ML17209A507

Final Rept, Power Operated Relief Valve Failure Reduction Methods, Prepared for C-E Owners Group.Response to NUREG-0737 Encl

ML17209A507
Person / Time
Site:	Saint Lucie
Issue date:	12/31/1980
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML17209A508	List: ML17209A507 ML17266A380
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.02, TASK-TM NUDOCS 8101060378
Download: ML17209A507 (143)

Text

ENCLOSURE 3 PORV FAILURE REDUCTION METHODS FINAL REPORT PREPARED FOR THE C-E OWNERS GROUP NUCLEAR POWER SYSTEMS DIVISION HECEMBER 1980 C-E POWER SYSTEMS COMBUSTION ENGINEERING, INC.

WINDSOR, CONNECTICUT Sl oz 060

E I I(

TABLE OF CONTENTS Section Title ~Pa e BACKGROUND PURPOSE DESCRIPTION OF PORV SYSTEM 1 PORV OPERATING EXPERIENCE 5 PRIMARY SAFETY VALVES 7, METHODS FOR REDUCING PORV SYSTEM FAILURE 8 IMPLEMENTATION OF FAILURE REDUCTION. PROGRAM 12 ANALYSIS AND RESULTS OF FAILURE REDUCTION PROGRAM 13

SUMMARY

AND CONCLUSIONS 15 10 REFERENCES 17 TABLE 1 C-E PRIMARY SAFETY VAL'VE AND'ORV DATA 18 TABLE 2

SUMMARY

OF EVENTS INVOLVING PORV OPERATION 19 TABLE 3

SUMMARY

OF EVENTS RESULTING IN POTENTIAL CHALLENGE TO PORV 20 FIGURE 1 TYPICAL PRIMARY SYSTEM OVERPRESSURE PROTECTION FIGURE 2 TYPICAL ELECTROMATIC RELIEf VALVE 22 Appendix A C-E ANALYSIS OF REFERENCE PLANT (SL-2) FAULT TREE FOR POWER OPERATED RELIEF VALVE LOSS OF COOLANT ACCIDENT

4

1. BACKGROUND The failure of a power-operated relief valve (PORV) to close subsequent to its actuation during an overpressure condition was a key factor in the Three Mile Island-2 (TMI-2) accident. As a result, the operating history of PORVs on all operating light water reactors (LWRs) was investigated by the Nuclear Regulatory Commission (NRC). On an overall basis, the results of the inves-tigation indicated that the probability of a small break loss of coolant accident (LOCA) due to the failure of a PORV to close appeared to be a major contributor to the total probability of a small break LOCA from all causes. (I)

Consequently, the NRC has requested ) that methods for PORV failure reduction be evaluated by C-E for possible implementation to increase plant safety.

2. PURPOSE The purpose of this study is to review PORV failures, to evaluate methods for failure reduction, to describe the plant changes made or recommended to reduce PORV failures, and to evaluate the effectiveness of these changes for C-E operating plants.

3. DESCRIPTION OF PORV SYSTEM

3. I Introducti on A brief description of the provisions for overpressure protection of the typical C-E Nuclear Steam Supply System (NSSS) primary coolant system and clarification of the supporting role of the PORVs is provided below. '.

Overpressure protection for the primary coolant system is based on the combined action of the primary safety valves, secondary safety valves, and the reactor protection system. At operating conditions the PORVs are not formally part of the overpressure protection system; although the presence of PORVs increases the primary coolant system relieving capacity.

3.2 Function of the PORV To reduce the qumber of challenges to the primary safety valves, and thus reduce the probability of gross safety valve leakage or weeping, pressurizers on all C-E operating plants (except for ANO-2) are provided with two PORVs having actuation set points below that of the primary safety valves.

Figure 1 shows a typical installation arrangement for primary system over-pressure protection. Isolation valves are provided upstream of each PORV.

Throughout this report, the term "PORV System" is used whenever the PORV and its isolation valve is being considered in combination. Design and operating parameters for the primary safety valves and PORVs at C-E operating plants are given in Table 1.

Additional functions, not considered in the initial NSSS design, have sin been assigned to the PORVs. These functions include low temperature over-pressure protection, venting, and long term cooling subsequent to a LOCA.

These auxiliary PORV functions have been documented elsewhere and are not included in the scope of this report.

3.3 PORV Desi n Basis The PORVs are designed to have. an opening setpoint pressure below that of the primary safety valves and to provide sufficient relieving capacity to ensure that the pr'imary safety valves do not lift or weep during over-pressurization transient conditions such as uncontrolled rod withdrawal, loss of load, or loss of all non-emergency AC power. The PORY opening setpoint pressure is sufficiently high to ensure that the PORVs do not, open in response to normal maneuvering transients.

3.4 PORV Descri tion All PORVs in operating C-E NSSSs are Dresser electromatic relief valves which are pilot actuated, reverse-seated, and which use pressurizer pressure to operate the valve (Figure 2). Mhen pressurizer pressure exceeds the valve setpoint pressure, the solenoid on the pilot valve is energized; this causes its plunger to actuate a lever to open the pilot valve. The main- valve's pressure chamber above the valve disc is vented

through the open pilot valve and the resulting pressure difference across the main valve disc causes the main valve to open and discharge pressurizer fluid. When pressurizer pressure decreases below the setpoint value, the solenoid is deenergized, the pilot valve closes, and steam pressure builds up in main valve pressure chamber and forces the valve disc closed.

3.5 ~PORV 0 The PORVs are designed for automatic or manual operation. In automatic operation, the PORVs are opened by the high pressurizer pressure trip signal in the reactor protective system, which is actuated by a two out of four channel logic system. The PORVs, which are actuated by the same bistable trip units which actuate the reactor trip, open whenever the pressurizer pressure exceeds the high pressure reactor trip setpoint and they remain open until pressurizer pressure falls below the valve reset pressure. In the manual mode the PORVs can be operated independent of system temperature and pressurizer pressure.

The PORV actuation setpoints vary somewhat from plant to plant, at a nominal value of approximately 2400 psia, about 100 psi below the primary safety valves setpoint and 150 psi above normal operating pressure (Table 1).

3.6 PORV Isolation Valves To permit isolation of a PORV in case of excessive seat leakage or failure to close, mot'or-operated block valves are provided upstream of each PORY.

During power operation the block valves are normally open. However, one or both PORVs may be isolated (block valves closed) because of excessive leakage. Also, operation with one PORV isolated may be considered to avoid excessive reactor coolant discharge due to both. PORVs lifting.

3.7 PORY Leaka e Detection Several methods were used prior to the TNI accident for the detection of excessive PORV leakage or failure to close. These methods include moni-toring PORV discharge piping temperature, PORV'pilot valve position indica-tion, and quench tank pressure, temperature, and level. Readouts from each

of these measurements are generally available in the plant main control room. Subsequent to the THI-2 accident, the NRC required a reliable, direct means for PORV position indication. Action to respond to this requirement is described in Sections 6 and 7 ~

3.8 Electric Power Su lies In performing their function to reduce the frequency of primary safety valve challenges, the PORVs provide equipment protection and as a con-sequence, are not considered as part of the plant safety system. There-fore, the valves as installed in the field were not provided with safety grade power sources and no credit was taken for their operation in safety analyses. Subsequent to the THI-2 accident, consideration was given to providing the PORVs and their isolation valves with emergency power sourc Further actions on PORV system power suppl.ies are discussed in Sections 6 and 7.

3.9 Com arison with Other PWRs The PORV systems provided in pressurized water reactors (PWRs}- supplied by Babcock and Wilcox (BQI) ), We'stinghouse (W)

(4) and C-E differ in details such as the type, number, capacity, setpoint, valve vendors and control circuitry. Certain important differences among the PWR vendors'ystems are described in the following sections.

On C-E plants, the initial design function of the PORVs was solely to reduce the challenges to the primary safety valves during power operation.

The'ORVs on BSW and 'W plants had an additional function, namely, to reduce the frequency of reactor trips due to high pressure. The PORU actuation set point on C-E plants coincides with the high pressure reactor trip setpoint, whereas, the other PWR vendors required that the PORV actuation pressure be below the high pressure reactor trip setpoint in order to reduce the number of high pressure trips. The C-E design allows the specification of a higher PORV actuation pressure., and therefore a greater margin above the normal plant operating pressure than do the othe PWR designs. Typically, the margin between normal operating pressure and

the PORY actuation setpoint was about 150 psi for C-E plants, 100 psi for W plants, and 70 psi for B8W plants. This difference provided an incremental margin to PORV challenges in C-E plants compared with those of the other PWR vendors.

The BSW plants are equipped with the same type 'of PORVs as those of C-E, namely, the Oresser electromatic solenoid pilot-operated valve described in Section 3.4. The majority of W plants use Copes-Vulcan spring-loaded, air-operated valves, Air pressure on the control diaphragm overcomes the spring force to open the valve. Venting the air pressure from the control diaphragm allows spring force to close the valve. A few W plants use PORVs manufactured by Masoneilan (3 plants), Oresser (1 plant), ACF Industries (1 plant), and Control Components (1 plant).

4. PORV OPERATING EXPERIENCE

4. 1 Combustion En ineerin Plants The operating experience of PORVs in C-E plants has been compiled in Table 2 based on information supplied by the various plant operators during a survey conducted in early 1980. The PORV actuations noted in Table 2 do not necessarily represent the total number which have occurred, since PORV actuations were not reportable events and were not routinely recorded. Therefore, some actuations may have been overlooked. Also, since the available means for the detection of PORV actuation was not direct,, but generally dependent upon an integrating effect, such as increasing quench tank level, for example, some actuations may have gone undetected.

Table 3 is a tabulation of high pressurizer pressure reactor trips occurring in C-E operating plants for which PORV actuations were not reported. The data was obtained from a review of published data, mainly from the NRC. Since, by design, a high pressurizer pressure reactor trip should be accompanied by PORV actuation, it is inferred that the actuation did occur, though it was not reoorted.

Table 2 indicates a total of seven confirmed PORY actuation events. Four events occurred during PORV testing or system maintenance. In two of these events the PORVs failed to close satisfactorily. The remaining three actuation events occurred during power operation, with the PORVs operating satisfactorily in each case. Table 3 indicates a total of sixteen high pressurizer pressure reactor trips, eleven of which resulted from turbine runbacks. Tables 2 and 3 extend the PORV actuation data presented in NUREG 0635(1)

It was inferred that the high pressurizer pressure trips listed in Table 3 were accompanied by PORV actuations. Combining the confirmed PORV actuation events during power operation listed in Table 2 with the inferr actuation events from Table 3, a total of nineteen events or thirty-eight PORV challenges is obtained, with no failures being reported. A total of about 29 reactor-years of operation is covered by this data.

The two PORV failures-to-close on C-E plants listed in Table 2 occurred during maintenance or testing.

The Palisades incident occurred when the Reactor Protection System (RPS) was deenergized for maintenance, which caused the PORVs to open. Oue to an ambiguity in the pertinent wiring diagrams the technician failed to perceive that his action would cause PORY actuation. The spring-return-to-Auto feature of the PORV selector switch contributed to the incident since the selector switch could not be retained in the "Manual" mode and "Shut" position unless held there by the operator. Corrective action was taken to clarify the pertinent wiring drawings and eliminate the spring-return-to-Auto feature of the PORV selector switch. The PORV failure-to-close in this instance was not due to the failure of the valve.

The second PORV failure-to-close occurred in Calvert Cliffs 81 during valve operational testing following valve maintenance, The valve failed to shut completely. Modified replacement parts had been installed in the

valve because original, replacement parts were. unavailahle due,to vendor upgrading 'of the valve design. Following adjustment of the pilot valve stroke, satsifactory valve closure was obtained.

4.2 Ex erience at Other PWRs Westinghouse PWRs in the U.S. have not reported any PORV failures but since they are equipped with a different type of PORV their reliability experience'isnot relevant to C-E PORVs.

It has been estimated that in B8W plants there have been approximately 150 actuations of PORVs with six cases of failure-to-close properly.

One failure oc'curred during low power testing upon loss of a vital bus, another during star tup testing due to improper venting, and a third was

.a leaky valve. Three failures occurred during power operation, giving approximately 3/150 = .02 failures per demand.

5. PRIMARY SAFETY VALVES 5.1 0 eratin Ex erience No primary safety valve liftshave been reported for C-E operating plants during approximately 30 reactor-years of operation. Westinghouse plants also have not reported any primary safety valve lifts. One primary safety valve lift has been noted in a B8W plant, but no details were given.

In view of the. lack of challenges to the primary safety valves, a direct quantitative estimate of their reliability based on experience cannot be made.

5.2 Probabilistic Anal sis The main steam safety valves (MSSV) are much more subject to challenges than are the primary safety valves, so that data regarding their reliability has been developed. This data does not have direct applicability to the primary safety valves since, even though the MSSV bears some similarity to the primary safeties, there are distinct differences with respect to service conditions, materials, and other design features. Lacking data on the primary safety valves, the MSSV data may provide. some indication of'rimary safety valve reliability.

'A study of PWR HSSV operating experience up to May, 1978 was performed by C-E. The data sources used were iHPRDS Failure Report Summaries, License Event Report Summaries, and Operating Units Status Reports.

The period reviewed included 137 reactor-years of operation at 38 PWR "plants with an estimated population of 570 HSSVs. During this period there were an estimated 2070 MSSV test demands (pre-operational and annual).

V Assuming one demand on MSSVs for every ten scrams or turbine trips, about

'580 operational NSSV demands were estimated. The total number of HSSY demands in the study period were estimated'o be 5650.

During this period two events were reported (none from C-E operating plant in which MSSVs failed to close following a'emand. The first event occurred at Turkey Point Unit 4 in 1974 when a missing cotter pin caused one MSSV to fail open. The second event occurred at Three Nile Island Unit 2 in April, 1978. A common mode failure of six MSSVs to close occurred due to cocked sleeves in the bellows assembly. Thus, the total number of NSSV failures to reseat reported during the study period was seven.

Based on the seven reported HSSV failures and the 5650 estimated HSSV demands, a failure rate of 1.24 x 10 per demand is estimated, This

-2 failure rate is lower than the value of 2 x 10. estimated for power operated relief valves in NUREG 0560. Assuming that the HSSY reliability data are to some degree applicable to the primary safety valves, the data suggests that the primary safety valves may be more reliable'han the PORVs.

More definite conclusions must await development of operational and/or test data on primary safety valves.

6. METHODS FOR REDUCING PORV SYSTEM FAIL'URE 6.1 Reduction of PORY'Challen es The frequency of PORV system failures can be reduced by decreasing the fr quency of challenges to the .PORVs. These reductions must be made without adversely impacting safety or incurring unacceptable economic or performance

penalties. Methods for potentially decreasing the frequency of PORV challenges on C-E plants and a brief summary of their impacts on the plant are provided below.

6.1.1 Raise PORV Set oint High pressurizer pressure trips the reactor when the pressure exceeds the trip setpoint pressure and the output from the same bistable comparator also actuates the PORV. Therefore, only one setpoint is available. Raising this Reactor Protection System (RPS) high pressurizer pressure reactor trip setpoint would invalidate the safety analysis and increase the challenges to the primary safety valves.

6.1.2 Lower Hi h Pressurizer Pressure Tri Set oint This requires the concomitant lowering of the PORY actuation setpoint as described above. Doing so would increase the number of challenges to the PORVs.

6. 1.3 Raise the set oint for the existin PORV 0 enin /Hi h Pressurizer Pressure Tri and Add Another Hi h Pressurizer Pressure Reactor Tri at 2400 si The setpoint for the existing PORY Opening/High Pressurizer Pressure Reactor Trip would need to be raised approximately no higher than 20-40 psi to prevent primary safety valve challenges during a full loss of turbine load without a simultaneous reactor trip while simultaneously precluding PORV openings'uring milder pressure increases. The benefits of this alternative would be very small since only a very small fraction of the PORY openings would have been avoided by this modification (i.e., full load rejection where PORV opening was desired to precl.ude primary safety valve opening and the inadvertent initiations would not have been affected).

