scenarios were conducted in the Fort St. Vrain Control Room Mock-up. These walkthroughs were conducted by the author using a licensed operator with 3 years of on-the-boards experience at Fort St. Vrain. Also in attendance during the walkthroughs were a plant equipment operator and the authors of the NED report. In general, the walkthroughs and interviews support the PSC contention that a design-basis steam or feedline pipe rupture can be isolated within four mintes. Certain assumptions in the NED report are overly conservative in that they result in much longer response times than are credible. For example, each operator and plant action is assumed to take place serially. In reality, the East-and West-end operators work in parallel especially during emergencies. I w

{ In one case (hot reheat line break downstream of the Pressure switches), the conservatism of the NED report results in an elapsed time of four minutes until the leak is isolated. This assumes that the Reactor is not scrammed for 140 seconds after the Turbine trips due to loss of condenser vacuum. From the walkthrough, however, it is obvious that the operators would manually scram the Reactor in well under one minute after such a severe vacuum loss. This one change would considerably reduce the postulated time required to isolate the rupture. One area in which the NED report appears to be non-conservative involves the time estimated for operators to read emergency procedures and decide on the proper follow-up actions. In all cases, the NED report estimates 30 seconds for the operator to pull an emergency procedure and determine the proper course of action. In these cases, however, emergency procedure B1 is applicable. It is considered to be a relatively involved procedure and some of the discussion concerning follow-up actions is ambiguous. The operator who participated _in the walkthrough estimated that approximately two minutes is a more realistic estimate for this step. After conducting these walkthroughs, it is apparent that, at least for design basis pipe breaks, the four-minute isolation criterion is likely to be met. However, there is one question that relates to design basis pipe breaks that persists and probably should be addressed by PSC. That question, at l_ east in my own mind, is how long after a break is " isolated" might heat still be input to the Reactor or Turbine Building. As to this question, consider a cold reheat line break in the Turbine Building. The NED report considers this rupture to be isolated once the operator closes the main steam bypass valves into the bypass flash tank. This is, in fact, the last thing the operator can do and closure of these valves does stop the input to the flash tank, which is feeding the line break. It seems intuitive to me, perhaps with no technical basis, that, even after the input to the flash tank is closed, BTUs would still be available to exit through the cold reheat line. Hence, even though the line break is " isolated" in under three - minutes (according to the NED estimate), the Turbine building temperature may continue to increase, or not start decreasing, for some time beyond this point. I'm not technically astute enough to know if this is a problem, but I wanted to raise the question. s w.. b

r ] , - In summary, it looks as if the NED conservatism and non-conservatism cancel each other out,- rendering their time estimates reasonable for the design basis pipe breaks they analyzed. Very truly yours, hb M M. E. Middox, Senior Scientist MEM/sa cc: ~D. Glenn S. Marquez 9 x.

p! ~ r ENCL OSURE C b d, ,C.iCEIVED aus 1 6 1984 ' a.' l -_ : b ~.J .. :.c ...n-:: ;L :.~.u :, ~. .a. GA Trz: :;ies Inc. P o. Box 81608 SAN DEGo. CAUFoRNIA 92138 (619) 455 3000 August 9, 1984 GP-2325 Mr. H. L. Brey, Manager Nuclear Energy Division Public Service Company of Colorado 2420 hest 26th Avenue, Suite 100D Denver, CO 80211

Subject:

FSV Steam Break Accident - 10/20 Min. Leak Termination Analysis

Dear Mr. Brey:

At the request of Mr. J. Reesy, GA performed an analysis of the reactor and turbine building environmental temperatures for steam line Dreak accident conditions. The analysis was based on leak termination after 10 minutes and after 20 minutes. The results, as summarized in Table I (below), show higher environmental temperatures than for the 4-minute case on which previously reported results were based. TABLE I PEAK ATMOSPHERIC AIR TEMPERATUARES 20 FEET FROM LEAK SOURCE Steam Leak Termination Peak Atmosphere Building / Reheat Stesam Line Minutes Temperature, OF Reactor / Cold 4 429 Reactor / Cold 10 429 Reactor / Cold 20 442 Reac'Ar Hot 4 506 l l Turbine / Hot to 571 Turbine / Hot 20 612 deft'-@f e ld ' e ' LL v

g + L l 1-GP-2325 prgs 2 The work is documented by SD&PD:CJR:083:84 which is transmitted for your information. Should you have any questions regarding this transmittal, please contact R. Rosenberg at (619) 455-2174. Very truly yours, ,a 'M_ D. J. Kowal, Director Fort St. Vrain Services Enc'losure cc: J. Reesy / 4 e i )

(MS-333 j Enclorura to GP-2325 GA CORRESPONDENCE FORM r 1476 r IN REPLY g FROM: C. Rodgers REFER TO: SD&PD:CJR:083:84 I.,,Rosenberg '- DATE: June 28, 198k R TO:

SUBJECT:

Fort St. Vrain reactor building cold reheat and turbine building hot reheat steam line breaks with 10 minute leak termination Summary ' Environmental qualification of safety-related equipment is required at Fort St. Vrain to ensure plant safety. Postulated accidents that may signifi-cantly alter the environment inside the reactor and turbine buildings and possibly affect equipment performance are studied. Operating conditions in the .' buildings are then predicted for these events. Qualification tests are per-formed on safety related. items to ensur,e their satisfactory performance during these postulated accident conditions. The atmospheric conditions inside the buildings must be compatible with the operating limits of all safe shutdown cooling equipment to ensure plant safety. One postulated accident that could cause changes inside the buildings that may affect some of these. components is a steam pipe break. In the ' reactor building, a cold reheat steam line break upstream of the loop isolation valves was considered to be the worst case. In this event, automatic controls put the plant into a shutdown mode. -Then diagnosis and manually initiated corrective action are necessary to isolate the rupture. In the turbine building, the hot reheat steam line break was considered the most, severe accident because of the pipe size and steam enthalpy. No remedial action occurred prior to the manually initiated leak termination for this event. Four, ten, and twenty minute delays from the initial rupture until manual termination of the steam flow.were assumed for both pipe breaks in this anal-ysis. The turbine and reactor buildings both have louvers that vent to the outside so pressure build-up is not considered a problem. However, temperaturc l gradients developing from the steam leak scurce may become undesirable. g .==se 4- -. =

. cms-33).- c '[unn 28,1984 4' RODGERS, Peak atmospheric air temperatures predicted in the Fort St. Vrain reactor and turbine buildings by the CONTEMPT-G code for several steam pipe leak termi-nation times at a distance of 20 ft from the source are summarized in Table 1.. The cold reheat steam pipe leak in the reactor building and the hot reheat steam pipe leak in the turbine building were terminated at,4,10, and 20 min-utes. Plots of atmospheric air temperature in the reactor building are shown in Figure 1 for the three leak termination times. Temperatures in the turbine building are shown, in Figure 2 for each leak termination. These temperatures (' .v) are all at a distance of 20 ft from the steam source.. In-both the reactor and turbine buildings the predicted atmospheric air l temperatures appear higher and remain higher for longer periods of time with successively longer steam leaks. This is because there is mor,e energy added to the environment by the released steam than can be absorbed for a steady state so air and component temperatures rise. The higher temperatures resulting from 4 the longer leaks. decay at the same rate as the four minute leak. They just begin decaying at a higher temperature so remain higher for a lodger period of time. Thus longer steam leak terminations lead to equipment exposure at higher j ' temperatures for longer periods of time. / Evaluation 1 The predicted response of the atmospher.ic air temperature inside the reactor building' to a postulated accidental cold reheat steam line rupture was requested for a ten minute and twenty minute steam leak. Temperatures are to be predicted at a distance of 20 ft from the leak source. The worst location for. this break would be upstream of the loop isolation valves. This rupture b l would cause excessive movement of the turbine rotor which will trip the turbine l l generator and low hot-reheat. steam pressure will scram the reactor. Normal secondary coolant system control actions will cause the main steam loop isola-( tion check valves to close and put the plant into a shutdown mode; however, the pipe rupture will still be. fed with steam from the flash tank. Isolation of the pipe rupture can be accomplished remotely from the control room by increas-ing the setpoints on two pressure control valves, which are located upstream of + r ..m e- -.v_-. .,y y