Further, there is no more room in the protective system cabinetry in some of the operating plants to accommodate additional bistable trip units and other circuitry that would be required. Adding additional trips would be expensive and would take a considerable amount of time to incorporate.

6. 1.4 Block Out and/or Deactivate PORV Durin Power 0 eration In the event of a full power incident which causes the turbine admission valves to close rapidly (e.g. full load rejection, electrical system over-frequency, turbine control failure), the reactor would trip on high pressurizer pressure in the absen'ce of a turbine trip signal. The pressurizer pressure would continue rising above the 2400 psi setpoint until the reactor trip quenched the power output of the core and caused the pressurizer pressure to decrease. It is prudent to use the power operated relief valves to preclude challenging the primary safety valves during this transient. There are PORV block valves which can be closed in the unlikely event of a PORV failing to close. Such block valves are unavailable to mitigate the cons'equences in the unlikely event that a safety valve fails to reclose.

6. 1.5 Reduce 0 eratin Pressure A reduction in operating pressure would tend to reduce the number of PORV openings, but by only a small proportion. Also, the lower the operating pressure, the higher the overshoot in pressure after a load rejection is terminated by the high pressurizer pressure trip. The higher overshoot in pressure results from the delay in the reactor trip. This increases the potential for challenging the primary safety valves. More importantly, decreasing the primary operating pressure would decrease the operating ONB ratio thus causing the core to be operated closer to one of the safety limits.

6.1.6 Elimination of Turbine Runback Table 3 indicates that a relatively large number (11[ of high pressure trips (and presumably 22 PORV actuationsj occurred during turbine runback events. A review of this plant feature indicated that its elimination would not adversely affect plant operation, while at the same time reducing PORV challenges to a significant degree.

6.2 Im roved Ca abilit for Countermeasures The frequency of PORV system failures can also be reduced by improving the capability for appropriate countermeasures (PORV isolation) sub-sequent to a PORV failure to close. Methods for potentially improving the capability to take appropriate action and a brief summary of their impacts on the plant are discussed.

6.2. 1 Automaticall Close Block Valve Whenever PORV Fails to Close on Command There are several ways this could be implemented. The block valve closing signal could be armed by an initial PORV opening signal so that the block valve would remain open in normal operation but would be automatically closed if the PORV failed to close on command.

Another approach would use the concurrence of an open PORV valve and and PORV valve closure command to automatically close the block valve.

Although automatic valve closure would remove the requirements for operator action upon PORV failure, the additional control circuitry would introduce additional complexity to the system and would itself be subject to its own failure modes. These schemes require further detailed evaluation to determine their positive and negative impacts on oyera11 plant safety. A simpler approach is to assure that the operator is able to utilize existing inplant instrumentation to identify a stuck-open PORV and to close the block valve..

6.2.2 PORV Position Indication Reliable and positive control room indication of PORY position would provide vital information to the operator in a clear and timely manner to permit him to take the appropriate action necessary to prevent escalation of a minor incident into a LOCA. An ultrasonic flow-meter, located at the discharge piping of the PORV, w'th flow indication and alarm in the control room, would provide direct, positive, rapid-response, and reliable indication of PORV position.

An advantage of this instrument is that it does not require any penetration of the piping. Alternatively, the PORV could be provided with a position indicator for the main valve disc position.

6.2.3 Electric Power Su lies The PORVs and their associated block valves, which. were designed for an equipment protective function rather than a safety function, were not initially provided with emergency power supplies. The provision of emergency power to these valves would maintain the availability of the relief system and also permit its isolation, if necessary, upon loss of all non- emergency power sources.

6.2.4 Im rovement of 0 erator Ca abilit The evaluation of the TMI-2 incident indicated that a program to improve operator performance, particularly during emergency conditions, would significantly reduce the potential for serious nuclear incidents.

Upgrading operator capability to recognize and to respond appropriately to a PORV failure-to-close should significantly reduce the possibility of the subsequent occurrence of a small break LOCAL

7. IMPLEMENTATION OF PORV SYSTEM FAILURE REDUCTION PROGRAM The following actions to reduce PORV system failures have been completed or are pending:

The turbine runback feature has been eliminated from C-E operating plants.

2. The motor operators for the PORV block valves and the. pilot solenoids for the PORVs have been provided with. emergency power supplies to permit them

-to function upon the loss of all non-emergency power.

3. Ultrasonic flowmeters are being installed on the PORV discharge piping to provide a direct measurement of steam flow and therefore, of PORV position, with indication and alarm in the control room.

4. Operator training programs have been initiated to provide the operator with a more comprehensive understanding of plant operation under emergency conditions. Guidelines and detailed emergency operating procedures have been developed to aid the operator to cope with a spectrum of emergency conditions. This includes the conditioning of the operator to recognize and respond promptly to PORV failure to prevent escalation of the failure to a small break LOCA.

8. ANALYSIS AND RESULTS OF FAILURE REOUCTION PROGRAM An analysis was performed to provide an estimate of the reliability of the PORV system as well as an estimate of the improvement in reliability expected as a result of the various actions taken or to be taken as noted in Section 7.

Appendix A presents a description of the reliability analysis and the results obtained. Thi's section provides a discussion of the analysis and results.

Table A-I gives challenge frequencies for the PORVs and demand failure rates used in the analysis for various aspects of PORV and block valve operation.

The frequency of challenges to the PORVs is based on the C-E operating presented in Section (4. 1). The PORV demand failure (failure-to-plants'xperience close) rate is based on the B8W operating experience described in Section 4.2.

The reasons for using the BSH data as a basis are that:

1. The C-E PORV system design basis and other NSSS features as discussed in Section 3.0 tended to keep PORV actuations to a minimum, so that only a small statistical data base for PORV actuations on the C-E:NSSS was available.

2. BSW operating plants had experienced a relatively large number of PORV actuations, and in addition, their operating plants are equipped, with one exception, with the same type of PORVs from the same supplier as are C-E operating plants.

i

3. Westinghouse operating plant experience was not included due to the fact that, in general, they used a different type of PORV from different vendors than did C-E and 8'.

The specific value of the 88W PORV demand failure rate used in the Appendix A analysis was 0.02 failures-to-close per opening. If the C-E plant experience (38 challenges with zero failures} was statistically combined with the 88W data, the demand failure rate would be reduced by about 20/ to 0.016, A value of 0. 155 was used for the probability of failure of the operator to isolate the failed-open PORV. This value is based on data in WASH 1400 (5) and is taken as the mean between the operator's normal stress level and severe stress level failure probabilities.

Table A-2 provides the estimated frequency of an unisolated failed-open PORV, (i.e. small break LOCA due to a failed-open PORV) for a C-E plant to which various features have been incorporated. It shows the progressive reduction in the recurrence frequency of a small break LOCA due to a failed-open PORV as the various methods for PORV system failure reduction noted in Section 7 are implemented. Case 1 is the reference case prior to elimination of the turbine runback feature. This case takes no credit for operator action to isolate the failed-open PORV on the assumption that the available instrumen-tation did not provide clear, positive valve position, indication to the oper-ator. Case 2 assumes elimination of the turbine runback feature, with no credit for operator action. Case 3 is similar to Case 2, except credit is taken for operator action on the basis that appropriate instrumentation has been added to give the operator clear, positive indication of PORV position.

Cases 4 and 5 assume that provision for automatic closure of the block valve upon failure of the PORV to reclose has been incorporated. Case 4 assumes a control grade design which involves reliable components but has only a single isolation valve and hence is not single failure proof. Case 5:assumes a safety grade design with series isolation valves to provide single failure protection for closure.

The estimates in Table A-2 show that the elimination of the turbine runback feature and taking credit for operator action (based on positive valve position indication and alarms) serves to reduce the estimated recurrence frequency of a small break LOCA due to PORV failure by a factor of about 14.5 (or about 18 for a PORV demand failure rate of .016). The estimated recurrence frequency for a small break LOCA due to a pORV failure is 1.8 x 10 per reactor-year

-3 (or about 1.4 x 10 per reactor-year for a PORV failure demand rate of .016),

which is well within the 905 confidence range of a small break LOCA due to a

-2 -4 pipe break, 10 to 10 per reactor-year, as estimated by MASH-1400. Two factors which would further reduce the recurrence frequency of a small break LOCA due to PORV failure from the value before the TMI-2 accident have not been quantified. One is the improvement in operator capability and reduction in the probability of operator error due to new intensive operator training programs, and the updating of plant emergency procedures based on guidelines which consider the realistic response of the plant to transients and accidents.

The second is the provision of emergency power to the PORV block valves to allow PORV isolation, if necessary, after loss of non-emergency power. These factors provide some additional confidence regarding the conservatism of the analytical results.

Table A-2 also shows that provision of control grade automatic block valve closure upon PORV failure to close would reduce the recurrence frequency of a small break LOCA due to PORV failure nearly to the lower limit of the range

-2 of 10 4 per reactor-year estimated for the small break LOCA due to pipe rupture by MASH-1400. The provision of a safety,-grade, single-failure-proof design for automatic block valve closure by the addition of redundant isolation valves reduces the recurrence frequency to a negligable value.

9.

SUMMARY

AND CONCLUSIONS The C-E operating plants after approximately 29.reactor-years of operation have experienced no PORV failures during power operation. The elimination of the turbine runback feature and the provision of a direct reliable means for indica-ting PORV position to the operator provided significant improvements in system reliability. The recurrence frequency of a small break LOCA due to PORV failure

has been reduced by an estimated factor of about 15 to a value. of about

-3 1.8 x 10 per reactor-year. This recurrence frequency is well within the 90% confidence range of the recurrence frequencies of 10 to 10 per reactor-year for a LOCA due to a small pipe rupture estimated in WASH-1400. Improved operator training programs and emergency procedures, as well as the provision of emergency, power to the PORVs and to their block valves, though not quanti-fied, has reduced the small break LOCA recurrence frequency even further.

The incorporation of the feature of automatic block valve closure upon PORV failure would further increase PORV system reliability.

10. REFERENCES

1. NUREG-0635 - Generic Evaluation of Feedwater Transients and Small-Brea@-

Loss-of-Coolant Accidents in C-E Designated Operating Plants, January 1980.

2. NUREG-0737 Clarification of THI Action Plan Requi.rements, Nov. l980

3. NUREG-0560 - Generic Assessment of.Feedwater Transients in Pressurized Water Reactor Designed by the Babcock and Wilcox Company, May, 1979.

4. NUREG-0611 - Generic Evaluation of Feedwater Transients and Small-Break-Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants, January, 1980.

5. MASH-1400 - Reactor Safety Study, October, 1978 Appendix III, Table III 6-1.

TABLE 1 C-E PQIlgRY 5AFET'('tALItE AND P ATA A. PRIMARY SAFETY VALVES DATA Plant Valve Vendor Val ve

~T)e Number

~er lant Setpoint

~Sl

• Rated Minimum 11/h ~i.h

• Maximum Actual

/h Ft. Calhoun Crosby HB-BP-86 2530 216,000 240,000 2485 212,000 236.000 Pal i sades Dresser 31739A 2565 230,000 256,000 2525 230,000 256,000 2485 230,000 256,000

• St. Lucie 1 Crosby HB-BP-86 2485 212,000 236,000 Maine Yankee Dresser 31709KA 2535 218,000 243,000 2510 216,000 240,000 2485 214,000 238,000 Calvert Cliffs 1 and 2 Dresser 31739A 2550 304,000 334,000 2485 296,000 329,000 Millstone 2 Dresser 31739A 2 2485 296,000 329,000

• Capacity indicated corresponds to 3X accumulation above set pressure B. PORV DATA Valve 1 1 Num er Setpoint *Relievinq Capacity Plant Vendor ape ~er lant ~si lb/hr Ft. Calhoun Dresser 31533 VX 2 2385 111,000 Palisades Dresser 31533VX 2385 155,000 St. Lucie 1 Dresser 31533VX-30 2385 159,000 Maine Yankee Dresser 31533VX 2385 150,000 Calvert Cliffs 1 and 2 Dresser 31533VX-30 2385 159,000 Millstone 2 Dresser 31533VX-30 2400 148,000

• Rated value at OA accumulation, provided by vendor

TABLE 2 Summar of Events Involvin PORV 0 eration PLANT INITIATING PLANT DATE COND IT IONS EVENT DESCRIPTION Consumers Power*

Palisades Sept. 8, 1971 Mode 3 Technician deenergized PORV opened when RPS deenergized.

RPS for maintenance Baltimore Gas 5 Elec.

Calvert Cliffs-1 July 6, 1979 Mode 5 Test of PORV During operational test of PORV valve failed to fully close.

Adjusted pilot valve stroke 2 -

August 20, 1980 100K MSIV Closure PORVs cycled on high pressure Florida Power 5 Light Feb. 21, 1977 100K 100'A load rejection PORV cycled during test when St. Lucie -1 reactor tripped on high pressure.

Omaha Public Power Dist.

Fort Calhoun Hay 28, 1978 80Ã Turbine control valve PORV's cycled when plant tripped closed on high pressure.

Fort Calhoun Dec..20, 1978 Mode 5 Troubleshooting PORV's opened when technician pressure recorder pulled recorder fuses.

Northeast Utilities Aug. 10, 1979'ode 5 Troubleshooting PORV opened on loss of AC Millstone-2 to emergency bus.

Maine Yankee Atomic Power Company Ho PORV Operation Events Maine Yankee

• Palisades has operated since 1972 with PORV block valve shut.

TABLE 3 Summary of Events Resulting In Potential Challenge to PORV PLANT INITIATING PLNlT DATE CONDITIONS EVENT DESCRIPTION Consumers Power Mar. 19, 1973 85K Circuit Noise Spurious high pressure trip Palisades (Note 1) Aug. 31, 1976 100Ã MSIV shutting High pressure trip due to MSIV shutting.

Nov. 26, 1976 15Ã Generator Synchronization Spurious high pressure trip while bringing generator on line.

Hay 22, 1978 100% Closure of both t1SIV Hiqh pressure reactor trip.

Baltimore Gas 8 Elec.

Calvert Cliffs -1 July 8, 1975 lOOX Turbine runback High pressure trip due to turbine runback. Unable to verify PORV operation due to loss of plant computer.

Jan. 26, 1975 20$ Power reduction with High pressure reactor trip.

manual pressurizer spray control Northeast Utilities Apr. 13, 1976 805 Turbine runback High pressure reactor trip.

Millstone -2 Apr.. 23, 1976 ~

100K Turbine runback High pressure reactor trip.

tiay 10, 1976 100K Turbine runback High pressure reactor trip.

Hay 24, 1976 100K Turbine runback High pressure reactor trip.

Hay 25, 1976 100Ã Turbine runback High pressure reactor trip.

'June 8, 1976 100Ã Turbine runback High pressure reactor trip.

June 10, 1976 100Ã Turbine runback High pressure reactor trip.

June 19, 1976 lOOX Turbine runback High pressure reactor trip.

June 21, 1976 100K Turbine runback High pressure reactor trip.

Aug. 13, 1976 lOOX Turbine runback High pressure reactor trip.

Note 1 - Palisades has operated since 1972 with PORV blocking valve shut.

SAFETY VALVE TO QUECICH PORV TANK SAFETY TO. VALVE QUENCH BLOCK

. T,Ai)K' VLAVE S

PORV BLOCK VALVE PRESSURIZER FIGURE 1 TYPICAL PRII;ARY SYSTEt<

OVERPRESSU iE PROTECT IO)3

44

)IA 2) 22 20 ) IA 2I 204 )S" + I/O 25 (I)5.I I.4 ON)

OIINANIIINC 15

)I"I((Ill.O I I CN

">J.

OO)

Ill.

CINIIO Of Cllllll Il~

74A I 4C i~ r ~

5 I

(177.l NN IC 52A 42 I 44 I

IA IO IC 4I 12

( IA I <II 24-)/4 (III.I NO)

FIGURE 2 TYPICAL ELECTROMATIC RELIEF VALVE

REF.

QTY. NOMENCLATURE NO.