~ LMB-3'3) ~ ~ * ^ ~ ~ - ~ ~ ^ ,6

• RODGERS 'Junn 28, 1984 I

i the main steam de'superheaters, to a valve above the settings for the loop pres-sure ' relief valves. The superheater steam would then be discharged to the atmosphere instead of to the flash tank and the rupture would be isolated. Diagnosis and corrective action will require some time to terminate this leak. i Also, the predicted response of the atmospheric air temperature inside the i turbine building to an accidental hot reheat steam pipe rupture for a ten minute and twenty minute leak was requested. These temperatures were to be determined at a distance of 20 ft from the l'eak source. The hot reheat steam pipe rupture will dump more heat energy into the building for a limited time interval than any comparable superheat or cold reheat steam leak because of its pipe size and steam enthalpy. It was aseumed that remedial.1ction such as reactor scram did not occur before the manually initiated leak termination. Calculations made in 1972 for similar postulated pipe leaks using a four ~ minute termination were recalled to form the basis for this new study. These results were published in Reference 1. Figures 3 11 and 3 12 of this reference are reproduced here as Figures 3 an'd 4 for the re' actor bu'ilding and turbine l building atmospheric temperatures after the rupture. Table 2, line 1 shows the ' peak temperatures for the reactor building 20 ft and 30 ft from the leak and the turbine. building 20 ft from the source from this reference. A trial calcu-lation was made to reproduce these results. This would verify the model and data before proceeding with the new conditions. The CONTEMPT-G code was used to predict the atmospheric air temperature inside ~ the turbine and reactor buildings. It was retrieved from the CSD archive library reference number THSD0560. This version of the program can be used to analyze steam pipe breaks within an open containment building. The code is described in Reference 2. Two computer runs were retrieved from the April.1972 calculation files. They calculated temperatures in the reactor building 35 ft from the, leak and in the turbine building 20 ft from the leak. Peak atmospheric temperatures are recorded in Table 2, line 2 for these two cases.. These examples were rerun l using the same input data and the retrieved CONTEMPT-G code. Results are plotted for atmospheric air temperature in the reactor buildins in Figure 5 and b l

[HB-333' l kODGERS Juns 28, 1984 l turbinebuildinginh'igure6. Peak temperatures are listed in Table 2, line 3 ~ The input data for these cases were the same as before, but the retrieved CONTEMPT-G code is a more recent version. Calculated turbine building peak temperature was 469'F, 23*F higher at four minutes and the same distance than the recovered run and reactor building i peak temperature was 372*F, 40*F higher than the recovered run. These temper-stura differences -ire due to using a more recent version of CONTEMPT-G since the data input was the same. It is not know specifically what the differences are between the two code versions. The code version used for the 1972 l calculations no longer exists. However, the plots of temperature versus time for these two cases have the same characteristic shape as shown in Figures 3 and 4. The discrepancy in peak temperature was small and deemed acceptable, so this version of CONTEMPT-G was used for the remaining analysis. Data for the steam blowdown rate which includes steam addition rate and g ) i energy added is very critical for this analysis. It directly affects the atmospheric air temperature' peak and profile with respect to time. And yet j ,availabl's sources consisting of some recovered runs and Reference 1 differed about steam addition rates. e The steam blowdown curves shown in Reference 1 for the reactor building cold reheat steam line rupture (Figure 3 3) and the turbine building hot reheat steam line rupture (Figure 3.4) are reproduced here as Figures 7 and 8, respec-tively. Scaling values from these curves into tabular data for use. in the CONTEMPT-G code yields the data shown in Table 3. 'This data was supposedly used for the four minute leak termination study in 1972. However, it is con-siderably different than the values recovered from tele runs in April 1972 which are shown in Table 4. In particular, the reactor building blowdown rate at 239.99 seconds is 193 lb/sec from the recovered run and 113 lb/sec from Reference 1. This dit- - forence yields atmospheric air temperaEure profiles that are quite different. ~ Results for the reactor building cold reheat' steam pipe break for.a four minute leak termination at a ' distance of 20 ft from the source are shown using these l 4 4 6 e 4 em9 .s,.. .,,.r.,, ,._.-~,-r------v.-,.-..m..,v __y _y,e_._,

LW*MB { ,'. RODGERS..Juns 29, 1984 two rates. Steam blowdown rates from Reference 1 were used for Figure 9 yield-Il ing a pea 5c temperature of 416*F. Data from recovered runs of April 1972 were used for Figure 10 yielding a peak temperature of 429'F. Comparing these two temperature profiles with the results reported in Figure 3, lead one to surmise that the blowdown data actually used in the previous analysis was the data from f the recovered runs in April 1972 and not what is shown in Figure 7 Substan-tiating this is some data found in Reference 3, Figure 1, which is reproduced here as Figure 11. This shows steam flow through the rupture frod the initial break to seven ' minutes later. Scaling data from this curve agrees with the values for the four minute leak from the recovered runs of April 1972 listed in Table 4. This data was chosen to be the best available blowdown data for the cold rel$ eat steam pipe break. Flow was assumed to be a constant 70 lb/sec after 400 seconds for the teb minute and twenty minute, peak cases. Table 5 shows this blowdown data tabulated. The blowdown curves tabulated in Tables 3 and 4 for the turbine building hot reheat steam line rupture are different initially, but agree after about two seconds. Its initial difference does not affect the air temperatures becaus,e it occurs for such a short time period. Results are shown for the tur- .bine building atmospheric air temperature usin's Reference 1 data in Figure 12 and recove. red data in Figure 13 for a four minute leak and a distance of 20 ft from the source. Peak temperatures and curve shapes are the same. Another source to justify this data was found. Blowdown data from Reference 4, Fig-ure 1 is reproduced here as Figure 14' which supports the recovered data of j April 1972. Therefore, the data lis'ted in Table 4 was chosen to be the best available blowdown data for the hot reheat steam pipe break with four minute termination. Flow was assumed to be a constant 625 lb/sec after four seconds 4 o - Table 6 shows the blowdown data for hot reheat pipe for each case studied. rupture with ten minute and twenty minute leak termination. fi Heat transfer surface areas of major components (heat sinks) in each building for various distances from the leak sou ce were determined in the 1972 study. The derivation of these areas is described in Reference 1. This data wastabulatedforthereactorbuildinginTable31ofReferencek'andthetur-bine building in Table 3 2 of ' Reference 1 and are reproduced here as Tables 7 S i i m-

ou vers NODGERS ~ funa 28, 1984 and. ' 8, respectively. These values were used directly as input for the CONTEMPT-G code in this study. l ~ Reactor building atmospheric temperatures were then predicted for ten minute and twenty minute leak terminations. Blowdown data was taken from Table 5 and the heat transfer surface areas of each heat sink were taken from Table 7 - for a 20 ft distance from the postulated leak. These results were plotted along with the four minute termination results in Figure 1. Peak temperatures are listed in Table 2, item 4. Turbine building atmospheric temperatures were also predicted for ten minute and twenty minute leaks. Table 6 blowdown data was used and Table.8 heat transfer surface areas of each heat sink for a distance of 20 ft from the leak were used. Results were plotted along with the four minute termination predictions in'Figu a 2. Peak temperatures are listed'in Table 2, item 4 Discussion In the reactor building the predicted atmospheric air temperature for an a'ccidental cold reheat steam pipe break 20 ft from the leak source is summar-r- ized in Figure 1. The postulated four minute leak appears to cause the air temperature to reach a maximum of 429 8F at four minutes,,then decay very rapid-ly for several seconds. Following this initially rapid decay period the decay rate slows. Initially the pipe rupture releases a jet of steam into the atmos-phere which expands into a steam cloud. This event adds a lot of energy into the immediate area which causes the air temperature to rise rapidly. As the steam cloud expands, components in its path begin absorbing some energy. The energy of the steam released into the air reduces over time. This siv.[a the i rise in temperature. When the leak terminates suddenly the heat source is removed so temperatures drop quickly. They then decay tw.rds an equilibrium as the released steam energy is dissipated. This curve characteristic shape agrees with the one predicted by the previous study shown in Figure 3. The peak i;emperature predicted now is 24*F higher than what was reported before. This is due to using a more recent version of the CONTEMPT-G code. It was assumed that this difference was acceptable. b ~

V @dj-]fs

. g,.