MAIN BASE-PILOT, BASE ASSEM.

(WELDED, INTEGRAL ASSEM.)

1A 1 INLET, FLANGE 1B 1 OUTLET FLANGE 1C 1 CAGE 1D 1 TUBE INSERT 1E 8 MAIN BASE INLET STUD 1F 1 PILOT BASE 1G 4 PILOT BASE STUD 2 8 INLET STUD NUT 3 1 MAIN DISC 3A 1 PISTON RING 4 1 MAIN DISC SPRING 5 1 GUIDE 6 1 GUIDE GASKET 7 1 GUIDE RETAINER PLUG 8 1 RETAINER PLUG CAP SCREW 8A 1 CAP SCREW LOCKWASHER 8B 1 LOCK SCREW 8C 1 LOCK SCREW LOCKWASHER 9 1 SEAL WIRE 10 1 PILOT DISC 11 1 PILOT DISC SPRING 12 1 SEAT BUSHING 12A 1 LOWER GASKET 12B UPPER GASKET 13 LOWER SPINDLE 14 BELLOWS ASSEM.

(WELDED, INTEGRAL ASSEM.)

14A BELLOWS 14B FLANGE 14C PISTON 15 UPPER SPINDLE 16 PILOT STUD NUT 17 SOLENOID BRACKET 18 LEVER 19 LEVER PIN ASSEM.

19A SHOULDER SCREW 19B NUT FIGURE 2 - TYPICAL ELECTROMATIC RELIEF VALVE

. Sheet 2 of 3

REF.

NO. QTY. NOMENCLATURE 19C 1 BRACKET BUSHING 19D 2 LEVER BUSHING 19E 1 COTTER PIN 20 1 ADJUSTING SCREW 20A 1 LOCKNUT 21 1 BRACKET PLATE 22 4 BRACKET PLATE CAP SCREW 22A 4 LOCKWASHER 23 1 SOLENOID 24 4 SOLENOID CAP SCREW 24A LOG KWASHER 25 1 PLUNGER HEAP 26 1 LEFT HAND SPRING GUIDE 27 1 RIGHT HAND SPRING GUIDE 28 2 PLUNGER SPRING 29 2 PLAIN SPRING WASHER 30 2 SPRING COTTER PIN 31 2 GUIDE BRACKET 32 1 GUIDE BRACKET BOLT 32A 1 LOCKWAS HER 32B 1 NUT 33 1 SWITCH 34 2 SWITCH MACHINE SCREW 34A 2 LOCKWAS HER 35 3 SPRING GUIDE CAP SCREW 36 1 SPECIAL SPRING GUIDE SCREW 37 4 SPRING GUIDE NUT 37A 4 LOCKWAS HER 38 1 BRACKET COVER ASSEM.

38A 1 LEFT HAND COVER 38B 1 RIGHT HAND COVER 38C 5 MACHINE SCREW 38D 5 LOG KWASHER 38E 5 NUT 39 1 SOLENOID COVER 39A 6 MACHINE SCREW 40 1 NAMEPLATE 41 1 TAG PLATE 42 1 CAUTION PLATE 43 1 SOLENOID NAMEPLATE 44 10 NAMEPLATE SCREW FIGURE 2 - TYPICAL ELECTRONATIC RELIEF "VALVE Sheet 3 of 3

~1 APPENDIX A C-E ANALYSIS OF REFERENCE PLANT (SL2) FAULT TREE FOR POWER OPERATED RELIEF VALVE LOSS OF COOLANT ACCIDENT

TABLE OF CONTENTS Section Ti tie ~Pa e 1.0 PURPOSE A-1 2.0 SCOPE A-1 3.0 SAFETY FUNCTION ELEMENT DESCRIPTION A-1 4.0 ANALYSIS ASSUMPTIONS 5-2 5.0 RESULTS A-3

6.0 REFERENCES

A-4 Tab 1 es Ti tl e ~Pa e A- 1 COMPONENT AVAILAB'ILITYDATA FOR PORV A-5 LOSS OF COOLANT ACCIDENT A-2 RECURRENCE FREQUENCIES FOR PORV LOSS OF COOLANT INCIDENT A-7

~Fi ures Title A-'1 POWER OPERATED RELIEF VALVES SCHEMATIC A-8 A-2 FAULT TREE LOGIC DIAGRAM FOR PORV LOSS OF COOLANT INCIDENT A-9

1.0 PURPOSE This report presents the results of a reliability analysis for loss of reactor coolant through the power operated relief valves.

2.0 SCOPE The reliability analysis considers the performance of the safety function element (SFE) strictly as defined in Sections 3 and 4, Safety Function Element Description and Analysis Assumptions. In this form, the analysis will not be applicable to all initiating events but presents a model which was determined to be most useful in terms of applicability and most amenable to later modification for application to special cases.

3.0 SAFETY FUNCTION ELEMENT DESCRIPTION The safety function element, Relieving Reactor Coolant System Pressure through the Powered Operated Relief Valves (PORV), refers to the opening of I

the PORV due to high Reactor Coolant System pressure and reclosing these val ves once the Reactor Coolant System pressure decreases bel ow the val ve setpoint. Included in this SFE 'are the 'opening and reclosing of the PORVs. Also included is the operator's capability to close the PORV block valve, from the contr'ol room, if the PORV fails to reclose.

A schematic of the PORV layout is shown in Figure A-3,. There are two 50/

flow capacity PORVs. Both PORVs receive a signal which causes them to open during a high Reactor Coolant System pressure transient. Once the Reactor Coolant System pressure decreases below the PORV setpoi nt, the PORVs reclose to preclude excessive loss of Reactor Coolant System inventory.

However, if either'r'both PORVs do not reclose the operator has the capability of terminating flow through the valve(s) by closing the block va 1 ve (s).

4.0 ANALYSIS ASSUMPTIONS The following assumptions were made in performing the reliability analysis:

1. PORV loss of coolant incident is defined as the inability to terminate flow through both PORVs to preclude excessive loss of Reactor Coolant System inventory.

2. At the actuation of the PORVs, the operator's normal stress level changes to a level intermediate between normal and severe stress (average of normal and severe stress levels).

3. Both PORVs have identical setpoint.

4. Fai led components are not repaired during this SFE.

5. High pressurizer pressure condition exists at the actuation of the PORVs.

6. The reactor is at power prior to actuation of the actuation of the PORVs.

7. The component availability data for PORV loss of coolant incident which was used is given in Table A-j..

A-2

5.0 RESULTS The fault tree logic diagram for power operated relief valve (PORV) loss of coolant incident is shown in Figure A-2. The minimal cutsets consist of at least three components. Therefore, all three component events must occur in order for a PORV loss of coolant incident to occur.

Best estimate recurrence frequencies for the PORV loss of coolant .incident were calculated for the following cases:

1. Turbine runback feature and no operator action

2. Without turbine runback feature and no operator action

3. Without turbine runback feature and with operator action

4. Without turbine runback feature and with automatic closure of block valve

5. Without turbine runback feature and with automatic closure of series redundant block valves The results are shown in Table A-.2. Cases 4 and 5 assumed potential improvements to the current plant design.

A-3

6.0 LIST OF REFEREHCES Fault Tree

Title:

PORY LOSS OF COOLANT INCIDEHT Ref. No. Descri tion User's Manual and Output Guide for C-E Reliability Evaluation Code (CEREC), Rev. 1, W.S. Chow.

2. Combustion Engineering Interim Data Base - Failure Rates for Nuclear Power Plant Components, D.J. Finni curn.

3 ~ IEEE STD500-1977, IEEE Guide to the Collection and Presenta-tion of Electrical, Electronic, and Sensing Component Reliability for Nuclear Power Generating Stations.

4, WASH 1400 (HUREG-75/014) Reactor Safety Study, An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, Appendices III and IY, (Tables III-2-1 and III-6-1).

5. Combustion Engineering Reliability Data System, Initiating Event Report (1-1-61 to 12-31-77), R.G. Sider.

6. NPRDS 1977 Annual Reports of Cumulative ystem and Component Reliability, September, 1978.

7. St. Luci e I I SAR, Sect i on(s ) 5.5.12

8. Post-TMI Evaluation Task 3 Follow-up Peport, Pressurizer Systems and Emergency Power Supplies, Combustion Engineering, November, 1980.

9 ~ NUREG-0560, Staff Report on the Generic Assessment of Feedwater Transients in PWRs Designed by Babcock & Wilcox Company, U.S. NRC, May; 1979.

Drawings St. Lucie II, Sequence of Events Auxiliary Diagrams St. Lucie II, Reactor Coolant System P&I Diagram, E-13172-310-109, Rev. 03 A-4

TABLE A-COMPONENT AVAILABILITYDATA FOR PORV LOSS OF COOLANT INCIDENT Component Descr iption Code Frequency Ref. Demand Ref.

Identification (1/yr.) Failure Rate Power Operated Opens on Demand PORV100D 6.60E-Ol Relief Valve (With Turbine Runback) PORV200D 6.60E-01 Opens on Demand PORV100D . 2.78E-01 (Without Turbine Runback) PGRV200D 2.78E-01 Opens Spuriously PORV10S 2.80E-03 PORV20S 2.80E-03 Fails to Reclose PORV1FTR 2.00E-02 PORV2FTR 2.00E-02 Block Valve IA Mech. Malf. BV IAMM 6.59E-05 Valve Motor Fails BY IANT 2.02E-04 Valve Breaker Fails BV IABR 1.00E-06 to close Automatic Signal BVIAAS 1.20E-02 not Received Operator Fails BV14030P 1. 55E-Ol to Close Valve Block Valve IIA Mech. Malf. BVIIAMM 6.59E-05 Valve Motor Fails BVI IAMT 2.02E-04 Valve Breaker Fails BVI IABR 1.00E-06 to close Automatic Signal BVI IAAS 1.20E-02 not Received Operator Fails BV14050P 1.55E-01 to Close Valve

TABLE A-1 (continued)

COMPONENT AVAILABILITYDATA FOR PORV LOSS OF COOLANT INCIDENT Component Description Code Frequency Ref. Demand Ref.

Identification (1/yr ) Failure Rate Block Valve IB t1ech. t1al f. BV I BMH 6.59E-05 Valve Motor Fails BV I Bt1T 2.02E-04 Valve Breaker BVIBBR 1.00E-06 Fails to Close Automatic Signal BVIBAS 1.20E-02 not Received Block Valve IIB Mech. Half. BV I I BMM 6.59E-05 Valve tiotor Fails BVI IBHT 2.02E-04 Valve Breaker BVIIBBR 1.00E-06 Fails to Close Automatic Signal BVI IBAS 1.20E-02 not Received Best Estimate Using 246.2 Possible Reactor Years

• • Values Mere Obtained from Data in Ref. 4

Table A-2 Recur rence Frequencies for PORY Loss of Coolant Incident FREQUENCY DESCRIPTION {1/YR.)

Turbine runback feature and no 2.6E-02 operator action Without turbine runback feature and 1. 1E-02 no operator action Without turbine runback feature and 1'.8E-03 with operator action Without turbine runback feature and with 1. 4E-04 automatic closure of block val ve Without turbine runback feature and with 1. 7E-06 automatic closure of series redundant block valves A-7

V-1402 PORV TO QUENCH V-1403 TQi<K BLOCK , !

V.%LVE V-1404 PORV V-1405 BLOCK VALVE PRESSURIZER FIGURE A-1 POHER OPERATED RELIEF VALVE SCHEi'1AT I C

Goer oue 1b P6KV PAIN LGFT OPSY PATH Z, g Cmwl t4S OPSY R

(Spic T 3)

Figure A-2 Fault Tree Logic Diagram for Power Operated Relief Valve Loss of Coolant Incident SHEET 1 oF 3

Fault Tree Logic Diagram for Power Operated Relief Valve Loss of (SHEET i)

Coolant Incident PjL%L +

R'6'HAIHS os'H ILloYG': UO7TE'o I3oXGS ACE IHiwuPEP OHLY FOK IAPGRAPISP PLAHT OCS ICrHS, POLY V-Itio1 GLocIC VLYQ) oPEHs E', FAILS FA I VS)

To KECLCSE To CLoSC PofCV POKV Fqutt m OFE'K ~ FAL LS OFEHS FA I LS TO F+ILuICS TO CLOSE KCCLOSS VIA/.

Vins Pyt.uiFYIC 5V Iq O3+P Potcv POICV I3LOCx VLV. l aLoCIC YLV. l (n4Ye ~)

OPI..HS oM PE H*HP

/III'ANIC oPCHS Sl LLK(OuSLY XA r I3 FA>L5 To CLo56 LFAtt- To Ct+

P>IEV i MECH. VLV. N'tx. +Aufo SIGHS

~ NE'cH. l ~YLK nTK. 1 fAu&.ctomi 1 viv. p,KKK I FAILS HoY HALF FALL5 Ho'T FAILS TO gKECf IVGOJ

<J SVXAHH I3VXAMY GvrAAS ovrA,M GvzII~m GvrGHT PLIrISAS IsvZESK SHg OF 3 Fi ure;A-2

Fault I'ree .Logic Diagram for Power Operted Relief Yalve Loss of 68rC6'VIATIOPS g (ZHrrT 1) -.

Coolant Incident 5RCAkflt PATH X COIITKOL ROOM Ke'MAIVS OPCH WAIF ~ NALFMHCTIOQ VlCCH. g(-, < HAIIICA I Mo ToR'PCXATOR 0PEi7.

PORv PoulFR OPCgATEP gg.lFF VALVC PORV V- I QOLI GLCCK VLVQ ope~ e sAILsl I AIL') VLV. . VALvF To R ~LE To CLOK PORV OP@'OS BIZ.V FAIiM TO RECLOSE GquIP+CIlV F AI LII R5'pIoR. To rAILs Cl.ose VLV.

CHOIR V2F'TR GV II405 4 P pohcy OPENS'H PORV oPCI4S GLosl< yLY.

'XA

+gLaai vw.

I ~g I (N>i's i)

DEHAIID PuRIOuSLY OILS To CLOM ~fhiLS To CLOs~e I I f'ORV 2IPOIe pazvz@s VLV. HTK 4%lb. SI6gA~L VLV. QRKR FAILS WOT i PAILS

/~

~KECeWeo J To CLoS'E gVllA~n GV ZA~T OV nAAS CV XC GR GYXSMQ 4)

DVJLGKf GyILSAS Gy L50K SHgPV 3 OF 3 Fi gure A-2

~g ENCLOSURE 4 RESPONSE TO NUREG 0660 XTEH XX.K.3.17 REPORT ON OUTAGES OF ECC SYSTEMS This report details the outages of ECC Systems since licensing of, St.

Lucie Unit 1 on March 1, 1976. The report was generated on the following bases:

A. Only those outages which resulted in less than the minimum required ECCS capacity for the plant mode are included.

B. Failure of a component or removal from service of a component which does not place the plant- into Technical Specification action statement (e.g;;removal of third Dump from service in the component cooling water system); were not considered because the plant still has the minimum systems,.

The report 'lists"outages ot.her than scheduled preventative.maintenence, chronologically by system. Scheduled preventative maintenance is indi-cated by notes in the listing for each system.

Xt should be"noted that with the 18 month, fuel cycle currently"in .effect *

"annual" preventative maintenance and overhauls performed during

-'he refueling will be extended to a eighteen month cycle. Additionally, St.

Lucie mechanical and control systems have been designed such that most periodic testing can be done without: taking the systems out of service.

k~ith one exceptio'n, tl>ose tests which require taking a system out of service .

are performed in operating. modes which do not require the system to be operable.

Cumulative reactor availability for St. Lucie Unit 1 was 26,356 hours0.00412 days <br />0.0989 hours <br />5.886243e-4 weeks <br />1.35458e-4 months <br /> as of October 31, 1980. No ECC system train had an outage rate of over 1%

for this period using the bases specified above. The approximate percen-tages are noted with the data for each system.