,RODGERS Juna 28,1984 I. The ten minute leak termination does not appear to cause the peak atmos-pheric air temperature to increase beyond the' 429'F calculated for the four sinute leak. It does, however, cause the temperatures to remain higher for a longer period of time than the shorter leak. After four minutes the air tem-l 4 perature drops to a minimum at about seven minutes, then rises a few degrees I until the leak terminates at ten minutes. After ten minutes temperatures deep sharply then decay as before. Between four and seven minutes after the rup-ture, the rate of steam energy added to the environment is decreasing. Compo-nonts in the reactor building are absorbing energy at a slower rate du' ring this time because - there' is less heat to transfer from the steam. Air temperatures begin to drop. The heat transfer coefficient, thermal conductivity, volumetric heat capacity, and surface areas of each component modeled are considered when analyzing temperatures. Some components begin to loose heat during this time 7 such as the steel decking, ducting, electrical conduits, and cable trays. These' items tend to follow the air temperature more closely than the others. 4 Between about. seven and ten minutes the rate of steam energy release is con-1 stant. This heat addition rate is more than the components and environment can absorb for a steady state so temperatures begin rising again. They rise until I the steam leak terminates. At this time air temperature has peaked at about 409'F. Terminating the leak source causes the air temperature to drop rapidly and decay towards a new steady state. l The twenty minute leak termination air temperature curve follows the same 4 j predicted curve as the ten minute leak except that after ten minutes it con-tinues to rise to a peak air temperature of 442*F at twenty minutes. It then drops rapidly and decays towards a new equilibrium. Once again the constant f steam energy added between ten and twenty minutes-is transferred to the air and i components causing their temperatures to, rise. l' It was assumed that there was an adequate cold reheat steam supply to maintain the constant steam leak energy addition rate of.70 lb/sec 1359 stu/lb from 400 seconds until either the ten or twenty minute termination. An esti-

l mate of the total volume of water / steam required to sustain these leaks was t

based on the flow rates in Table 5. The volume of water released from the pipe for a twenty minute leak is approximately 144,000 lb. There appears to be more than sufficient water available-to supply this leak. o f l j

RODGERS ' Juns 28, 1984 e In the turbine building the predicted atmospheric air temperature for an accidental hot reheat steam pipe rupture 20 ft from the leak source is summar-ized in Figure 2. The postulated four minute leak appears to cause the air temperature to reach a maximum of 506*F at four minutes then decay very rapidly for several seconds. The decay rate then slows as the system moves towards' a steady state. Initlally the hot reheat steam pipe releases a jet of energy This cloud expands into the tur-into the atmosphere producing a steam cloud." bine building atmosphere. The steam energy released decreases quickly for four seconds then remains constant for the duration of the leak. This constant heat source causes temperatures to increase inside the building until the leak is terminated. The sudden terminatics of steam energy addition causes the air temperature to drop quickly then decay to an equilibrium. This curve charac-teristic shape agrees with the one predicted by the previous. study shown in Figure 4. The peak temperature predicted now is 26*F higher-than what was reported before. This is due to using a more recent version af CONTEMPT-G. It was assumed tha't this difference was acceptable and longer leak termination times analyzed. The ten and twenty minute steam leak terminations appear to cause air temperatdres to continue rising until the leaks are stopped. This is because there is more energy added to the environment by the released steam than can be absorbed for a steady state, so temperatures rise. The predicted peak atmos-pheric air tempeiature for a ten minute leak was 571*F and for a twenty minute leak was 612*F. These higher temperatures decay at the same rate as the four minute leak. " Temperatures also remain higher for the increased steam leak terminations' for a longer period of time. It was assumed that there was adequate hot reheat steam to maintain a steam leak foi ten and twenty minutes. A constant energy source of 625 lb/sec, 1360 Btu /lb from four seconds until leak termination at ten or twenty minutes is a considerable amount of water. An estimate of the total volume of water / steam required to sustain these leaks was based on the flow rates in Table 6. l The volume of water released from the pipe for a twenty minute leak is approxi-mat'ely 754,000 lb. There appears to be more than enough water available to supply this leak.

(MS-333 p RODGERS Jun2 28, 1984 Conclusion The CONTEMPT-G predicted peak air temperature of 429'T in the reactor building appears unchanged for a four minute and ten minute cold reheat steam pipe rupture 20 ft from the source. Thus extending the leak termination time from four to ten minutes only increases the length of time components will be exposed to higher temperatures. A twenty minete leak would produce a peak temperature of 442*F, only slightly higher than the shorter leaks. The pred.icted peak air temperature in the turbine building appears to be 506*F for a four minute hot reheat steam pipe rupture 20 ft from the source. The ten and twenty minute leaks appear to produce air temperatures of 571*F a'nd 612*F, respectively. These predicted temperatures are significantly higher - titan the peak for a' four minute leak. They also maintain temperatures higher for a longer period of time. It appears from a rough estimation that there is more than enough steam 4 , supply for the long leak terminations of ten and twenty minutes. The blowdown -l rate from the cold reheat pipe rupture in the reactor building is much smaller than the hot reheat pipe rupture in the turbine building. Thus the volume of water required to supply the cold reheat pipe leak is less-than the hot reheat pipe leak. l l e e i i I e ,-e--

---

,r.

m-p - -,.,y

---

,y

,.,y- .,vr-r w*. +-rw r- ..e= r-,.

~. . cms-333 RODGERS~ June 28, 1984 I References 1. Benham, R. G., E. J. Burwell, D. S. Duncan, and H. F. Menzel, "Qualifica-tion of Fort St.' Vrain Safe Shutdown Equipment for Steam Environment Resulting from Pipe Ruptures," GA Technologies Report GA-A12045, May 30, 1972. 2. Houghton, W. J. and C. L. Turner, " Operation of the CONTEMPT-G Containment ~ Analysis Pr'ogram on t,he UNIVAC 1108 Cocputer," GA Technologies Report ~ GA-A12692Supp'lementA(GA-LTR-6[, February 8,1974. 3 Mediger. *W. To E. V. Haake "PSC: Cold Reheat Steam Line and Main Steam Line Failu[*e," Gd Technologies Internal Memo Number WH:1157:67, November ~ ,e 27, 1967. g .4. Potter, R. C. to H. F. Menzel, "PSC Predicted Leakage to Turbine Building from a Hot Reheat Line Rupture," GA Technologies Internal Memo Number l RCP:PSC SE311:72, March 6, 1972. 1 5.* Joseph, W. M., " Interactive Plotting Routine, SUPER

• PLOT," GA Technologies Internal Me[no Number SAB:025:WMJ:81, March 2,1981.

~ i i CR MAB Idi ces S. B. Inamati A." 5. Shenoy F. A. Silady e

m g RODGERS Jun3 26,1984 TABLE 1 ~ PEAK ATMOSPHERIC AIR TEMPERATURES 20 FEET FROM STEAM LEAK SOURCE BUILDING / REHEAT STEAM LEAK TERMINATION PEAK ATMOSPHERIC STEAM LINE (MINUTES) TEMPERATURE (*F) reactor / cold 4 429 reactor / cold 10 429 reactor / cold k0 442 turbEne/ hot 4 506 turbine / hot 10 571 turbine / hot k0 61k S t 0 O e S e e O s 4 9 L \\

• We h

r.,.