HIGH PRESSURE SAFETY INJECTION SYSTEM OUTAGE LENGTH DATES (HOURS CAUSE COMPONENT (S) ACTION TAKEN 4/26/76 16 Leaking Seal A HPSX Pump Repair 2/8/77 1 Dirty CCW Flowmeter A HPSI Pump Clean CCN Flowmeters 5/20/77 1 Leaking CCN Union A HPSI Pump Remake Coupling NOTES: (a) B and C-HPSI Pumps are redundant when pxoperly aligned mechanically and electrically. For preventative maintenance they are not xemoved from service simultaneously. A HPSX pump when removed from service. for preventative maintenance places one HPSI header out of service.

(b) Annual Preventative Maintenance Mechanical 'scheduled in December Time: 6 houxs Electrical - scheduled 'in Septembex Time: 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> t

(c) Semiannual 'Preventative Maintenance Scheduled in June and December . Time: 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (d) Quarterly Preventative Maintenance Scheduled in March, June, Sept '& December Time: 3 1/2 hours (e) Total annual A header preventative maintenance outage time excluding periods when system is not required is 34 houxs per year..

(f) The cumulative out of service pexcentage is .67,

LOW PRESSURE SAFE ECTION SYSTEM OUTAGE LEiNGTH .

DATES (HOURS) CAUSE 'COMPONENT (S) ACTIOiN TAKEN 2/5/77 3 Dirty CCW Flowmeters 1A LPSI Pump Clean CCW Flowmeter 2/9/77 '2 Dirty CCW Flowmeters 1B LPSI Pump Clean .CCW,Flowmeter 5/10-11/77 30 Cracked Weld 1A2 SIT Cooldown Parts (I,ER 77-29) 9/25/77 1 Check Timers 1A LPSI Pump Adjust ECCS Timers 9/25/77 1 Check Timers lB LPSI Pump Adjust ECCS Timers 11/21-22/78 38 Valve Operator Failed MV-07-1B Repaixs (LER 78-44) 2/22/79 1/2'4 Leaking Fill and Drain , 1B1 SIT Refill and Repressurize, Valve Maintenance (LER 79-007) 2/23-24/79 Allignment Check lB LPSI Pump Check Pump Alignment 2/22/79 12 Annual PM/Maint. 1B LPSI Pump 9/17/79 2 L'evel Trans Drift 1B1 SIT Repair (LER 79-29) 2/26-27/80 16 Annual PM/Maint 1B LPSI Pump 12/3-5/80 46 Seal Leakoff High lA LPSI Pump Repair NOTES: (a) Annual preventative maintenance schedule'd for Febxuary Time: 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Time: 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (b) Semi-annual preventative maintenance scheduled for March and September (c) The 2/22/79 and 2/26-27/80 preventative maint'enance dates axe included in the above list because the outage provided a convenient time to perform some additional discretionary maintenance.

(d) Total annual preventative maintenance time, including main-tenance during periods LPSI is not required, is 18 'nours per year per train.

(e) The cumulative out of service time for train A.was 0.4% and train B ia 0.7%, Total plant operation with less than required LPSI trains is 1.1%.

CONTAliPiENT S i OUTAGE LENGTH DATES (HOURS) CAUSE COMPONENT(S) ACTION TAKEN 2/11/77 2 Dirty CCH Flowmeter 1A CS Pump Clean Flowmeter 2/11/77 2 Dirty CCM Flowmeter 1B CS Pump Clean Flowmeter 8/17/77 1 Valve Not Fully Open 1A Header Valve Open Valve (LER 77-34) 9/8/77 6 PC/M 246-77 1B CS Pump Complete PC/M 9/8/77 6 PC/M 246-77 1A CS Pump Complete PC/M 9/25/77 1 Adjust Safeguard Timer lA CS Pump Adjust 9/25/77 1 Adjust Safeguard Timer 1B CS Pump Adjust NOTE: (a) Annual preventative maintenance is scheduled in March. Time: 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> (b) To avoid inavertant spray down of containment and Time: 1/2 hour per train introduction of sodium hydroxide to primary systems, per month the respective train is taken out of service fo" the monthly pump run.

(e) Cumulative time out of service for A header is 0.17% and for B header is 0.16%.

I C CONTAINMENT COOLERS NOTE: To date maintenance and preventative maintenance has:been performed in modes 5 or 6 when the coolers are not required. The plant intends to continue this policy. Three coolers operate at all times in operation. This is indicative of the reliability of these components.

COMPONENT COOLING WAT . STE1 OUTAGE LENGTH DATES (HOURS CAUSE COMPONENT'(S) ACTION TAKEN 8/30/77 .98 Leaking Mechanical Seal/ 1B CCW Pump Repair (LER 77-31)

Operator Error 10/2/80 .75'4 Battery Charger/ Elect. 1C and 1B CCW Pump Repair/Realign (LER 80-61)

Configuration 12/1-3/80 Leaking Line 1B CCW Heat Exchanger Repair (LER 80-67)

NOTE: Annual, semi-annual and quarterly preventative maintenance is performed with either the plant in a mode not requiring both CCW loops or the extra "C" "CCW pump is employed to maintain independent CCW loops.

INTA'OOLING WATER SYSTEf 8/26/77 Plugged Seal Line/ 1B ICW Pump Repair/Realign Electrical Electrical Lineup Plant (LER 77-35)

NOTE: Annual or semi-annual preventative maintenance is performed with either the plant in a mode not requiring both systems or the extra "C" pump is employed to maintain CCW requirements.

CHARGING AND VOLUME CONTROL SYSTr~f (CVCS)

OUTAGE LENGTH DATES (HOURS) CAUSE COMPONENT(S) ACTION TAKEN 2/20/77 8 Seal Lube Pump Failure A & B Charging Pumps Replace (LER 77-11) 2/26/77 13

'1 1/2 Seal Lube Pump Failure A Pump (B still OOS) Replace (LER 77-14) 10/2/78 Leaking Relief and Elec- B Pump (A OOS for Repair (LER 78-40) trical Interlock Maintenance)

NOTES: (a) Semi-annual and quarterly preventative maintenance is scheduled such that only one pump is out of service at a;time. Only two of three pumps are required.

occurred and resulted in L

(b) Outages of-individual charging pumps have generation of LERs. However; these events have not resulted in less than two charging pumps being in service at a time.

I AUXILIARY FEEDWATER SYSTEM OUTAGE . LENGTH DATES (HOURS) CAUSE COMPONENT(S) ACTION TAKEN 4/8/76 48 '0 Faulty Governor 1C AFW Pump Replace (LER 76-11) 4/13/76 Improper Wiring 1C AFW Pump Rewire (LER 76-13) 5/21/76 Lw Corroded Contacts 1C AFW Pump Replace/Seal (LER 76-22) 7/9/76 Moisture in Control Circuit 1C AFW Pump Dry (LER 76-35) 7/17/76 Moisture in Control Circuit 1C AFW Pump Dry & Seal (LER 78-35) 12/8/76 12 Corrosion in Latch Mech. 1C AFW Pump Dry/Seal (LER 78-47) 2/17/77 40 Moisture in Terminal Box 1C AFW Pump See Note (e) (LER 77-10)

Fail to Start Cause Un-

~

8/11/77 . 1 1C AFW Pump Op-Check (LER 77-33) known 2/9-10/78 28 Partially Shorted Winding MV-09-11 (C AFW Pump Check Motor (LER 78-6)

~ 'low Control) 2/10/78 Partially Shorted Winding MV-09-11 Replace (LER 76-6) 6/14/79 Steam Inlet Failed to Open MV-08-3 .(C AFW Pump Op Check (LER 79-20)

Steam Inlet).

i NOTES: (a) Annual preventative maintenance has been done during Mode 6 (refueling).

Plant intends to continue this policy.

(b) Semi-annual preventative maintenance for A, B & C pumps scheduled Time: 3 Hours in April and October.

(c) 'emi-annual proventative maintenance for C pump turbine scheduled Time: 4 Hours in April and October. C (d) Quarterly preventative maintenance for C pump turbine throttle Time: 1 Hour scheduled in January, February, July and October.

(e) Following the 2/17/77 incident on C AFW pump, steam line drains were rerouted, and electrical controllers were sealed. The absence or subsequent moisture and corrosion problems indicates that the change resolved this problem.

(f) Total annual preventative maintenance performed in modes requiring syste (g) The cumulative .time with less than the required AFW pumps is 0.6%.

DIESEL GENERA OUTAGE LENGTH DATES (HOURS) CAUSE COMPONENTS ACTION TAKEN 5/18/76 8 Clogged:Filter (Air Start) lA DG Clean (LER 76-21) 6/2/76 1/2 Adjust Timer 1A DG Breaker Return to Service 6/2/76 . 1/2 Adjust Timer 1B DG Breaker Return to Service 5/10/76 3 DG will not stop from CR lB DG Repair Switch 1/ll/77 1 Breaker Test Light Out 1B DG Breaker Replace 1/18/77 69 Turbocharger Failure 1B DG Replace (LER 77-2) 1/19/77 1 1/2 Stuck Linkage 1A DG Lubricate (LER 77-3) 1/20/77 8 Check Phase Balance 1B DG, Check, Return to Service 3/1/77 8 Operator Error lA DG" Reset Trip (LER 77-15) 9/20/77 65 Turbocharger Failure lA DG Repair (LER 77-42) 2/27-28/78 2 Clean Switchgear 1A DG Switchgear Return to Service 3/10/78 1 Install PC/M 1A DG 3/13/78 2 Install PC/M ' 1B DG 9/5/78 3 1/2 Breaker Failure to Close 1A DG Breaker Repair (LER 78-36 10/16/79 Failure of Voltage Reg. 1A DG Repair (LER 79-32) 9/3/80 6 Leaking Relief Valve 1A DG Replace (LER 80-55) 10/1/80 12. Leaking Relief Valve lA DG Vent Sys (LER 80-56)

NOTES: (a) Annual preventative maintenance has been performed during Mode 6 (refueling). Intent is to continue this practice..

(b) Semi-annual preventative maintenance scheduled .for Time: 3 Hours January and July (c) Monthly preventative maintenance scheduled monthly Time: 1 Hour (d) Total time req'uired for semi-annual and monthly preventative maintenance is 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

(e) The cumulative total out of service time for A DG is 0.7% and B DG is 0.6%. Total plant time is 1.3%.

HHERGENCY SAPHGUARD FEATURES ACTUATION SYSTEM A number of instanceshave occurred where single or'multiple HSFAS, channels have drifted out of specification or been found to be incorrectly set. These instances are listed below:

4/21/76 'wo containment pressure and two steam generator pressure bistables drifted out of specification.

Bistables were conservatively reset. (LER 76-17) 5/5/77 One steam generator pressure channel out of specification.

Incorrect setting was due to not setting following maintenance. (LER 77-25).

8/1/77 Two refueling water storage tank (RMT)

Levels out of tolerance duc to setpoint drift.

Instruments reset, (LER 77-32).

10/28/77 Channel A RRT level setpoint drifted. Bistable trip unit replaced. (LER 77-48).

1/3/78 'hannel A RMT Channel xeset.

level,setpoint drifted.

(LER 78-1).

Cause unknown.

2/19/80 Channel D RNT level setpoint drifted low. Channel was reset. (LER 80-'ll),

ENCLOSURE,5 FLORIDA POWER 6 LIGHT COMPANY ST LUCIE PLANT UNIT I CONTROL ROOM HABITABILITYREPORT

The St Lucie Unit 1 control room has been specifically designed to assure that the control room operators will be adequately protected against the effects of accidental release of toxic and radioactive gases and ."that the plant can be safely operated or shut down under design basis accident conditions.

The St Lucie Unit 1 control room habitability systems are designed to:

a) Limit control room personnel doses to within GDC 19 and SRP 6.4 guidelines, b) maintain C02 levels below one percent and 02 levels at a minimum of 17 percent at all times, c) maintain the ambient temperature required for personnel comfort and equipment operation at all times, d) withstand design basis earthquake loads without loss of function, and e) permit personnel occupancy during a chlorine, or other toxic .

chemical, release accident.

The location of the control room within the Reactor Agxiliary Building is shown in Figure 1. The St..Lucie Unit.l contiol =room 'ventilation and air conditioning system has undergone. extensive mode.fications, the results of which are shown. on the diagram provided as Figure 2.

The St Lucie Unit 1 control room habitability systems have been designed assuming that ten persons would be present in the control room during the accident. By assuming ten occupants, not only are the operators on the shift at the time of the accident accounted for, but also additional personnel such as health physicists.

I'I A potable water supply 'of 1.0 gallon per man per day is provided in a number-of plastic containers stored in the control room. The total amount of potable,.'water stored exceeds lOQ gallons. This water requirement conservatively allows for evaporated moisture and moisture losses in urine and feces. By Reference 1, 3.0 quarts per day per man is required at 75 P drybulb, while one gallon per day per man is the total recommended for drinking, food preparation, personal hygiene, and medical requirements.

To provide means for disposing of garbage, trash and human waste, the equivalent of sanitation kit III as recommended by the Office of Civil Defense for fallout shelters is provided and includes the following:

Paper, toiler tissue 5 rolls Plastic commode seat Heavy sanitary pads 1 dozen Regular sanitary pads 1 dozen Polyethylene gloves 1 pair Tie wires, bag closures Cups and lids 35 Cpn opener'anual}

Commode chemical

Polyethylene bag liners Instruction sheet Fiberboard boxes Fiber drum The above is recommended for a two week stay by 25 occupants in a survival shelter and is therefore conservative for this application.

A supply of food is stared in the control room which is sufficient to maintain habitability for ten men for a week.

The control room contains portable fire extinguishing equipment to permit the timely extinguishing of control room fires. Storage provisions for bottled air or chemox canisters are provided in the, control room 'for six hours of occupancy.

Toxic Chemical Release, A review of the locations and distances of industrial, military, and transportation fac'ilities and routes in the vicinity of the St Lucie site is included in Sections 2.2.1 and 2.2.2 of the St Lucie Unit 2 P

FSAR (see Attachment I}. This review has been performed using the guidance .from Standard Review Plan- Sections- 2.2 1-2.2.2 Rev. 1 and since both Units 1 and 2 occupy the same site, is applicable for the identification of potential hazards to the St Lucie Unit 1 control room personnel.

I A

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Because of the close proximi.ty of the St Lucie Unit 1 control room outside air intake to the St Lucie Unit 2 air intake, the evaluation of offsite toxic chemical sources contained in St Lucie Unit 2 FSAR Section 2.2.3 is also applicable to the St Lucie Unit 1 control room (see Attachment I). Since the issuance of the St Lucie Unit 2 FSAR it has been learned that chlorine, used,.for 'the city, water system and the sewage treatment facility, is stored in tanks located near the northwest corner of the St Lucie site. Because of the close proximity to the site an analysis of effects of a chlorine release accident on the St Lucie Unit 1 control room habitability has been performed. The results are given in Table 1.

The short separation distances between the air intakes and the locations at which toxic chemicals are stored onsite require that a separate analysis be performed for the St Lucie Unit. 1 control room for the chemicals: ammonium hydroxide, carbon. dioxi.de and cyclohexylamine. The arguments used in the St Lucie Unit 2 FSAR to eliminate the other chemicals stored onsite remain valid for the St Lucie Unit 1 control

$ 'goya ga well (pee pttachgent 7), Xn additi>n to the above listed chemicals which. will be common to both St Lucie Units. 1 and 2, chlorine

$ 'a. and wi;11 be used for the cieculating treatment at St Luci;e Unit 1 until the installation of a hypochlorite generator.

An analysis of the effects of an accidental release of the three above mentioned toxic chemicals and of onsite and offsite stored chlorine has been performed. The locations of these chemicals and of the St Lucie Uni.t 1 outside air intakes are shown in Figure 1,. The results of that analysis are provided in Table 1 and are summarized below.

Ammonium hydroxide is stored onsite in two 55 gallon drums at 30 percent concentration by weight. The concentration at 'the outside air intakes of the control room is calculated assuming all the ammonia in the solution becomes airborne instantaneously following a postulated rupture of the container. This very conservative assumption results in concentrations at the outside air intakes well is excess of those which can be actually expected.

In the case of carbon dioxide, complete vaporization is assumed immediately following accidental releases. The airborne transport of the puff is modeled using the instantaneous release diffusion model presented in Regulatory Guide 1.78. Since the control room is located a short distance from the release point and the amount of chemical is small, the model is adjusted to allow for additional dispersion in the 0

vertical direction by assuming uniform mixing between the ground and the elevation of the fresh air inlet (a 19 meter elevation from ground level is used}.

The concentration of cyclohexylamine in the feed tank in the Turbine Building is 10 percent. The chemical is delivered to the site in 55 gallon drums in the concentrated form and stored in a 20 ft by 12 ft storage room. Conservatively it is'ssumed that a 55 gallon drum fails. The evaporation rate of cyclohexylamine is calculated using Equation 2.1-18 in Reference 2. The transport of va'por is modeled by the short term, continuous release diffusion equation, presented in Regulatory Guide 1.4.

It is conservatively assumed that the centerline of the plume remains I

incident on the control room outside air intakes during the entire time it tak'es the liquid to evaporate. A ground level release is assumed. However, credit-is'<aken--'fox-addktional~dXs'persion. in the vertical direction by assuming uniform mixing between the ground and the elevation of the outside air intake.

The offsite chlorine is stored in two 150 lb cylinders near the sewage treatment facility and one 150 lb cylinder near the city water storage tanks. The airborne transport of a puff release of 25 percent of the closest tank contents is modeled using the instantaneous release diffusion model presented in Regulatory Guide 1.78. As with the carbon,.

ii dioxide analysis, the model is adjusted to allow for additional dis-persion in the vertical direction.

Chlorine used in the treatment of the circulating water is stored onsite in one-ton cylinders. In order to provide control room occupants protection against an accidental chlorine release, seismic Category I chlorine'etectors have been installed at the control room outside air intakes. An analysis has been performed to evaluate the concentration in the control room based on the failure of a one-toni.chlorine tank and the instantaneous release of 25" percent of its contents. The guidance given in Regulatory Guide 1.78 has been followed in modeling the diffusion of the chlorine cloud. The results of the analysis indicate that a.-maximum chlorine concentration of 25 ppm occurs in the control room 30 minutes following the detection of the gas. '.The toxicity limit is reached in 9 minutes. These time intervals .are

greater than the Regulatory Guide 1.78 requirement that the minimum time from detection to 15 ppm be at least two minutes, consequently, the control room occupants will have sufficient time to don their self-contained breathing apparatus. Table 2 lists the assumptions and parameters used in this analysis, The models described above are used to calculate concentrations ofc toxic chemicals at the control room outside air intakes. As indicated in Table 1, the concentration of ammonia and chlorine at the outside air intake of the control room exceed the toxicity limits. Since carbon dioxide and cyclohexylamine concentrations at the outside air intake are below the toxicity limits, the concentrations inside the control room are not required to be calculated.

The concentrations inside the control room are calculated based on the following equation:

C(.t) vt'e X(t')

dt'here:

C (t)= chemical concentration inside the control room at time t X (t')= chemical concentration outside, the air intake at time t'=

control room air exchange rate of 0.52 per hour which.'.is based on normal air intake rate of 920 cfm

I Por the chemicals analyzed, except chlorine, it is found that the concentration remains well below the toxicity limit under the assumptions that the control room is not isolated:. and no action is taken by the operators:following the accident. In the case of chlorine stored onsite, there is sufficient time for the control room occupants to don breathing apparatus. Therefore, an accidental release of toxic chemicals stored on and off site poses no threat to the control room operators.

Radiolo ical Release In the event of a postulated accident (LOCA) there could be three major sources of radiation exposure to the control room personnel: 1) direct radiation exposure from radioactive material outside of the control room, 2} submersion exposure from radioactive material within the control room, and 3) inhalation exposure from radioactive material within the control room. An analysis has been performed to insure that the St Lucie Unit,l control room personnel do not receive a combined dose from any accidental release of radioactivity which exceeds the limit of GDC 19 of 10CPR50 and the dose guidelines of SRP 6.4.

The source of the largest potential direct dose to the control room personnel is the control room emergency filtration system which is located in a room adjacent to the control room, separated from it by a one ft thick concrete wall. Using conservative assumptions, this system will contribute less than 1 rem over a 30 day period due to buildup of radioactive material on its filters.

The next largest potential direct dose is from the external atmosphere sur-I rounding the control room. This source arises 'from assuming a 0.5 percent I

ft I per day leakage rate of the containment atmosphere to the external atmosphere for the first day following a LOCA, and a 0.25 percent per 'day leakage rate E

for subsequent days. The control room is shielded from this source by at least two ft and as much as four ft of concrete. The time integrated dose calculated using an infinite cloud model,,which overestimates the answer, amounts to less than 0.2 rem.

The third largest potential direct dose arises from the radioactive'aterial released to the containment atmosphere following a LOCA. This source com-bines the dose from the material remaining in the atmosphere as well as that which plates out inside the containment. The magnitude of the initial source was determined assuming that 100 percent of the noble gas inventory and 50 percent of the halogen. inventory of the core is released immediately following a LOCA. Further,' halogen plateout factor of 0.50 is used. The activity at subsequent times was determined considering radioactive decay of the isotopes.

The control room is shielded from these sources by at least the three ft thick concrete shie3;d building. wall, the two inch thick steel containment wall, and the two ft thick control room wall. The source was modeled>>, as a cylindrical volume source, and no credit was taken for any additional shijlding from structures in the interior of either the Shie3d Building or the Reactor Auxiliary Building. The total dose then calculated. amounts to less than 0.1 rem.

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The remaining source of direct radiation,"..the Shield Building ventilation emergency filtration units, the atmosphere outside the control room but inside the Reactor Auxiliary Building, and the containment sump water, contribute a total dose to the control room of less than 0.005 rem for 30 days post-LOCA. The Shield Building ventilation emergency filtration units are. separated from the control room by at least four to six ft of concrete shielding. The atmosphere inside the Reactor Auxiliary Building becomes radioactive from the relatively slow leakage of contamination from the containment. This low leak rate combined with concurrent h

radioactive decay, results in a low dose to the control room. The 'final source, the radioactive water in the containment sump, is separated from the control room by at least, nine ft of concrete (measured perpendicularly through the shielding). As a consequence, its'.dose contribution is negligible,.under the most conservative shielding assumptions.

The analysis of the dose which the control room personnel would recieve following a postulated accident (L9CA) via the air introduced into the control room has been performed using the methodology of Murphy and Campe-".,

The St Lucie Unit 1 control room is designed to be maintained at a slightly positive pressure after a LOCA. The inhalation and submersion doses to its occupants are proportional to the makeup air intake rate, necessary to maintain the control room envelope pressurized at a dif-ferential pressure of at least 1/8 inch water gauge. The assumptions and parameters used to determine these doses are provided in Table 3, while the doses to control room personnel are reported in Table 4.

The analysis shows that the control room design meets the requirements of GDC 19 and SRP 6.4 without the use of self-contained breathing apparatus, bottled air, or potassium iodide.

REFERENCES

2. Toxic Va or Concentrations in '.the Control Room Followin a Postulated Accidental Release, J Wing, U S Nuclear Regulatory Commission, NUREG-0570, June 1979.

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0 TABLE 1 Toxic Chemical Evaluation Distance From Peak Conc. (ppm) Time to Reach Toxicity Control Room+ Quantity At Inside Toxicity Limit Note Ammonium 55. gal.;30%

Hydroxide 500 210 by weight 7.69(+4) 1.32(+2)

Carbon Dioxide 10000 450 360 SCF 4.40(+3) 1p2 Cyclohexylamine 20 100 55 gal;100% 3.45 273 by weight Chlorine (offsite) 15 580 38 lb 1.58(+3) 4.86 Chlorine 15 328 500 lb 25 (onsite)

• See Figure 1 for locations No'tee:1). The concentration was determined assuming that the toxic chemical becomes'instantaneously airborne following the rupture of the container.

2) Not a design basis event, because the calculated concentration at the outside air intake of the control room is less. than the toxicity limit.

3) The concentration was determined assuming that the toxic chemical evaporates following the rupture of the container.

'I TABLE 2 Regulatory Guide 1.78 Onsite Chlorine Release Evaluation Calculation Assumptions Chlorine container size 1 ton Distance to outside air intake ducts 100 meters Control room filter efficiency for chloiine 99%

Meteorological conditions Pasquf1L F 6 lm/sec Control room chlorine detectors sensitivity 5 ppm Control room volume 106, 920 ft Normal Control room air intake rate 920 cfm Time for chlorine activation detection signal* 6 seconds

• Measured from time when chlorine concentration outside control room intakes is 5 ppm, to the time of isolation.

TABLE 3 Assum tion and Parameters Control Room Volume Control Room ft P'2,550 Technical Support Center 44,370 ft3 TOTAL Outside Air Intake Rate Normal 920 cfm Emergency 450 cfm Emergency Recirculation Rate Through Charcoal Adsombers".; 1550 cfm Emergency Filtration System Charcoal Adsorber Efficiency Elemental Iodine 95%

Oganic Iodine 95%

Particulate Iodine 99%

Noble Gas 0%

Unfiltered Infiltration Rate 3 cfm Atmospheric Diffusion Factor 0-8 hours 4.86(-4) sec/m 3 8-24 hours 4.17 (-4) sec/m3 1-4 days 2. 80 (-4) sec/m3 4-30 days 1.59(-4) sec/m Occupancy Factor 0-1 day 100%

1-4 days 60%

4-30 days 40%

Finite Cloud Approximation Used to Estimate Whole Body Gamma Dose

TABLE 4 Radiolo ical Doses to Control Room Personnel Dose GDC 19/SRP 6.4 (Rem) Limit Rem)

Whole Body i Direct 1.Q Submersion 0;6 TOTAL 1,.9 5.0 Ski.n Submersion 30.0 Thyroid Inhalation 3,'8 ..0 30.0

/

ATTACHMENT 1 I'

SL2-F SAR 2.2

( ' z.'z.i NEARBY INDUSTRIAL LOCATION AND ROUTES TRANSPORTATION, AND MILITARY FACILITIES The St Lucie site is located on Hutchinson Island approximately six miles southeast of Ft. Pierce, Florida. Within five miles of the St Lucie site are: (a) six primary, secondary and light duty highways, (b) one rail line, (c) two airways, (d) intracoastal shipping lanes, and (e) sand mining operations (see Figure 2.2-1). Available data indicate. that no other facilities exist within either a five mile radius or, in terms of signif-icant facilities, a 10 mile radius of the plant (e .g., oil and gas(yips lines, military bases, chemical plants, drilling operations, etc.> )

2.

2.2 DESCRIPTION

S 2.2.2.1 Descri tion of Facilities There are no significant facilities within the plant vicinity that produce hazardous materials. For a description of nearby facilities, refer to Subsection 2.1.3.3.3.

2.2.2.1.1 Transportation Facilities One primary highway (US 1), three secondary highways (SR A1A, SR 712, SR 707) and two light duty roads (Walton Road, Easy Street), are within five miles of the plant. As indicated by Table 2.2-1, average daily traf-fic volumes on these routes during peak seasonal times (fall and winter) during 1977-1978 ranged from 1,016 to 19 f/51 )Increases in traffic volumes could be as high as six percent annually ', 13) The shortest linear distance from the center of the Reactor Building to each of "these highway corridors is listed below:

a) US 1 - 4.8 miles WSW b) SR A1A 0.2 mile E c) SR 712 - 3.8 miles NW d) SR 707 1.8 miles WSW e) Walton Road 3.4 miles SW f) Easy Street 3.1 miles NW Rail Paralleling the western shore of the Indian River (2.0 miles west south-west of the Reactor Building) is a Florida East Coast Descriptive statistics concerning this facility are provided below (~j)way.

g 2 ~ 2-1

SL2-FSAR a) Average daily number of 'trains - 21 b) Average train, size - 50 to 55 cars c) Maximum train size - No limit d) Number of passenger trains .- None e) Commodities transported " Rock, autos, building materials, perish-

'bles, piggyback shipments (FAK/Freight of all kinds) and any haz-ardous materials meeting the tariff regulations of the Interstate Commerce Commission (ICC) f) Tonnage shipped annually past the site - 7,959,098 tons It

~Airwa s Two airways are located approximately two miles to the east of the plant:

V259 and V3E. The two airways are used extensively by both IFR traffic (instrument flight rules; primary/y)commercial) and VFR traffic (visual flight rules; primarily private)

Wate~rwa s Commercial shipping lanes are located east and west of the plant. The Intracoastal Waterway .is located 1.2 miles to the west of the plant. The St Lucie County portion of the Intracoastal Waterway (a north-south trans-portation route extending the length of the east coast) passes through the Indian River. Atlantic Ocean shipping lanes are about 10 to 15 nautical miles east of the play(<)with north bound traffic lanes located farther east than southbound lanes 2.2.2.1.2 Quarrying/Mining Operations A small sand mining operation (employing two people) is located along the western shore of the Indian River approximately four miles of the plant site. No explosives are employed by these operations ~~r()gest 2.2.2.2 Descri tion of Products and Materials 2.2.2.2.1 Railroads The Florida East Coast Railway may transport any hazardous material com-plying with ICC tariff regulations past St =-Lucie Unit 2. The principle explosive substance transported is liquid petroleum gas (maximum tank size, 33,000 gallons): the princi~Je toxic substance transported is chlorine (maximum tank size, 90 tons) . Such materials may be included on all trains.

Within the past 10 years, two minor rail accidents (both derailments) have occurred within five miles of the plant: (a) May 15, at milepost (MP) 248.2- "wrung journal" (i.e., broken axle) on car FEC 12295, (b) August 26, 1974 at MP 257.1 brake rod broke, derailing car NW 292587. Neither 2.2-2

SL2-FSAR a

inczdent involved hazardous materials and totaj pcorded damage (involving only equipment and.track) approximated $ 762.00 Truck Carriers I'.2.2.2.2 No data were available on truck traffic or truck shipments within five miles of St Lucie Unit 2, although existing records indicate that no truck rela)yf)accidents involving hazardous materials have occurred within the area Since there was very little information on tru~)0)raffic in the vicinity of the plant, the Applicant performed a survey to get an indication of the amount and type of truck traffic on the roads within a five mile radius of the site. The survey was performed between January 30, and February 6, 1979 and consisted of collected information on US 1, SR A1A, SR 707, SR 712, and Walton Road. The survey consisted of the following:

a) A collection of existing information from the state and the county regarding traffic counts and accident characteristics.

b) Twenty-four hour truck classification counts taken at 13 locations.

Each truck was classified by the number. of axles and whether it was marked a's carrying hazardous material. The type of hazardous material was also noted.

c) A roadside interview on US 1 of trucks marked as carrying material was conducted on VS 1 for a total of 18e5 hours over a two day period. Information on the type and amount of hazardous material being carried by these vehicles was collected. The interview sta-tion was located on US 1 because it is the most heavily traveled roadway in the five mile area and has the majority of truck traffic of the roads in question.

d) Automatic traffic recorder counts were obtained at 10 locations for a seven day period.

e) Contacting propane gas supply companies in the area to determine if deliveries were made within the five mile radius; the type of material being transported; and size and capacity of these trucks.

The survey locations are shown on Figure 2.2-2. The results of the survey are given in Tables 2.2-2, 2.2-3 and 2.2-5 and shown on Figures 2.2-3 and 2.2-4.

a The average daily traffic count and the average weekday traffic count de-termined fran the survey are given- in Table 2.2-2 and shown on Figure 2.2-3. The count data are consistently higher than the volume figures obtained frcn the State of Florida. The difference is apparently caused by the increase in seasonal activity on all routes in the area during this time of year (January-,February), and in traffic volumes on SR A1A due mainly to construction activities at St Lucie Unit 2.

2 ~ 2" 3

SL2-FSAR The truck volume and the volume of trucks displaying hazardous material placards are also represented in Table 2.2-2 and Figure 2.2-3 and are based on the trucks observed during the classification counts.