__.__.,._p -.,...-_-...-,,3 .-..7,___-.__ r..,

TABLE 2 PEAKATHOSPHERIdTEMPERATURES(*F) ) 20 Feet From Source 30 Feet From Sour Reactor Building Turbine Building Reactor Buildini 4 min 10 min 20 min 4 min 10 min 20 min 4 min 333 e 1,1 480

1) Reference 1 results 405 s

326 (Figure 3 g and Figure 3g2) 332 e 2.8 446

2) Recovered runs April 1972 327 469(2) 372 e 2.5

3) Steam blowdown and heat transfer surface 365 area data from recovered runs

4) Steam b1cwdown from recovered runs, heat 429 429 8 3 8 442 506 571 612 same as transfer surface area data from Reference 1 409 case 3

5) Steam blowdown and heat transfer surface 416 e 2.2 507 area data from Reference 1 392

• N:tes:

L .y (1) peak atmospheric temperature at cut-off time unless otherwise noted. (2) heat transfer areas not as reported in Reference 1 O e on.m,,

- ~ - -. -.. _ _ - _ _ - - a TABLE 3 BLONDONN DATA FOR 4 MINUTE LEAR FROM REFERENCE 1 TIIE STEAM ADDITION RATE ENERGY (ENTNALFY) ADDITION RATE hour second Ib/hr lb/sec' Btu /hr Stu/lb deactor building, cold reheat pipe 0. 0. 9.216 E+6 2560 1.252~ E+10 1359 1.389 E-4 0.5 4.896 E+6 1360 6.654 E+9 2.083 E-4 0.75 3.'492 E+6 ~ 4.746 E+9 970 2.778 E-4 1.0 2.412 E+6 670 3.'278 E+9 ~ 6.'944 E-4 2.5 1.465 E+6 407 1.'991 E+9 ~ 1 389 E-3 5.'O i.260 E+6 350 5.'712 E+9 l 2.778 E-3 10.0 1.152 E+6. 320 1.566 E+9 F a 2.778 E-2 100. 1.080 E+6 300 1.468 E+9 i{068 E+ii 113 5.528 E+8 6.666 E-2 239.99 I i C, 6.667 E-2 240.0 0. 0 0. 0.51 1836.0 0. 0 0. Turbine building, hot reheat pipe l ~ 1 746 E+7 4850 2.654 E+10 1520 0. O. 1 389 E-4 0.5 i.~004 E+7 2790 1.506 E+io iS00 ~ k.'778 E-4 1.0 6.'516 E+6 1810 h.'644 E+9 5480 ~ ~ !) 4$167 E-4 i.5 4.356 E+6 i210 6.'360 E+9 i460 5 556 E-4 2.0 3 391 E+ 6-942 4.883 E+9 1440 ~ j 8.333 E-4 30 2.534 E+6 704 3 548 E+9 1400 ~ l 1.111 E-3' 4.0 2.257 E+6 627 3.'070 E+9 i360 ~ ~ l 6.66d E-2 239.99 2.'257. E+6 627 3.'070 E+9 i360 ~ l 6.667 E-2 240.0 0. 0 0. O. ~ 1836.'O 0. 0 0." 0. 4 c.51 em.s

TABLE 4 BIAWDOWN DATA FOR 4 MINUTE LEAK FROM RECOVERED RUNS OF APRIL 1972 TIME STEAM ADDITION RATE ENERGY (ENTHALPY) ADDITION RATE hour second lb/hr Ib/sec~ Btu /hr Stu/lb R Tctor building, cold reheat pipe 0. O. 9.504 E+6 2640 1.291 E+10 1359 153888E-4 0.'5 4.'752 E+6 1320 65457 E+5 3{24 E+6 900 4.402 E+9 2.'0833 E-4 0 75 2.7777 E-4 1.0 2.304 E+6 640 3 13 E+9 6.'9444 E-4 2.5 1$44 E* 6 400 1.'h57 E+9 1 3888 E-3 5'o i.'26 E*6 350 i.'712 E+9 2$7777E-3 10.0 1.152 E+6 320 i.565 E+9 ~ 2.7777 E-2 100. IIO8 E+6 300 i.'467 E+9 h.'.424E+8 6.'6666 E-2 239.99 6.936 E+5 193 6.67 E-2 240.0 0. 0 0. 0 51 1836.0 0." O' O." I Turbine building, hot reheat pipe O. O. 1.854 E+7 5150 2.8181 E+10 1520 i.'08 E+7 3500 1.6it E+10 1500 1.388 E-4 0.5 2.777 E-4 1.0 6.'66 E+6 1850 9.857 E+9 i480 ~ ~ 4$166 E-4 1$5 4.5 ' E+6 1250 6.57 E+9 1460 5.555 E-4 2$0 342 E+6 950 4.925 E+9 i440 ~ ~ 8.333 E 3.'O 2 52 E+6 700 3 528 E+9 1400 1.111 E-3 4.0 2.25 E+6 625 3.06 E+9 1360 j 6.666 E-2 239599 2.25 E+6 625 3 06 E+9 1360 ~ 6.67 E-2 240IO 0." 0 0. 0 0$51 1836.'0 0. 0 0." 0 ~

i .[MD-333 t, RODGERS Juns 28, 1984 TABLE 5 BLOWDOWN DATA FOR COLD REHEAT PIPE RUPTURE IN REACTOR BUILDING WITH 10 MINUTE AND 20 MINUTE TERMINATION TIME STEAM ADDITION RATE ENEROY ADDITION i hour ~second lb/hr Ib/see Stu/hr Stu/lb 1) 10 minute steam termination O. 0.' 9.504 E+6 2640 1.291 E+10 1359 1 3888 E-4 0.5 4.752 E+6 1320 6.457 E+9 2.0833 E-4 0.75 3 24 E+6 900 4.402 E+9 2.7777 E-4 1.O 2.'304 E+6 640 3.'13 E+9 + 6.9444 E-4 2.5 1.44 E+6 400 1.957 E+9 1 3888'E-3 5.0 1.26 E+6 350 1.712 E+9 k.~7777E-3 10.'o i.152E+5 320 i.565E+9 2.7777 E-2 lbo. i.*b8 E+5 300 5.'457E+9 l 1.111 E-1 400. i.~52 E+5-70 3.'4'25 E+8 ~ 1.666 E-1 599.99 2.52 E+5 70 3 425 E+8 .1.667 E-1 600. O. 0 0. 0.51 1836. O. 0 0. u i l

2) 20 minute steam termination O.

O. 9.504 E+6 2640 1.291 E+10 1359 1.3888 E-4 0.5 4.752 E+6 1320 6.457 E+9 2.0833 E-4 0.75 3 24 E+6 900 4.402 E+9 l 2.'7777 E-4 1.'0 2.'304 E+6 640 3."13 E+9 1 7 i 6.9444 E-4 2.5 1.44 E+6 400 1.957 E+9 i 1 3888 E-3 50 1.26 E+6 350 .1.712 E+9 l k.7777E-3 10.0 i."152E+6 320 i.~565E+9 ~ l 2 7777 E-2 100. 1.08 E+6 300 1.467 E+9 1.111 E-1 400. 2 52 E+5 70 3.425 E+8 0 33352 1199.99 2.52 E+5 70 3.425 E+8 0 33333 1200. O. 0 0. 0.51 5836.' O.' o 0." y i l

CMB-333 2000ERS Jun3 28, 1984 TABLE 6 BLOWDOWN DATA FOR HOT REHEAT PIPE RUPTURE IN TURBINE BUILDINO 1 WITH 10 MINUTE AND 20 MINUTE TERMINATION 4 l 1 TIME STEAM ADDITION RATE ENER0Y ADDITION 4 hour 'second Ib/hr Ib/sec Btu /hr Btu /lb 4

1) 10 minute steam termination 0.