In summary, the trucks observed during the study period comprised from 1.3 to 6.4 percent of the total traffic. These values are cpy~rable to the normal truck percentages, which are usually five percent , Trucks carrying hazardous material comprised from 0 to 16.7 percent of the total truck traffic as indicated on Table 2.2-2.

A summary of the trucks marked hazardous and interviewed on US 1 is given in Table 2.2-3. Though this is a limited sample, it does give an indica-tion of the type and amount of hazardous material transported within the five male radius.

On SR A1A, the majority of the truck traffic services the St Lucie site.

Table 2.2-4 xs a description of the type, size and frequency nf truck shipments of compressed gases and process chemicals to the St Lucie site .

In addx,tion to interviews and classification counts, four propane gas companies known to make deliveries in the area were contacted to determine

, the type and amount of material being transported and the size and capar-ity of their trucks; the results are as follows:

a) Tro x as Stuart transports liquid propane gas in 2,000 gallon tanks on two axle and three axle trucks into the area of the five mile radius. They deliver once a month on SR 707 as far north as the plant site and they deliver once a month to "Venture Sales" located on SR AlA in the area of Nettles Island.

b) Tro i as, Fort Pierce - is similar to a> above . However, they deliver once a month during summer months and twice a month during winter months to locations along SR 712 and SR 707 as far south as the plant site.

c) Tri-Count Gas Inc, Stuart - transports liquid propane gas in quan"

.tities up to 2,150 gallons on three axle trucks into the area of the five mile radius. They presently service St Lucie Unit 2 once or twice a week for welding operations. They also service "American Resort" in the vicinity of Nettles Island once a month and deliver along SR 707 and SR 712 once a month.

d) Econ-O-Gas, Inc, Stuart - makes no deliveries into the site area.

Information on truck accidents is presented in Table 2.2-5 and Figure 2.2-4. Between January 1, 1973 and December 31, 1977, 19 accidents involving trucks occurred on US 1 and one accident occurred on SR AlA within the five mile radius. Between January 1, 1973 and December 31, 1976, one accident occurred on SR 707. Within the last five years there were no ~Mjdents within the St Lucie County involving hazardous material

2. 2-4

SL2-FSAR 2.2.2.2.3 Waterborne Commerce As is indicated in Table 2.2-6, 21 different types of.commodi)jp are regularly shipped past the site via the Intracoastal Waterway ing 1975-1977, residual fuel oil constituted 56 percent of all shipments (by weight/tons). Other major types of commodities shipped past the site during this same period included nonmetallic mineral products (10.3 percent) and sugar (9.0 percent). Although no data are available concern" ing shipping in the Atlantic Ocean, the U.S. Coast Guard estimates that 40 to 50 ships pass the site each day. Approximately half of t)jg)traffic (i.e., 25 to 30 ships) is estimated to carry petroleum products Onsite Products and Materials '.2.2.2.4 Compressed gases and process chemicals located on the St Lucie site for operation and maintenance purposes (and stored in standard industrial high pressure cylinders) include the following:

Com ressed ases a) Acetylene - approximately 25 bottles, {360 scf) b) Oxygen - approximately 25 bottles, (360 scf) c) CO - approximately 80 bottles, '(360 scf) d) N2 40,000 scf tube, trailer 40 bottles, (360 scf)

Liquid Dwyer (several hundred gallons)

Hydrogen 40,000 scf tube trailer 80 bottles, (260 scf)

Process Chemicals a) Cyclohexylamine - two 55 gallon drums b) Ammonium hydroxide two 55 gallon drums Hydrazine eleven 55 gallon drums d) Potassium Dichrnmate 200 pounds e) Sodium Hydroxide 3800 gallons Sulfuric Acid 3000 gallons Other ases limited to a small number of bottles a) Argon b) Met)lane c) Propane

2. 2-5

SL2-FSAR d) Laboratory specialty gases 2.2.2.3 ~Pi elines No pipelines are located within five miles of the plant (3,4) 2.2.2.4 ~Wetetwe s The Intracoastal Waterway (10 foot channel depth) passes 1.2 miles west of the plant. All intake structures are located to the east of the plant and open into the Atlantic Ocean. Four types of vessels utilize the Intra-coastal Waterway: (a) self propelled passenger and dry cargo vessels, (b) non-self propelled dry cap~)vessels, (c) non-self propelled tankers, and (d) towboats and tugboats 2.2.2.5 Air orts and Airways No major airports exist within 10 miles of the plant. Approximately nine miles west northwest of the site there is a private airport ~p))ed Sunrise.

Within nine and 50 mil.es of the plant, there are 26 airports . Loca-tion data for all airports within 50 miles of the site are provided in Table 2.2-7. Based on available data, no within the 50 miles area records operations at or beyond(/yves)s2gf2$ )0d airport (within 10 miles) or 1000d (within 10 to 50 miles) ' ' The estimated number of IFR flights occurring in airways V295 and V3E (calculated within a 20 mile radius of the site) is 250,000 annually; VFII )~affic in this same area is estimated to equal 500,000 flights annually 2.2.2.6 Pro'ections of Industrial Growth Between 1980 and the year 2000, light manufacturing activity(jg)St Lucie County is projected to increase by approximately 111 percent . In 1980, approximately seven percent of total earnings** or $ 13.1 million is expected to be derived from manufacturing activities. By the year 2000, earnings derived from light manufacturing ($ 27.7 million) are projected to approximate two percent of total earnings. During each of these benchmark years, approximately 60 percent of manufacturing-based earnings is expected to be derived from two industrial sectors: foods and kindred products, (primarily citrus) and chemicals and allied products (primari1y fert-ilizers). Throughout this time frame, relatively small amounts of manufac-turing activity are expected to occur in four additional industrial sec-tors: printing and publishing; metal fabrication; machinery manufacturing; and el.ectrical equipment manufacturing.

Such increases in manufacturing activity as may occur in St Lucie County are expected to be contained in relatively high jg)ensity nuclei located aLong major highways - especial 1y I 95 and US I . Given such a development strategy, increases in manufacturing activity within five miles of St Lucie Unit 2 may be anticipated primarily along US 1 in the vicinity of Port St Lucie (i.e. approximate Ly five miles west of the pLant). s

2. 2-6

SL2-FSAR Refers to. distance (d) in miles from the site.

• "'arnings include income derived from wages, salaries, proprietary and miscellaneous income.

2 ~ 2-7

SL2-FSAR 2.2.3 EVALUATION OF POTENTIAL ACCIDENTS There are no design basis events7external to the plant that have a probability of occurrence of 10 per year or greater and have potential consequences serious enough to affect the safety of the plant to the extent that 10CFR100 guidelines could be exceeded.

2.2.3.1.1 Transportation of Explosives and/or Flammables on the Atlantic Ocean and'ntracoastal Waterway The Atlantic Ocean shipping lanes are about 10 to 15 nautical miles east of the plant (refer to Subsection 2.2.2.1.1). Hence, with a distance of 10 miles, no ship or barge explosion can affect the plant structures.

Due to the Intracoastal Waterway channel depth of 10 ft, the size of barges passing the plant site is limited. The waterway depth is nominally assumed to be capable of handling nin~3@et draft vessels which transport a max-imum load of about 16,000 bbl '. However in actual practice, trans-porters are .,limited to loads of about 7,000 bbls per trip on barges of no more a six feet draft because the Intracoastal Waterway is not dredged often ))g . As indicated in Table 2.2-6, the commodities of concern regarding explosions are gasoline'nd petroleum.

Gasoline is used as an example for calculating explosion overpressures.

According to Robert P g~yjict, the upper limit of flammability for gasoline is 7.9 percent ~ The highest limit of flammability for the gasoline family stated by the Bureau of Mines is 10.5 percent for cyclo-propane (3z) . Mr. Benedict has stated that although the density of gasoline vapor at the highest limits of flammability is unavailable, the combination of a 10.5 percent limit of flammability and a gasoline vapor density of 0.245 ibm/ft (which corresponds to the vapor density of heptane) at this limit is conservative.

Thy free volume of 16,000 bbl barge is 16,000 bbl x 42 gal/bbl x 0.1337 ft /gal 89,846 ftg

. Using a conservative 10.5 percent gasoline-~ir mixture (i.e., 0.105 of volume) at a vapor density of 0.245 ibm/ft there are 2311 ibm of gasoline in a 16,000 bbl barge. Assuming an ex-tremely conservative upper bound of mass equivalency at 240 percent, 2311 ibm of gasoline vapor yields a detonation equivalent. to 5547 ibm TNT.

From Equation 1 of Regulatory Guide 1.91, "Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants,"

February 1978 (Rl), the calculated safe overpressure distance is R >k W > 45 (5547) > 797 ft.

1 where; R ~ distance in feet from an exploding charge W ~ pounds of TNT

2. 2-8

SL2-F SAR K = constant = 45 s

Therefore a one psi peak positive overpressure will not occur at a distance greater than 797 ft. for a 16,000 bbl barge of gasoline vapors. Since the Intracoastal Waterway shipping channel is over 6000 ft away from any safety related structures, no damage occurs from a barge explosion.

Explosion generated missiles are also considered as follows:

In order to determine the distance through which generated missiles can travel, consider the exploding volume of 89,846 ft as a sphere (radius =

27 '8 ft ) with energy equipartitioned through the exploding volume.

The explosive energy (E ) assuming 5547 ibm TNT is e

E ~ 5547 ibm x 500 kcal x 3.968 Btu = 1.1 x 10 7 Btu and the energy density (E) is 1.1 x 10 Btu = 122.5 Btu/ft free volume 89,8 ft In a deflagration type of explosion the maximum energy density imparted to potential missiles cannot exceed the energy of the explosion. Hence, the kinetic energy (KE) of a potential missile cannot exceed KE(Btu) ~ E(Btu/ft ) x M m

(ibm) -, P

,m (ibm/ft )

where M and p are the mass and density of the potential missile.

With a kinetic energy (KE) = 1/2 M V , where V is the missile speed, the maximum range (R ) of a missile is

, max R

max

= V2 2KExl~2EMm I= 2E g M g Pm M g Pmg m m Using steel as an example with a density of 489 ibm/ft , the maximum range calculated of a potential missile is x 122.5 x 778 ft-1bf x 32.2 1bm-ft 2 ~bt ft Btu 1bf set-R max

= 389.8 ft 32.2 ft/sec x 489 ibm/ft The above equation does not include consideration of air resistance or energy lost in rotation, which would. decrease the range of any generated missile. Thus there is no haza'rd from a barge explosion due to missiles.

2.2.3.1.2 Transportation of Explosives and/or Flammables by Truck or State Road A1A A review of the truck traffic reveals that the governing explosive and/or flammable event would arise on SR A1A which passes about 750 ft east of

2. 2-9

SL2-FSAR the diesel oil storage tanks due to a liquified propane truck accident.

Based on the limited amount of hazardous truck movements past the site, the probability of having a potential accident whose consequence can result in ~adionuclide releases in excess of 10CFR100 guidelines is less than 10 per year as described below.

Based on accident data for a five year period (January 1, 1973 through December 31, 1977) provided by the Florida Department of Transportation, there has been only one truck accident within five miles of($ site on SR AlA in a total of 2,600,000 truck vehicle miles traveled /~ . There-fore, the probability of any type of truck accident is calculated to be 3.8 x 10 truck accidents per vehicle mile. IL'his site specific probability is much smaller than the 1.3 x 10 truck accidents per vehicle mile probability for a r~~lc truck accident in the "minor" severity category predicted in WASH-1238 . Therefore the WASH-1238 probability is used and gives a conservative estimate of the frequency of truck ac-cidents in the site vicinity.

To calculate the probability that hazardous flammable liquids explode due to a spill, it is necessary to determine the conditional probability of a spill and the conditional probability of an explosion occurring due to a spill. Although there have been no accidents within St i~~ie County involving hazardous material within a period of 1973"1978 $ , the prob~)j)ity of a spill as a result of an accident is estimated at 0.02 , since two percent of accidents involve a tank truck with suf-ficient impact to cause rup'ture of tank. The probability of an explosion due to a spill as determined by f)))Department of Transportation's Office of Hazardous Materials is 0.0113 Thus, the probability(Pe) associated with an in-5ransit explosion of a truck is 1.3 x 10 x 0.02 x 0.0113 mz 3.6 x 10 explosions per vehicle mile.

The number of vehicle miles per year for the transport of hazardous material in the one mile stretch of SR AlA incident to the site can be estimated. The annual number of liquified propane gas truck deliveries on SR AlA in the vicinity of the site is 27 shipment/yr (as described in Subsection 2.2.2.2.1). Assuming all these trucks travel through the one mile stretch of road incident to the site, 27 chicle mile/yr can be estimated. Using the probability of 3.6 x 10 explosions/vehicle mile, the probability of an explosion is 9.72 x 10 per year, in the one mile stretch of SR AlA closest to the site.

Since the probability is less than 10 per year, an explosion of a truck carrying hazardous material is not a design basis event.

2.2.3.].3 Transportation of Explosives and/or Flammables on the Florida East Coast Railway The Florida East Coast Railway runs about two miles west southwest of the plant site (refer to Subsection 2.2.2.1.1). Since the rail line can ap-proach the safety related structures no 'closer than the distances computed in Figure of Regulatory Guide ) .91(Rl ), no further consideration 1

0 need be 2.2-10

SL2-FSAR given to the effects of blast in plant design. This two mile distance is greater tha~ ge ranges of fragments of the train accident in Laurel, Mississippi . The range of the "rocketing" car in the Laurel, Mississippi. accident was 1100 ft while small fragments had a maximum range of 1600 ft.

Thus there are no hazards from "rocketing" rail cars or their fragments for St Lucie Unit 2 safety related structures.

2.2-11

SL2-FSAR 2.2.3.2 Desi n Basis Toxic Chemical Events

2. 2.3.2. 1 Introduction The accidental release of toxic chemicals may affect control room habit-ability. Based on information presented in Subsection 2.2.2, the potential sources are analyzed in detail to determine the threat to the control room operators.

Table 2.2-8 contains a list of each of the toxic chemical stored or trans-ported in the vicinity of the plant. T)ose specific events which are found to have a probability of less than 10 per year are considered not to be design basis events.

2.2.3.2.2 Assumptions and Methodology Based on information presented in Subsection 2.2.2, Table 2.2-8 includes a list of hazardous chemical sources which are considered in evaluation of potential accidents. Consideration is limited to those chemicals which are present within a distance of five miles from the control room air intakes.

Chemicals stored or situated at distances greater than five miles from the facility are not considered because, if a release occurs at such a dis-tance, wind speed and atmospheric dispersion will dilute and disperse the incoming plume to such a degree that there will be 'sufficient time for the control room operators to take appropriate action, if any is required . In addition, the probability of a plume remaining within a given sector for a long period of time is quite small.

Facilities Located within five miles of the plant do not sto're, use or

, produce large quantities of hazardous substances. However, some quantities are stored on site as indicated in Table 2.2-8. There are no toxic chemi-cals transported by waterborne commerce and road in significant quantities that may affect the safety of the plant following accidental releases ~

Consequently such sources are not evaluated. There are no pipelines lo-cated within five miles of the plant and so this source is aLso not con-sidered. The amounts of toxic chemicals transported by the Florida East Coast Railway (FECR) are greater than those specified in Regulatory Guide 1.78. "Assumptions for Evaluating the Habitability of a Nuclear Power PLant Control Room During a Postulated Hazardous Chemical Release",

June 1974 (RO). Therefore releases of toxic chemicals due to railroad accidents are considered in the analysis..

In order for the control room operators to become exposea to one of the toxic chemicals listed in Table 2.2-8, the following chain nf events must occur . First, the container in erich a given chemical is enclosed must somehow fail and release its contents. Second, the chemical must be sufficiently volatile to become airborne. Third, at the time of release, the direction of the wind must be such as to transport the airborne material from the point of release to the control room outside air intakes. The airborne material has to be sufficiently stable in air not to condense on the ground, or burn or explode, or otherwise Lose its toxi-city prior to reaching the outside air intake. The quanti ty of the chemi-cal which becomes airborne has to be sufficiently large and dispersion in air sufficiently low, for the concentration nf the toxic agent to build up

2. 2-12

SL2-FSAR to toxic levels in the control room atmosphere before the operators can take protective action.