0. 1.854 E+7 5150 2.8181 E+10 1520 l 1.'388 E-4 0.'5 h.'08 E+7 3500 1.65 ' E+10 1500 ~ 1 2 777 E-4 1.0 6.66 E+6 1850 9.857 E+9 1480 l 4.166 E-4 iI5 4I5 E+6 5250 6.57 E+9 k460 5.555 E-4 2.0 3.42 E+6 950 4.925 E+9 1440 { ,8.533E-4 30 2.'52 E+ii 700 3.'528 E+9 i400 ~ 1.111 E-3 4.'O 2."25 E+6 625 3.'06 E+9 i360 1.666 E-1 $99.99 2.25 E+6 625 3 06 E+9 1360 1.667 E-1 600.0 0. 0 0. 0 5.'.'51 1836'O 0.', 0 0.' 0 ~ I i

2) 20 minute steam termination j.

l 0. O. 1.854 E+7 5150 2.8181 E+10 1520 1.388 E-4 0.5 1.08 E+7 3000 1.62 E+10 1500 2 777 E-4 1.0 6.66 E+6 1850 9.857 E+9 1480 4.166 E-4 1.5 4.5 E+6 1250 6.57 E+9 1460 5."555 E-4 i.0 3.42 E+6 950 4.925 E+9 i440 8.'333 E-4 30 2.52 E+6 700 3.'528 E+9 k 400 ~ ~ 1.'111 E-3 4.0 2.'25 E+6 625 3 06 E+9 5360 ~ ~ 7 0 33332 1199.99 2.25 E+6, 625 3 06 E+9 1360 0 33333 1200. O. 0 0. 0 0.51 1836. O. 0 0. 0 e b

TABLE 7 REACTOR BUILDINO CONTEMPT-G CODE HEAT SINKS Heat Transfer Surface Areas (fts) are shown as function of distance from postulated steam leak i i Arog) No. Heat Sink Material-PCRV 36.2 ft(2) 30 ft 25.ft 20 ft 15 ft 10.9 ft 4 4 l 1 Concrete walls, floor. PCRV Concrete 42.230 17.380 11.960 8.300 5.310 2.990 1.588 2 PCRV support ring Concrete 9.870 9.870 '6.790 ,4.710 3.020 1.700 90 3 Partition walls & floors Concrete 9.710 6.600 4.540 3.150 2.020 i.130 60< Steel 16.090 3.760 2.590 1.800 1.150 ' 650 38i . Thin steel wall (4) f 4 g 5 Composite steel wall steel i9.810 -(3) 6 Steel decking steel 43.140 19.440 13.370 9.290 5.940 3.340 1.717 '7 Structural steel, equipment Steel 28,800 i2.320 '8.470 5.890 3.770 2.120 i.15 8 Ducting, conduits, trays Steel 63.700 34.180 23.510 16.330 10.450 5.880 3.51 j 9 Piping (4) Steel ' 39.200 34.440 23.690 i6.450 50.530 5.920 3.10 i I Tctal Heat Transfer Surface Area (fts) 282.550 137.990 94.240 65.920 42.190 23.730 12.55 '.44 1 Tctal Volume (ft") 534.730 i98.220 113.100 65.450 ' 33.510 14.140 5 H=t Transfer Coefficient (Btu /hr-ft* *F) 5.0 93 13 2 18.6 28.i ' 48{0 86c (1) Section includes the total volume of reactor building below operating floor but without process area. .(2) This is the maximum radius (regional boundary) considered in the new analysis. Section includes the volume below U i PCRV which is inside and outside support ring, see Figures 3.6 and 3 7 of Reference 1. (3) The composite steel wall is an outside heat transfer surface which does not extend below EL. 4790'-0a. (4) Does not include insulated piping. i. i

~ TABLE 8 TURBINE ~ BUILDING CONTliMPT-G 00DE HEAT SINES Heat Transfer Surface Areas (ft') are shown as a function of distance from postulated steam leak II No. Heat Sink Material 70.2 ft 60 ft 50 ft 40 ft 30 ft '20 r 1 Concrete floor Concrete 30.400 24.640 19.320 14.350 9.780 5.69 2 concrete Structures concrete 5.810 4.710 3.6?0 2.740 1.870 1.09 i 3 concrete partition walls & floors ' concrete 44.6bo 36.150 28.350 21.050 15.340 8.35 4 Piping Steel 80.540 65.570 51.190 38,010 55,900 15.0@ C i 5 composite steel wall Steel 7.930 6.430 5.040 3.750 2.550 1.40 1 Steel 12.300 9.970 7.820 5.810 3.960 2,30 6 Steel decking 7 Structural steel, equipment Steel 62.710 50.820 39.850 29.600 20.170 11.79 4 89 8 Conduits & cable trays Steel 52.700 42.710 33.490 24.870 16.950 9 j 1 1 Tctal Heat Transfer Surface Area (fts) 296.990 240.690 188.750 140.170 95.520 55.6L Tctal Volume (ft') 750.000 547.200 380,000 243,500 136.800 60.81 H2at Transfer Coefficient (Stu/hr-fts..F) 5.0 6.1 7.6 10.1 14.4 23 7 4 (1) This is the maximum radius (regional boundary) considered in the analysis, see Figures 3 8 and 3.9 of Reference 1. L as (2) Does not include insulated piping. S l l -y. --,y

j ~ 4 I 1 1 REACTOR BUILDING AIR TEMPERATURE 20 FT FROM SOURCE i A 500 uar mmanen T M 4 MINUTE i 0 10 MINUTE

• S N

^,.... 2 P 5'0 bi~Nbki[ \\ ~ l H 400-E N i R s 1 g%'. C ,'~~~~ i T E 300 -- x 3( ~ 1* \\ f l E ~ R A T 200-U I -- R E 1 i l F 100-iiii iiii 6 i i i i i.. iiii ii i 0 5 10 15 20' 25 30 j TIME - MINUTES .._.... Reactor building atmospheric temperature at a distance of 20 fr. from i Fig. 1. leak source I l .i A~ ~ ' ~ ' ~

) 1 1 1 TURBINE BUILDING AIR TEMPERATURE 20 FT FROM SOURCE A 700 a uunimnos. T O M

• U is nINUTE' 1

P 600 -- ~...... H 26 nIHUTE / *1 E f R \\ '/ j I t l C 500 \\ t N,' T i. i E ~ ' Mi

l M

i' p 400 E --i 1 R [ ( A N ~ T U R 300 - i E i 1 i F 200 ii,, iiii iiii

iii, 30 O

5 10 15 20 25 i TIME - MINUTES i i Turbine building atmospheric temperature at a distance of 20 ft from Fig. 2. 1 leak source I

~ l 600 1 1 / -1 ~ l < =-ie.s l( I e / i\\ ~ m N,, 's 1 ~ / l e/n-iS, ~ ,-.- s / / [ ~ - - - , T A ~ 400 -[ [ I l N IIOTE: // R-20' N 20 FT 15 CLOSE5T 81STAIICE 70 , 5AFE SNUTOOWN EQUIPMENT I\\ / C I / n-25' l \\g \\ g / n-3o* '----.-_4 ~ { 300 A=36.2 N P 1 i b -s 1 ~~~ RE&iONAl. 000000ARY l l %EE TASLE 3.1 m I i~ l r - ti5 : AROUNO PCRV e ,8 SEE TABLE 3.1 1 ElI olN 300 4" M .ir =l v !ps - 3l-I I 'f l l 1 I I [20 o i 'I 4 5 10 15 .25 30 TIME (min.) Temperature of reactor building atmosphere during postulated cold reheat steam pipe rupture Fig. 3. accident as a function of distance "R" from possible source of steam leak (from Ref. 1). me. O e

• Wm& M

g 1 a l 1 t I l l I I' I 500 l _I W ~ 31 wi 5

l gl j

1 11 i n-no' "I I C 300 I I i, I. am L,. j jN e i $ 200 1 I U l I e. I m I I R-50' ol { R-60' EIg R=M.2* REGIONAL B3UNDARY l les l e-w l I8 IIOTE: I - r E 15 20 ft is closest distance to safe shutdown

1. ll;;

equipment I ~ I I I I I t I I 4 o e 2 3 8 5 to 15 20 25 30 l Tant (nin) Fig. 4. Temperature of turbine building atmosphere during postulated hot reheat steam pipe rupture j accident as a function of distance "It" from possible source of steam leak; no reactor scram i until leak termination (from Ref.1). i 'l M O O I <j ~~

I t 1 e l " ' ~ ' ' REACTOR BUILDING, 4 MINUTE CUT-0FF, 30 FT i A' 500 T- ~ l M o S p i H 4m E j R l I 1 C i T-I E 300 ow M a t P E 4 R l t A I i T 200-U I R [ i l E 1 i F ice i i i i iiii iiii ii,i iiii iiii 0 5 19 15 20 25 - 30 .1 i TIME - MIN j i esesses4 esississ sisese s Fig. 5. Reactor building atmospheric temperature at a distance of 20 ft from j leak source using data from recovered run. 1 i -. -. I

= y ~ TURBINE BUILDING, 4 MINUTE CUT-OFF, 29 FT .. *h Ge9 l T n g S P 500 - y ~ E R I C 4ee f T M k ~ s ~ P \\ 300-E N R A ~ ~ T U R N E F 100 i i i i iie i iii i i iiii iiii 0 5 19 15 20 25 30 TIME - MIN 05/16eS4 96:59:48 $78338 d Fig. 6. Turbine building atmospheric temperature at a distance of 30 ft from 1eak source using data from recovered run.

Y IL 1 e 2500 - m-G= ~ swo - U 3 E 8 g i t 3 lese - / enast III Tifet SCALE N Cl TEastienATieel GF STEAn LEAR l pg. +

• i t

i f At t i I g O I 2 3 4 5 'to M 120 180 234 tent.(stc) Fig. 7 Steam blowdown rate during postulated cold reheat pipe rupture in reactor building, '. enthalpy constant at 1359 Btu /lb (from Ref.1). ~ g 8 . w 1 I l

_.____..m-._..m j 5000 k 1550 i e 1500 nous i 1 I O l caemst In Tent stas.t T 1450 'D TimiseATieis OF STEM LEM .' 3000 - ~ \\ E j -A g 1 3 o N z m, o w. / i ;ogo. 1400 m x 6 a em w 1350 scoo - I i g. ,3.o o o a 2 3 4 5 to 60 120 leo 240 Ting (5tt) i Fig. 8. Steam blowdown rate and enthalpy input during postulated hot reheat pipe rupture in turbine i building without scram action (from Ref. 1). 1+ e-- ~

REACTOR BUILDING, 4 MINUTE CUT-0FF, 20 FT A 500 T M 0 g P H 400-I E R I C T N h \\ h P N i E N N R A T 200-U R E i l 4 F 100 i ii i iiii iiii

iii, O

5 10 15 20 25 30 TIME - MIN i l $76737 MAY 9,1984 15 45:48 j Fig. 9 Reactor. building atmospheric temperature at a distance of 20 fL from steam leak using blowdown data from Ref.1. 1 i

t I , e s .~. i ~...,.. ,'(.".4,e.? ' ' - l REACTOR BUILDING, 4 MINUTE CUT-0FF, 20 FT .27;"*f1500 . '; - :T - .~ ~ 1 .. - n.,. n -.g i P l -H 400- . E ~ i R I i C ~ { T j E 300-g N M ~ i j I E N ~ R 4 A T U 200-l R ~ E 4 F 100-i iii iiii

iii, ii i i ii,,

iiii O 5 10 15 20 25 30 l TIME - MIN ) ste744 Finy s.tss4 1sistise l Fig. 10.' Reactor building atieospheric temperature at a distance of 20 ft from steam leak using blowdown data from recovered' runs of April 1972.

I ..t - T:7I: t t ft. P a, j ~... M * ' I' Rscu4edt.fipe tem foluinel;.; E ' YtG M }'.. ksalk1 t 6 s Iselfcl.DeceNIfl.....*:1770 'l~ i lb 1 . I... g. .1 ;foi Presui e

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....l. ..:. I.L..,. :!..!....:..: (.s .a..:...l...... .s 11.3 A 1 E ll... .3 e. .t. i.- .:. e Fig. 11. Complete off-set rupture in cold reheat steam pipe in reactor build-i ing, 100% load, flow in the pipe between rupture and HP-turbine exhaust line (finsh tank).

iI 0 m 3 or i f t i f i 02 i T f 5 o F 2 e 0 c i n 2 a i t s i i F d F i O a. 0 2 4 t 2 1 T a i 4 U 4 i ef C 5 re i uR 1 E N t I a m i T r o U M e r i 4 pf N s m 5 - 9 e I 1 1 M ta E t i 9 ca 4 M ~ I v id i r T m en hw i G po N sd i ow I 0 6 mo D 7 tl 1 s ab L s I i TS gg U nn i B ii ds l u i E i N uk i ba I e B 5 el n R ( im U ba i T re 1 ut i Ts 2 i 1 ' 0 _ _ ~ - ~ - - _ g i 0 0 0 0 0 0 F 0 0 0 0 0 0 6 5 4 3 - 2 1 ATMOSPHERIC TEM (R URE F ~ i I ,i j 4 f: 1l

I TURBINE BUILDING, 4 MINUT'E CUT-OFF, 20 FT . } d. J..f ";--!'d.600' .....- J T - ? ..sn .o. S P 500-1 H E R I ~ C'400-( T i E h 300 E R A ~ T UR 200-E 'F 10'0 , i i i iiii iii i ii i i i i i i iiii 15 20 25 30 O 5 10 TIME - MIN i ST9722 my 8, 1984 16:35:28 l Fig. 13 Turbine building atinospheric temperature at a distance of 20 f t from t steam leak using blowdown data from recovered runs of April 1972. 4 1 e 1 i A4/PTUPE $7EAN TEMP AtB k ao b. { g sm /YM /kTSS//PE I $V/7//?/. GND/7//NS: .iStM MWENTCRV=2960 la TEMP c /402 'F" . /5tY3TU9E u 640P5'/A EN7NALPY = IS/S Ml/I! N yp anse ars=.c22 fr-t t 34tv Igm ^/53/X0 /.B D&WPED 772 224Pf/AE 47//l/J\\bVC /N 4.M/N e i /stW i 12WSTAN/~ C29 ff!ZC - i 0 / 2 3 + S to 7?Pf Jer'~*ITN.ON/2VR.? **.YC 2/.4CY 72 I N%'P Fig.,14. PSC hot reheat rupture 34" OD main header line, no scram. 9 _____._________m

r RODGERS Junn 28, 1984 l APPENDIX A COMPUTER ANALYSIS STORAGE Storage and retrieval of the computer studies performed for this analysis are described here. Computer programs and runstreams for each case analyzed l are saved in Computer Services Division (CSD) Archive Library. Printed output for each' case is stored by Records Management Department. o The computer code used for this analysis was the CONTEMPT-G code retrieved from CSD Archive Library reference number THSD0560. Runstreams, data'and out-put plot files for each transient case are saved in the CSD Ai* chive Library as a data tape. The entry is FSV-STEAMLK/5-84 reference number THSD3856. Retriev-l ) ~ ing a, data tape from the Archive Library follows the procedure given in the CSD i Users Guide GA-A10888. The computer runs for each transient case are listed in Table A1. This table lists the case number, batch run STJOB number, date and time run was created, building, distance from leak, leak cut-off time, source of leak rate, source of heat transfer data, output plot file name anr1 runstream element na=e. The run elements and data elements are all storedIin the runstream file. A change was made to the CONTEMPT-G code to allow plotting data for atmospheric air temperature to be written on an output file for later use with the SUPER

• PLOT program described in Reference 5.