Chemicals that are nonvolatile solids or liquids, or that spout~ueously combust in air do not pose a threat 'to control room habitability. Con-sideration of these factors leads to the elimination of the following chemical sources from toxic hazard evaluation. Solutions of sodium hy-droxide and potassium dichromat'e are eliminated because, while the sol-vent may evaporate, the solute is nonvolatile. SuLfuric acid is elimi-nated due to its low voLatility. It is an oily liquid with a vapor pres-sure oE only 0.0008 mm Hg at 25 C and its evaporation rate is negligible under ambient atmospheric conditions. Similarly hydrazine stored on site is, eliminated because its partial vapor pressure in the solution of 30

, percent concentration is only 3.7 mm Hg under ambient conditions.

Table 2.2-8 also indicates that many chemicals are eliminated because their potential for ignition constitutes a greater hazard than their toxicity.

When a flammable or explosive substance is released, it is highly likely that its vapor will explode or burn before reaching the control room.

Therefore, the only chemicals considered to present a potential danger to control room operators, are those whose toxicity limits are lower than their lower Limits of flammability. This leads to the elimination of chemicals such as hydrogen, acetylene, natural gas, propane, butane, and other flammable hydrocarbons.

These toxic chemicals in Table 2.2-8, which are not eliminated on the basis of criteria discussed above are shown to pose no threat to control room habitability by a detailed assessment of their atmospheric transport and potential for infiltrating ..co the control roan atmosphere. The at-mospheric dispersion condition is conservatively assumed to be stability Class F and 1.0 m/sec wind speed.

Ammonium hydroxide is stored onsite in two 55 gallon drums at 30 percent concent'ration by weight. The concentration at the outside air. intakes of the control room is calculated assuming all the ammonia in the solution becomes airborne instantaneously following a postulated rupture of the container. This very conservative assumption results in concentration at the outside air intakes well in excess of that which can be actually ex-pected .

In case of carbon dioxide, complete vaporization is assumed immediately following accidental release. The airborne transport of the puff is modeled using the instantaneous release di ffusion model presented in Regulatory Guide 1.78 (RO). Since the control room is located at a short distance from the release point and the amount of chemical is small, the model is adjusted to allow for additional dispersion in the vertical di rec-tion by assuming uniform mixing between the ground and the elevation of the fresh air inLet (a 23 meter elevation frcm ground level is used).

The concentration of cyclohexylamine in the feed tank in the Turbine Build-ing is 10 percent. The chemical is delivered to the site in 55 gallon drums in the concentrated Eorm and stored in a 20 ft by 12 Et storage room.

Conservatively it is assumed that a 55 gallon drum fails. The evaporation rate of cyclohexylamine is calculated using equation 2.1-18 in NUREG

2. 2-13

SL2-FSAR 0570 . The transport of vapor is modeled by the short term, con-tinuous release diffusion equation presented in Regulatory Guide 1.4 "As-sumptions Used for Evaluating the Potential Radiological Consequences of a LOCA for PWR", June 1974 (R2). It is conservatively assumed that the cen-terline of the plume remains incident on the control room outside air in-takes during the entire time it takes the liquid to evaporate . A ground level release is assumed. However, the credit for additional dispersion in the vertical direction by assuming uniform mixing between the ground and the elevation of the outside air intake is taken.

The models described above are used to calculate concentrations of toxic chemicals at the control room outside air intakes. As indicated in Table 2.2-8, the concentration of ammonia at the outside air intake of the con-trol room is the only chemical expected to exceed the toxicity limit foL-lowing the rupture of a 55 gallon drum containing 30 percent ammonia by weight. Since carbon dioxide and cycLohexylamine concentrations at the outside air intakes are below the toxicity limits, the concentrations in-side the control room are not required to be calculated.

The ammonium hydroxide concentations inside the control ro~m as calcu]ated based on the following equation (see Appendix 15B):

t C(t) = e ve. X (t')

dt'here:

C (t) = chemical concentration inside the control room at time t X (t') = chemical concentration outside the air intake at time t' control room air enchange rate of 0.46 per hour which is based on normal air intake rate of 750 cfm Based on the above equation, the concentration of ammonia inside the con-trol roan remains well below the toxicity limit under the assumptions that the control room is not isolated and no action is taken by the operators to

, do so following the accident. Therefore, ammonium hydroxide stored onsite poses no threat to the control room operators.

Chlorine is the principal toxic substance transported by the FECR (2.0 miles west south west of the plant). Since the quantity, pershipment, of chlorine (90 tons) shipped past the site is greater than the adjusted quantity given in Table C-2 of Regulatory Guide 1.78 (RO), the shipments are considered in the hazardous chemical analysis.

There have been a few minor railroad accidents within five miles from the plant in the past 10 years which resuLted in smaLl dama,es. nasea on information presented in Regulatory Guide 1.78 (RO), releases in the amount of 30 tons or more of chlorine at the railroad requi re considerarion in an evaluation of the control room habitability. A release of such magnitude is assumed to be equivalent to the total loss of a railroad car with a capacity of 90 tons or less, i.e ., the accidental release of the entire contents of. chlorine from a tank car is assumed to be an initiating event for' design basis accident.

2. 2- 14

SL2-FSAR The probability of such an event is given by the following equation:

P 1

P x N x M1 x D.

j x F.l j1 where:

Pil ~ annual probability of design basis event under atmospheric stability Class 1 involving the i-th chemical.

P = probability of a design basis accident for a mobile source H. =

1 annual numbers of trips involving the i-th chemical.

Ml = annual probability of an atmospheric stability class.

D.

, 3 the length of road; rail or river in sector j.

F jl wind frequency from sector j to outside air intake of the control room for stability Class n ~ number of wind direction sectors.

Based on tile number of trains, movements, cars per train (see Subsection 2.2.2.1.1) and the length of the track ne~r the site (approximately 9.2 miles), there are approximately 3.52 x 10 railroad car-miles traveled per year within five miles of the site. Within the past 10 years, two rail accidents have 'occurred within five miles of the Ilant. Therefore the frequency of an accident at the sjte is per mile. National statistics indicate that 33 x 10 events5.()g per mile result in 10 total loss of chlorine contents from a car. The ~g(jonal statistics also indicate that the frequency of rail road accident is 8.1 x 10 per mile. Therefore the frequency of an eveyt that results in the )otal loss of the contents from a car is 4.1 x 10 (=33 x 10 /8.1 x 10 )

per accident. Assuming the same relationship is applicable in the vicinity of )he plant, the8probability yf a design basis accident P, is 2.33 x 10 ( 5.68 x 10 x 4.1 x 10 ) per mile of travel distance. An avera~p yumber of 14 cars carrying chlorine are shipped per month by the FECR . Therefore the annual number of trips, Ni, involving chlorine are 168.

Table 2.3-81 indicates that the atmospheric stability frequency, M , is 4.16 and 1.13 percent for stability Classes F and G, respectively. Al-though atmospheric stability classes A through E are considered, the con-trol ronn habitabi lity is not affected under such meteorological condi-tions. The length of each segment of railroad, D., within each sector is shown on Figure 2';2-5. The required wind directidn from each segment of railroad towards the outside air intakes of the control roc m is also shown on Figure 2.2-5. The wind frequency F. , for stability Class F and G is obtained from Table 2.3-34. The proba6flity P. , of a design basis event under stability Class F in each segment is then:

il'.2-15

SL2-F SAR "2: -9 -9 Segment 1 P.

1 2.33x10 x168x0.0416x0.0617x2.66 2.'67xlO /yr Segment 2-3: P.= 2.33x10 x168x0.0416x0.1122x1.14 2.08x10 /yr 1

Segment 3-4: P.= 2.33x10 x168x0.0416x0.0813x0.823=1.09x10 /yr Segment 4-5: P.~ 2.33x10 x168x0.0416x0.0631x0.835=S.58x10 /yr L

5-6: -9 P.= 2.33x10 xl68x0.0416x0.0449x1.2~8.77x10 -10 Segment /yr Segment 6-7: P.~ 2.33xlO xL68x8.0416x0.0757x2.51=3.09xLO /yr Total 1.07x10 yr The probability of an event under stability Class F for the entire hazar-dous traveL dis)ance of 9.2 miles is the sum of the values calculated above and is 1.07x10 per year. Similarly, the proba)ility of an event under stability Class G is calculated to be 3.27 x 10 per year. Therefore an overall probability of an event that may affect control room ha)itabili ty is 1.4 x 10 per year. Since the probability is less than 10 per year, the'elease of chlorine due to a railroad accident is not a design basis event.

2.2.3.2.3 Results The accidental releases of chemicals stored on site and transported in the vicinity of the plant are found not to present undue risk to control room operators. Therefore, no detailed analysis is required in Section 6.4.

2.2.3.3 Fires in the Vicinit of the Site There are no industrial and chemical plants or storage facilities, or pipe-lines containing oil or gas adjacent to St Lucie Unit 2. The potential hazard from fires offsite are negligible because no flammable mass of appreciable size exists in the area.

In the unLikely event that a barge spills oil or gasoline accidentally on the Intracoastal Waterway, the spill would not only have to travel to the Hutchinson Island shoreline (approximately 3000 ft) but would have to tra-vel across 3000 ft of Big Mud Creek, basically a stagnant body of water.

The ultimate heat sink barrier wilL stop the flow of water from Big Mud Creek to the intake structure.

'uch a fire could affect the St Lucie Therefore, it is highly improbable that site..

'V 2.2.3.4 Collisions with Intake Structure and Li uid S ilLs Because the plant cooling water intake structure. is located in a commer-cially non-navigable area offshore in the Atlantic Ocean, no reasonable hazard exists from barges or ships that pass the site and no corrosive li-quids or oils accidentally reLeased could enter the intake structure.'.2-16

SL2-FSAR SECTION 2.2: REFERENCES Ament, G. - Personal Communication. St Lucie County Zoning 6 Building Department. October 5, 1978.,

2 ~, Brown, R.L. Personal Communication. Fort Pierce Utilities Authority, October 24, 1978.

3. Bekcham, R.E. Personal Communication. Florida Gas Transmission Company,,October 20, 1978.

Hurley, J. - Personal Communication. American Petroleum Institute, October 12; 1978.

5. Rudi, P.J. - Personal Communication. St. Luc ie County Office of Disaster Preparedness, October 5, 1978.

6. Sterba, D.E. Personal Communication. Florida Gas Transmission Company, 1978.

7 4 Walthers, L. Personal Communication. Treasure Coast Regional Planning Council, October 5, 1978.

8. Londy, M. - Personal Communication. Ft. Pierce Utilities, December 18, 1978.

9. Childress, R.J. Personal Communication. Superior Fertilizer and Chemical Company, October 26,~ 1978. '

10. Pearson, M.R. Personal Communication. International Minerals and

,Chemical Corporation, October 23, 1978.

11. Hillsman, R. Personal Communication. Indjantown Plastics Divi-sion, October 5, 1978.

12, St Lucia County. Traffic Corridors Input Data, 1978.

13. Ament, G. Personal Communication. St Lucie County Zoning and Building Department, December 20, 1978.

14. Wyckoff, R.W. Personal Communication. Florida East Coast Railway Company, October 27, 1978.

15. United States Department of Commerce, National Oceanic and Atmos-pheric Administration. Sectional Aeronautical Chart, Miami.

September 7, 1978.

16. Carr, M. United States Coast Guard, Miami. Florida. January 29, 1979.

17. Wyckoff, P.W.,- Personal Communication. Florida East Coast Railway Company, November 10, 1978 and February 8, 1979.

2.2" 17

SL2-FSAR SECTION 2.2: REFERENCES (Cont'd)

18. Hil'1, J.E " Personal Communication. Florida Public Service Commission, October 23, 1978,.

'19. Hill, J.E. - Personal Communication. Florida Public Service

-Commission, November 9, 1978.

20. Champagne Associates, Vehicle Classification and Product Containment

~Stud , Prepared 9 for Enviroaphere Company, March, 1979.

21. Washington, D.C. Highway Research Board, Hi hwa Ca acit Manual, 1965, PQ. 142.

22. Flaherty, H. P. Personal Communication. Department of the Army, Lower Mississippi Valley Division, Corps of Engineers, January 10, 1979.

23. Fort Pierce Port and Airport Authority Personal Communication.

October 24, 1978.

24. St Lucie Skyways - Personal Communication. October 30, 1978.

25 ~ Vero Beach Airport Personal Communication. November 15, 1978.

26. Palm Beach International Airport Personal Communications.

27. Carter, G.M. - Personal Communication. Miami Air koute Traffic Control Center, November 22, 1978.

28. U.S. Dephrtment of Commerce, Bureau of Economic Analysis.

29; The RMBR Planning/Design Group.. Com rehensive Plan: St Lucie

~Count June, 1974.

30. Mr Reed Patchen, Belcher Towing Company, Cape Canaveral, Florida, Personal Communication on July 31, 1979 and August 10, 1979.

31. Robert F. Benedict, Chief Supervisory Engineer Cryogenics, Chemical Engineers Construction Corp., New York (Personal Communication).

32. Coward and James, "Limits of Flammability nf Gases and Vapors",

Bureau of Mines, Bulletin 503, p. 130-134.

33. Strelow, R.A. and Baker, 'W.E., "The Characterization and Evaluation of Accidental Expl.osions", Report NASA CR 134779 for Aerospace Safety Research and Date Institute, Lewis Research Center, NASA, Cleveland, Ohio, June, 1975, NTIS 8N75-32191, (p.58).

34. Eichles, T.V. and Napadensky, "Accidental Vapor Phase Explosinns on Transportation Routes New Nuclear Power Plants", Report for Division of Engineering Standards, Office of Standards Develop~eat, XlC, May, 1978 NUREG/CR-0075 R5. (p. 40).

2. 2-18

SL2-FSAR SECTION 2.2: REFERENCES (Cont'd)

35. WASH - 1238 (1972) Environmental Survey of Transportation of Radio-active Materials to and from Nuclear Power Plants.

36. Arthur D. Little, Inc., "A Model Economic and Safety Analysis of Transportation of Hazardous Substances in Bulk", COM-74-11271, 1974.

37. Hazardous Materials Incident Reports from DOT F5800.1, July 1973 to December 1975. U.S. Department of Transportation, Materials Trans-portation Bureau, Office of Hazardous Materials Operations, Washington, D.C.

38. National Transportation Safety Board, Railroad Ace~dent Report, "Southern Railway Company, Tran 154, Derailment with Fire and Explosion, Laurel, Mississippi:, January 25, 1969", October 6, 1969.

39. ~

NUREG-0570, "Toxic Vapor Concentrations in the Control Room Follow-ing A Postulated Accidental Release".

40. A Modal Economic and Safety Analysis of Transportation of Hazardous Substances in Bulk, Arthur D Little, Inc. Cambridge, Massachusetts (1979).

41. M Deputy of the Florida East Coast Railway, Personal Communication on August '17, 1979.

2. 2-19

SL2-FSAR TABLE 2.2-1 AVERAGE DAILY AUTO AND TRUCK TRAFFIC COUNTS DURING PEAK (FALL AND MINTER) SEASON 1977-1978 HIGHWAY NO. HIGHWAY SEGMENT TRAFFIC VOLUME Us 1 712 t.o Malt. on Road 19,535 SR 712 U S 1 t.o SR 70? 6,406 SR A1A Ft.. Pierce t.o Hart. in Co. 2,731 Sk 707 Malt. on Road t.o Mart. in Co. 2,072 SR 712 t.o WalLon Road 1, 016 Walt. on Road NA Easy St.rect SOURCE: St.. Lucie Count y. Traffic Corridors Input. Dat.a, 1978.