These changes were incorpor-ated in a new version of CONTMT called CONTMT/P. The actual data writing onto the file is done in a new subroutine called PLOTT. A map element CONTMT/PMAP ~ incorporates these changes into absolute element named CONTMT/PABS. Table A2 1 - shows the changes made to CONTMT and the new map element. Table A3 lists l subroutine PLOTT. Runstreams,- input data and plot files described in Table Al as well as the changes to - the CONTEMPT-G code for plotting are all saved in Archive Library

• entry FSV-STEAMLK/5-84 reference number THSD3856." This data tape was written by the runstream shown in Table A4.

Temporary tape number 0086 was submitted

CMB-333 e l RODGERS June 28, 1984 Con-to the Archive Library and copied on a permanently stored archive tape. tents of this data tape entry are listed in Table AS. This table shows the ~ archive data tape file number, catalogued file name written on the tape and a brief description of the file contents. The Archive Librarian will retrieve this data tape from the archives when given a completed Retrieval Request (Form GA-1467) as described in the CSD Users Guide. Computer printouts generated during this analysis are stored in box GA-03130 with the Records Management Department for reference. Table A6 item-izes the contents of this box. e e e / e e 0 e 9 e b l l

e TABLE A1 CONTEMPT-G RUNS DISTANCE LEAK RAT HEAT. DATA [2) TRANSFER (3) START FROM LEAK CUT-OFF ELEMENT NAME( } PLOT FILE NAME

• ASE STJOB DATE TIME BUILDING (FT)

(MIN) 1 2336 May 16 8:59 T 20 4 2 2 FSV-PLOT *TB4M20F-1 TRBN-4 MIN /20FT-1 2 9722 May 8' 16:35 'T 20 4 2 1 FSV-PLOT #TB4H20F-3 TRBN-4 MIN /20FT-3 3 0704 May 8 i6:51 R 20 4 2 i FSV-PLOT #RB4M20F-3 RCTR-4 MIN /20FT-3 4 6676 May 9 15:44 7 20 4 1 h FSV-PLOT 8TB4H20F-2 TRBN-4HIN/20FT-2 5 6737 MAY 9 55:45 R 20 4 1 1 FSV-PLOT #RB4H20F-2 RCTR-4HIN/20FT-2 6 5833 May 15 9:57 R 20 10 2 1 FSV-PLOT #RB10H20F-1 RCTR-10HIN/20FT-1 RCTR I IN/30FT-1 30 '4 2 2 FSV-PLOT 8RB4H30F-1 7 5930 MayIS 9:58 R IM 8 8364 ~May 15 10:39 T 20 10 2 1 FSV-PLOT 8TB10H20F-1 TRBN-10HIN/20FT-1 9 1840 May 23 11:37 T 20 20 2 i FSV-PLOT 8TB20M20F-h TRBN-20 MIN /20FT-k I O 1880 May 23 11:38 R 20 20 2 1 FSV-PLDT8HB20M20F-1 RCTR-20 MIN /20FT-1 lotc3: '.1 ) Building 6 R - Reactor Y T = Turbine

2)

Steam leak rate taken from 1 - GA-A12045 2 - recovered run dated April 1972 ' 3') Heat transfer surface areas, coeffiolents taken fece 1 - GA-A12045 2 - recovered run dated April 1972 4). Element name is stored on catalog file FSV-STMLK85-84 l e i ~ e - ~.m m ws

- l e ~ TABLE A2 CONTMT UPDATES TO CREATE CONTMT/P AND MAP ELEMENT CONN!T/PMAP -~ ~~ ONTMT/P id H 0'G C 3FOR,5 CONTMT/0,CONTMT/P,CONTMT/P .-940 ~ CONTMT/0 ~ -- ~~ -' - --- C S10R E PLOTTING V ALUES C A LL PIO T T-(T S'E'C,'T M P 0701 -1026 CONTMT/0 C W R I T E ' P LO T T I N G ' V A L U E S ~ 0 N ~~U N I T ' 7 ~~-- CALL PLOTT (TSEC,TMPO,1) 3HDG CONTM T/PM AP ~ ~~~- 3 PREP ~3MA'PTIS CONTMT/PMAP,CONTMT/PABS ~ IN CONTMT/P 9 4 e p / e G e 9 9 O l

o , j TABLE A3 ~ ~ SUBROUTINE PLOTT dHDG PLOTT sFOR,$1 PLOTT,PLOTT SUBROUTINE ~PLOTT CTTE'C,TMPUTIRITE) C ~~ C--"- S A V E P LO T T I N G V ALU E S F O R TIME AND ATM0$PHERIC TEMPERATURE O C ~= TIME,~SEC --~~~--'-'-- ~ ~~ ~' ~ C TSEC C TMP0 = ATMOSPHERIC TEMPERATURE, F C I f1'T E = 0, S T O R E T S E C AN D TMPO 04fR'R'XT$ C 1, WRITE ARRAYS ON FILE 7 ~ P A'R A ME T E R NPLT=160 DIMENSION TIMPLT(NPLT), TEMPLT(NPLT), DATEC2), TIMEX(2) ~ D'A T A 'I P / 0 /, INIT/0/, DTPLOT/0.2/, TPLOT/0./ C IP = NUMBER OF POINTS C DT'Prori F R EQ U E N C Y~F O R S T O R I N'G PC0'MITA~~ MIN C TPLOT'* NEXT TIME TO STORE PLOT DATA, MIN C IF(INIT.EQ.1) GO TO 100 C ' INITIALIZE' FILE WITH DATET TIMEi J00NO CALL DATGT1(DATE) CALL G'T'I M ETT I M E X ) CALL PCT (JOSNO,1,0) I NI T =1 WRITE (7.7701) DATE.TIMEX,JOBN0 . TOE CONTINUE c IVffTrrfRTTET1 GO TO (200,300), IWRITE C C SAVE PLOTTING VALUES AT DTPLOT TIME INTERVALS -~70~0 c'ONTINUE'~ TMIN=TSEC/60 I FTfMDC7TPLTT ) Go TO 30u IP=IP+1 IF CI P.GT.NPCT) ~C ALL"ME R R TIMPLT(IP)=TMIN TEMPLT(IP)=TMPO ~~ TPLOT*TPLOT+0TPLOT so to 50u C. WRITE PLOTTING ARRAYS ON F' ICE-7 c 300 CONTINUE W RIT E C 7,7704) ' IP ~-- kRITE(7.7702) (TIMPLT(1),I=1,IP) w a ITE (7 i7703~)7'P WRITE (7,7702) (TEMPLT(I),Isi,IP) WRITE (6,7705) DATE,TIMEX,JOBN0 F00 RETURN C ~ ~--"' 7701' F ORM AT (13 A4) 7702 F ORM AT (6 E12.7) ~ 7703 F ORM AT (16, ATMO S PHERI C TE MP ER ATUR E, D EG-F') 7704 F ORM AT (16," TIME, MINUTES') ~770) FORM AT ('1 PLOTTING" ARR AYS ~ WRITTEN 'ON UNIT 7 FOR ',13A6)- END