NOTE: Separat.e truck count.s do not. exist. for t.he area within five miles of t.he plant..

SL2-FSAR TABLE 2.2-2 RESULTS OF TRUCK TkAF1 lC SURVEY Number uf Trucks Percentage Average Average Number of Classtfiea as Percentage v f Trucks Stat.on(') Daxly Weekday Trucks Carrying Ilazard- of Trucks vf Carrying Hazard Number Traffic( Traffic( Count.ed ous Hat,erial Weekda Traffic (2) uus Material (X) 6,895 7,262 129 3.1 210 6 1(3) 3.2 2 (5) (5) 3 (5) (5) 187 "

6 5 5(3) 3.2 4 2,511 2,574 50 1 1.9 2.0 5A 1,543 1,537 24 4 1.6 16.7 58 1,871 1,953 26 4 1.3 15.4 5C 1 >687 1,711 27 0 1.6 0 6A 5,163 5,289 340 6 4 2.4 68 (5) 859 43 4 5t4) 5.0 7 1 7,641 19,170 1,187 32 6.2 2.7 SA68 1,305 1,380 . 19 0 9 3,035 3,418 (6) 36) (6) (6)

Not,es:

(1) See Figure 2. 2-2 fur locattvns vf stations.

All count.s are non-directional.

Based on Average Weekday Trafftc recorded aL St,ation Hv. 9.

Based on Average Weekday Traffic recurded at. Stetson Hv. 7A.

As indicatea on Figure 2.2-2 - aul.omat,ic traffic recordings were made at. this locat.ion.

As indicat.ed vn Figure 2.2-2 classification counLs and roadside int.erviews were not maae aL t.his local>un.

Source: ChamPagne Associates, Vehicle Classificat.iun and Product, Containment. St,ud, February, 1979.

3 SL2-FSAR TABLE 2.2-3 HAZARDOUS MATERIALS FROM TRUCKS INTEkVIEWED ON US 1 JANUARY 30 AND 31 1979

~rrn ene Tot.al 5 Vehicles Maximum size (gallons) 2604 Average size (gallons) 1400 Gasoline Tot.al 18 Veht.cles Maximum size (gallons) 8300 Average size (gallons) 3100 Non-Flammable Gas:

(A)'ubal 1 Vehicle Size (ft. ) 12000 (B) Tot.al 1 Vehicle SIre (ibm) 230 (C) Tot.al size (gallons) 1 Vehicle Size (gallons) 100

~Ox en Tot.al 1 Vehicle Size (ft. ) 10,000 Bat.t.eries Total 2 Vehicles Maximum t,ransport.ed 400 Average Lransporbed 300 Diesel Oil Tot.al 9 Vehicles Maximum size (gallons) 8100 Average size (gallons) 1260 Bot.lied Gas Tot,al 1 Vehicle size (gallons) 50 Ch i ri enn Tot.al 2 I/2 Vehicles Max >mum ( gal ion s) 2400 Average (gallons) 970

O SLR-F TABLE 2.2-3 (Cont.'d)

Muriat.ic Acid Tot.al 1/2 Vehicle Size (gallons) 500 Combust.ible To t.al 1 Vehicle Size (t.ot.al) 6000 Corrosives Total 1 Vehicle Size (gallons) 3600 Tot.al Number of Vehicles:

Source: Champagne Associal.es, Vehicle Classificat.ion and Product.

Cont.ainment. St.ud , March, 1979

SL2-FSAR TAttLE 2.2-4 TRUCK DELIVERIES (CGHPRESSEO CASES PROCESS CHEhlCALS) TO ST LUCIE UNITS 1 AND 2 SHIPMENT SHIPhENT QUANTITY IIATEllIAL ~PRE IIEIICY HETliOD SHIPPED Acct.ylene weekly 5 Lon open Lruck 5-10 cylinders Oxygen weekly 5 t.on open Lruck 5-10 cylinders CO 2

semi-annually 5 Lvn vpen truck 80 cylinders N - t.razlers bi-monLhly Tube t.railer 40,000 scf ea. load 2

N, bvLt.les 3 t.imes/year 5 Lvn open Lruck 20-30 boLt.les 2

Irqurd monthly Liquru t.anker N N 1100 gal ~

Argon 3 Ltmes/year 5 t.on open t,ruck 1-2 cylinders HeLhane 3 t.imes/year 5 t.on open t.ruck 1-2 cylinders Propane 3 t.tmes/year 5 Lvn open t.ruck 1-2 cyltnders SpecraILy gases 3 t.imes/year 5 t.on open Lruck 1-2 cylinders Cyclohexylamine semi-annually Closed semi-t,railer 2 drums Ammoriium Hydroxide semi-annually Closed semi-Lrailer 2 drums Hydrazine bi-mont.h 1 y Closed semi-Lrarler 6 drums PoLassium t)ichromaLe semi-annually UPS 200 lbs Sodium Hydrvxide .monLhly Tank Lruck 3,800 gallons Sulfurrc Acrd mvnLhly Tank t.ruck 3,000 gallons Chlorrne mont.hly Open t,racLor t.railer 4 Lons Hydrogen-Lrailer b i-mont.h 1y .Tube Lrailer 75,000 scf Hydrogen-boLL les b i-mon Lh ly 5 t.on open t,ruck 20-30 bvLLles

Qy SL2-FSAR TABLE 2.2-5 TRUCK ACClDENTS WITNIN FIVE MILES OF ST LUCIE UNIT 2 1973-1977 Number Type Amounl of of Type of DaLe of Vehicles Truck of Properly Locaeion

• Acciaenc Involved Involved ~naca e ~na a e A 1973 2 SU PDO $ 10,000 B 1973 1 T-T I $ 2,000 C 1973 2 SU PDO $ 150 D 1973 2 T-T I $ 2>700 E 1973 2 SU PDO $ 150 1973 1 SU I $ 2,100 0 1974 3 T-T PDO 300 N 1974 2 T-T F $ 1,900 I 1974 2 SU PDO $ 175 J 1974 2 SU I $ 2,300 K 1977 2 T-T PDO $ 900 L 1977 3 SU PDO $ 2,500 M 1977 2 T-T PDO 800 N 1977 2 SU PDO $ 350 0 1976 2 SU PDO $ 300 P 1973 2 T-T PDO $ 800 1973 2 SU PDO $ 1,600 k 1974 1 SU PDO $ 100 S 1976 2 T-T I $ 9,000 T 1973 1 SU PDO $ 5,000 U 1977 2 SU PDO $ 900

• See Figure 2.2-4.

• • No accident. involved more Lhan one Lruck.

Legend:

SU Single Unit. Truck Source: Champagne Associales, Vehicle ClassificaLion T-T Tract. or Trailer and Producl Conlainmenl St.ud, March, 1979 I Injury F Falalily PDO Properly Damage Only

SL2- FSAR TABLE 2.2-6 FLUKIDA PUWEK & LICH'1'OMPANY CUhhODITY hUVEHENTS - 1975, 1976, & 1977.

DOMESTIC (ONLY) WATEKBOKNE CUHHEKCE PASSING THE APPKOXlhATE LOCATIOtt OF Tttb ST LUCIE UNI1' Short Tons per Calendar Year T e of Cvmmvdix '1 e vl Vessel 1975 1976 '977 Tot.als Shxps-and Boats t'assenger & Dry Cargu- 74B. 0 4, 030.0 12 > tt32.0 17,61U.U Self-propelled 1'resh 1'ish, except. Shellfish 15.0 22.0 37. 0 Ice 11.0 lb.u Misc Products uf Hanutacxuring 5U.U Furnxlure and Fxxt.ures Dry Cargu Nun-Seit- 28. 0 28. 0 propelled Hisc Wvn-mecal lie hinerai Prua. 19,615.0 9,372.0 7,203.U 36,190.U Irvn and St.eel, Bars, Rods, An- 4BU.O 75.0 65U.U 1205. 0 gles, Shapes and Sect.ions, In-cluding Sheex t'iling Iron and St.eel Pxpe and Tube 1>395.0 300.0 - 1,695.0 Fabricat.ed Het.al Pruducls ex- 3,801.0 2,940.U 5,225.0 11,966.0 cept. Ordnance, Hachxnery, and Transpur t.at.i vn Equi pmenx Hachi.nery excepx Elect.rxcal 6, 753.0 2, 955. 0 = 7, 056 ~ 0 16, 764. 0 Ele .t ricul Hachinery, Cquip- 1,190.0 3,050.0 1,610.0 "

5>850.0 ment. and Supplies Axrcratl and Paris 33.0 2BO. 0 ~ 313.0 Ships ana Bust.s 2B.U 275.0 303 ~ 0 Misc Shxpment.s nut. Ident.i.fiable 300. 0 615.0 2,255.0 3,170.0 by .Cvmmudxty Sugar 2,700.0 29,050.0 31,750.0 Aluminum and Aluminum Al luys, 4,000.0 5,UOO.O 9,000.0 UnworKed

SL2-FSAR TAHLE 2.2-6 (ConL'd) r ShorL Tons per Calendar Year T e vf CommodiL T e of Vessel 1975 1976 1977 Tot.als Sasic TexLile ProducLs, ex- Dry Cargo - Hon-Self- 50. 0 50. 0 cept. Text.ile Fibers propelled Timber, Post.s, Poles, Piling, 100. 0 100. 0 and oLher Wood in Lhe Rough Iron and SLeel Scrap 2>170.0 2,170.0 Sodium Hydroxide- Tanker " Hon-Self- 3,987.0 4,046.0 4,098 ' ,L2 i131. 0 (Caust. ic Soda) propelled Gasoline, including HaLural 100.0 100. 0 Gas Kesidual Fuel Oil 48,273.0 -

91,269.0 57,251.0 196,793.0 Asphalb> Tar, and PiLches 2,812.0 2,812.0 TOTALS 86,731.0 127,890.0 135,482.0 350t103.0 Source: DeparLment. of t.he Army, Lover Mississippi Valley Division Corps of Engineers, WaLerbvrne Commerce SLat.ist,ics Cent.er, January 10, 1979.

SL2-FSAR TABLE 2. 2-7 AIRPORTS WITHIN 9-50 MILES OF ST LUCIE UNIT 2 DISTANCE AHD DIRECTION PROM AIRFIELD SITE (STATUTE hILES)

Civil-Public Use Valkaria 45 NHH Sebasbian 34 HNff Vero Beach 22 UN'2 SL Lucie Co. NQ Qibham 13 S Palm Beach Gardens 38 S Palm Beach Internabional 48 SSE Okeechobee Co.= 37 Q Circle T Ranch 25 SSQ Palm Beach Co., Glades 48 SH.

Privare Fellsmere 36 NW Broocke 27 NNM New Hibiscus 25 HHH Indian River 22 QSf Nelson 1) HQ Raw@wild ll NH Peacock Ranch 13 QSQ Naked Lady Ranch 16 SSH Tropical Planrabion, 18 S Chem 42 SSQ Evans 21 SQ

)

SL2-FSAR TABLE 2. 2-7 (Cont.')

Mulgrew Ranch 40 RSVP Sunset. 41 4 Inaian Itammock 38 MNM Sunrise 9 NN'M Palm Beach Ranch Groves 37 S

~Beli orLs Sikorsky (Privat.e) 32 SSH SOURCE: Unit.ed St.at.es Department. of Commerce, Nat.ional Oceanic and Atmospheric Administ.rat.ion. Sect.ional Aeronaut. ical Chart., Miami, Sept. ember 1978.

• To nearest. mile, measured from Sect.ional Aeronaut. ical Chart.,

Miami.

Jgg SL2-FSAR TABLE 2.2-8 TOXIC CHEHICAL EVAMATION Distance From Hain Peak Concentration (ppm)

Source and Type Toxicity Limit/(Re f) Control Room OAI of Inside of Toxic Chem ( m) Control Rm Control Rm Remarks Acetylene 600 360 SCF Note 4 Ammonium 500/(4) 260 NW" 55 gal; 30 4.84 x 10 105 Note 2 Hydroxide by weight Carbon Dioxide 1.0 x 10 /(3) 600 NW 360 SCF 2.08 x 10 Note 1 Cyc lohexylamine 20/(6) 180 55 gal; 100Z conc 1.61 x 10 Note 10 Hydrazine- 5/(5) 180 400 gal; 35Z conc Note 9 Amerzinc Chem by weight Feed System Hydrazine- 5/(5) 60 550 gal; 5Z conc Note 9 Iodine Removal by weight Hydrogen 640 260 Notes 4,5 SCF'60 N

2

-gas 590 SCF Note 5 N

2

-liquid 620 NW 1100 gal Note 5 Potassium 60 100 gal Notes 3,6 Dichromate-.

TCCWS Potassium 220 ENE 50 gal Notes 3,6 Dichromate-CCWS Sodium 690 10,000 gal; 50Z Notes 3,6 Hydroxide conc by weight Sulfuric Acid 3 0

~m/(5) 700 10,000 gal; 60 Note 3 3 Baume m

• Concentration inside control room not determined since concentration at outside air intake is below toxicity limit

SL2-FSAR TABLE 2.2-8 (Cont'd)

Distance From Main Peak Concentration (ppm)

Source and Type Toxicity Limit/(Ref) Control Room OAI of Inside of Toxic Chem ( m) (Feet) (Direction) Quantit Released Control Rm Control Rm Remarks Road:

Chlorine 15/(7) 4.8 miles WSW 2,430 gal(2.8x10 lbs) Note 7 Combustibles 4.8 miles MSW 6,000 gal Note 4 Rail:

Chlorine 15/(7) 2.0 miles 90 tons Note 8 Liquid Petroleum 2.0 miles 33,000 gal Notes 4,5 Gas River:

Gasoline 1.2 miles M 16,000 bbl Notes 4,5 Sodium Hydroxide 1.2 miles W 16,000 bbl Notes 3,6 Notes

1) Not a design basis event because the calculated concentration at the outside air intake of the control room is less than the toxicity limit.

2) The concentration at the outside air intake of the control room calculated assuming all the toxic chemicals in the solution becomes airborne instantaneously following the rupture of the container.

3) Not volatile

4) 'rimarily a fire hazard

5) Simple asphyxiant. Toxic effect occurs at 33 percent volume in air.

6) Toxic agent is a solid under ambient conditions.

7) Quantity of toxic chemical, at the given distance, is lese than the maximum specified in Regulatory Guide 1.78 (RO), Table C-2.

1

8) Not a design basis event. The probability of occurrence, as indicated in Section 2.2.3 is lese than 10 per year.

9) The partial vapor pressure of the toxic chemical in the solution is lees than 10 mm Hg. The event eliminated based on guidelines provided in Regulatory Guide 1.78 (RO).

10) Concentration at the outside air intakes based on guidelines provided in Section 2.2 Reference (1).

SL2" FSAR TABLE 2.2-8 (Cont.'d) for Toxicit. Limit.s 'eferences

1) Crit.eria for a Recommended St.andard - Occu at.ional Ex osure t.o

~Ace!. leee, DHEH/PBU/HZOBH-76/195.

2) Sax, N Irving. Dan erous Pro ert.ies of Indust,rial Mat.erials, Third Edit.ion, Reinhold Book Corp., New York, -1968.

3) Regulat.ory Guide 1.,78, "Assum t.ions for Evaluat.in'he 1tabit.abilit.

of Nuclear Power Plant. Control Room Durin a Post.ulat.ed Hazardous Chemical Release."

4) Crit.eria for a Recommended St.andard - Occu at.ional Ex osure DHEW/PUB/NIOSH 74-136. NTIS-PB-246 t.o'mmonia.

699.')

Pat.t.y, Frank, A. Indust.rial H iene and Toxicit:, Vol II Toxicit.y (2nd Edit.ion Revised), Interscience Publishing .Co. New York, 1963.

6) Karel Verschueren. Handbook of Environment.al Dat,a on Or anic Chemicals, Van Nost.rand Hheinhold Company, New York.

7) Criteria for a Recommended St.andard Occu at.ional.Ex osure t.o Chlorine. DHEW/PUQ/NIOSH 76-170.

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