-.j f ~ o.. +.*. \\ TABLE A4 RUNSTREAM THAT WROTE TAPE FOR ARCHIVE ENTRY THSD3856 1. PASG,TX 0086,U95,0086U 2. 9HDG FILE 1 IS PSC* FILE 1 COPY,GM TO TAPE 0086 3. IPAT,TJLC PSC* FILE 1. 4. 'tCOPY,GM PSC* FILE 1.,0086. 5. SHDG FILE 2 IS FSV-STMLK*5-84 COPY,GM TO TAPE 0086 6. OPRT,TJLC FSV-STMLK*5-84. 7. ECOPY,GM FSV-STMLK*5-84.,0086. 8. EHDG FILE 3 IS FSV-PLOT *TB4M20F-1 COPY,GM TO TAPE 0086 9. IPRT,TJLC FSV-PLOT *TB4M20F-1, 10. SCOPY,GM FSV-PLOT *TB4M20F-1.,0086. 11. tHD3 FILE 4 IS FSV-PLOT *TB4M20F-3 COPY,GM TO TAPE 0086 12. IPRT,TJLC FSV-PLOT *TB4M20F-3. 13. ECOP.Y,GM FSV-PLOT *TB4M20F-3.,0081. 14. tHDG FILE 5 IS FSV-PLOT *RB4M20F-3 COPY,GM TO TAPE 0086 15. 9PRT,TJLC FSV-PLOT *RB4M20F-3. 16. SCOPY,GM FSV-PLOT *RB4M20F-3.,0086. 17. fHDG FILE 6 IS FSV-PLOT *TB4H20F-2 COPY,GM TO TAPE 0086 18. tPRT.TJLC FSV-PLOT *TB4M20F-2. 19. ECOPY,GM FSV-PLOT *TB4M20F-2.,0086. 20. IMDG FILE 7 IS FSV-PLOT *RB4M20F-2 COPY,GM TO TAPE 0086 21. IPRT TJLC FSV-PLOT *RB4M20F-2. 22. ICOPY,GM FSi-PLOT *RB4M20F-2.,0086. 23. fHDG FILE 8'IS FSV-PLOT *RB10M20F-1 DOPY,GM TO TAPE 0086 24. IPRT,TJLC F9V-PLOT *RB10M20F-1. - -tCOPYt6M' 'FSV-PLOT *RB10H20F-1.,0086. -25. 26. fHDG FILE 9 IS FSV-PLOT *RB4M30F-1 COPY,GM TO TAPE 0086 27. IPRT,TJLC FSV-PLOT *RS4M30F-1. 28. ECDPY,GM FSV-PLOTtRB4M30F-1.,0086. 29. .tHDG FILE 10 IS FSV-PLOT *TB10M20F-1 COPY,GM TO TAPE 0086 30. tPRT,TJLC FSV-PLOT *T310M20F-1. 31. ECOPY,GM FSV-PLOT *TB10M20F-1.,0086. 32. fHDG FILE 11 IS FSV-PLOT *TB20M20F-1 COPY,GM TO TAPE 0086 33. IPRT,TJLC FSV-PLOT *TB20M20F-1. 34. ECOPY,GM FSV-PLOT *TB20M20F-1.,0036. 35. EHDG-FILE'12 IS FSV-PLOTtRB20M20F-1 ' COPY,GM TO TAPE 0084 36. OPRT,TJLC FSV-PLOT *RB20M20F-1. 37. ECOPY,GM FSV-PLOT *RB20M20F-1.,0084. 38. tFREE 0086 2 e b 3 -w a n, e--y -w y.y -m-w--y.y*----.a

~ CMB-33) l , e '.t. RODGERS Juna 28, 1984 TAELE A5 ARCHIVE THSD3856 CONTD1TS TAPE FILE CATAI4G FILE CONTENT DESCRIPTION 1 PSC' FILE 1 CONTEMPT-G updated program, MAP and absolute element 2 FSV-STMLK"5-84 Runstreams and data 3 FSV-PLOT'TB4M20F-1 Plot file for case 1 4 FSV-PLOT *TB4M20F-3 Plot file for case 2 5 FSV-PLOT *RB4M20F-3 Plot file for case 3 6 FSV-PLOT *TB4M20F-2 Plot file for case 4 7 FSV-PLOT'RB4M20F-2 Plot file for case 5 8 FSV-PLOT *RB10M20F-1 Plot file for case 6 9 FSV-PLOT *RB4M30F Plot file f'or case 7 10 FSV-PLOT'TB10M20F-1 Plot file for case 8 s i 1 FSV-PLOTETB20M20F-1 Plot file for case 9 12 FSV-PLOT'RB20M20F-1 Plot. file for case 10 e h a e

w TABLE A6 BOX CA-03130 ~ RECORDS STORAGE MANIFEST Locati n inuo....am Organization Name Org. No. Project No. fo %7 /90 o fsfem Des;-ndP/ ant 2%en,;e s v v Rec rdsStorer Project Name Contract Number Box No. C.J. Rodmers Cort 51. Vrm.in Wsc> S 2-GA - 0 3 I3 o Date of Records " Date Stored (Yr/Mo/ Day) Disposition Date (Yr/Mo/ Day) $l44 /9fV 8 'l / M /.2 9 99/b/fo/ l Recorditescriptions Sucht.cc d.ress' frsAG4-A s2.6 Y5. buildin.. S minufe cuf-oM.204V. blowdow.< ~%rs, recover 2 l 7u~ rbine 6 J' suc% aters & GA-Anzovs. cSor buildi,m. Y Min U$c Cof o$P NR. blowd.aw % recovered en ' 'surfnc.e gree frain, G A -412.o 45 d' e Wurbine bolldino YMinale c.of-off D.0-ff,blowdau)n & 6A-A(10%

• • S Ar*^ #b~ N* *"

i 4'behr boildiw hntiAch co-f-off.ar>ff s blewd.om -fre m G4 - Ar2.od s d' suchce are.rt.s b GA'hA*W kPa eder bo*.(dW. ID ninofe co-{-d-{ a.+@f, b)eu,dso3n.fyd m re.e.coereJ ru ~ ! ' S""0"***rea-5 b FC*r'd C"^ 6' ecker boi{ dine.4 Stidiffe co-{-aN 30 f+ blowdau3n b cee.a ve n.J ru n '.for fa.ce ces -frax G4 - A 12.o45 7'YOditte buildine.ID m n0 fe col-a# ?^N. ble.udewa -bm rue, vere > run ' suchc.e arms From ree-overeA cun Of~bi Ae badd im. 4 n11ack c.utoM. cu 4,bl oms.owx 4re,n me.ove_eca r.; m. suche e ct. ems b GA-Alto 45

• bebiM c bu'ld",,tq,OD nGroke co{,-d[, SLOk blouad a% -from m ove W con

.soAc.= ares.:s h. ced-Aizovs b^dar builohm,on w.nok e.ofaf f ao ft. blewde b recow roa 2 l No$6: Sox & AC-o/H7 c.onkins. retajeg ga.ta. FSV .s+ea.m lea.k .shaces. -froth. previo o S l L

~ CALCULATION REVIEW REPORT I o.,*, e ~itTLE: F,, 7 S t. V r 4 w r-ncfe r b a.< /d co/d r c.AvoT ed Tu khs@ APPROVAL LEVEL beeldewy A*T r4Wf.ste,- 18 4 b.As witk /0 ~sn s,/c /cah QAL LEVEL rev~,_ a,- DISCIPLINE SYSTEM 00C. TYPE PROJECT 00CUf.1LNT NO. ISSU E is01LT R. INDEPENDENT REVIEWER: NAME O' M' bdI ORGANIZATION SdbYY 045/4a [4Y8) 7Eb N-REVIEWER SELECTION APPROVAL: BR MGR a DATE V' ( REVIEW METHOD: YES NO ERROR DETECTED ARITHMETIC CHECK 30 LOGIC CHECK V no ALTERNATE METHOD USED SPOT CHECK PERFORMED 50 COMPUTER PROGRAM USED / REMARKS: @TTACH LIST OF DOCUMENTS USED IN REVIEW) / t CALCULATIONS FOUND TO BE VAll0 AND CONCLUSIONS TO BE CORRECT: INDEPENDENT REVIEWER DATE SIGNATURE ,}}