ML20195J228

Rept for Boron Concentration Reduction/Boron Injection Tank Elimination

ML20195J228
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Site:	Zion  File:ZionSolutions icon.png
Issue date:	10/31/1985
From:	Bass J COMMONWEALTH EDISON CO.
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Text

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m REPORT FOR THE BORON CONCENTRATION REDUCTION /

BORON INJECTION TANK ELIMINATION FOR ZION UNITS 1 AND 2 J. C. Bass October 1985

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INTRODUCTION Westinghouse has developed improved analytical techniques which allow a reduction in the Boron injection Tank (BIT) concentration and, in most cases, removal of the BIT. This report provides background inf ormation on the BIT design basis, reasons why boron reduction or tank removal may be desirable, plant design features which allow the change to be proposed, as well as a sunrnary of analytical results which demonstrate the feasibility of boron concentration reduction for Zion Units 1 and 2.

BACKGROUND The BIT is a component of the Safety injection System whose sole function is to provide concentrated boric acid to the reactor coolant to mitigate the consequences of postulated steamline break accidents. Although the BIT acts to mitigate steamline breaks of various sizes occurring f rom any power level, the cases which serve as the Westinghouse steam 14.ne break licensing basis, and which define the existing requirements on the minimum BIT boron concentration, are as follows:

For the ' hypothetical' steamline break, i.e., double ended rupture of a main steamline, the radiation releases must remain within the requirements of 10CFR Part 100. This is the ANSI N18.2 criterion for Condition IV events, ' Limiting Faults.' Westinghouse conservatively meets this for the Zion Units by demonstrating that the DNB design basis is met, the i

criterion typically used for Condition 11 events.

For the ' credible' steamline break, i.e., the spurious opening of a single steam generator relief, saf ety, or turbine bypass valve, the radiation releases must remain within the requirements of 10CFR Part 20. This is the ANSI N18.2 criterion for Condition 11 events, "Faults of Moderate Frequency." Westinghouse has conservatively met this criterion in the past, by showing that no return to criticality is achieved, although showing that the DNB design basis is met is traditionally sufficient.

91770:10/102185 1

1 In order to assure the validity of the safety analyses performed to verify that the Westinghouse criteria are met, technical specifications are applied to the BIT and associated equipment. Specifically, these assure that the bo"ic acid concentration is maintained in excess of 20,000 ppm, approximately a 12 weight percent solution.

Heat tracing is necessary to maintain the tank and associated piping at a sufficiently high temperature so that the minimum concentration requirements may be met. Furthermore, the safety-related nature of the boric acid system requires that the heating systems be redundant.

The required solubility temperature imposes a continuous load on the heaters, and low-temperature alarm actuation and heater burnout have occurred in some operating plants. Violation of the Technical Specification on concentration in the BIT poses availability problems in that recovery is required within a very short time. If the concentration is not restored within one hour, the plant must be taken to the hot standby condition and borated to the equivalent i

of 2 percent 8K/K at 200'F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Thus, this requirement has a potentially serious impact on plant availability. In addition, the high boric acid concentration makes recovery from a spurious safety injection signal (which results in injection of the BIT fluid into the reactor coolant system) tine consuming and costly.

These potential difficulties unfavorably affecting plant availability, operability, and maintainability can be drastically reduced in severity or eliminated by reducing the boron concentration to a minimum level at which heat tracing would no longer be required or, where possible, by entirely-deleting the BIT system. The effect of these changes are discussed in the following section.

DESCRIPTION OF THE ANALYSES Accident Analysis The only accident analyses which are affected by boron reduction or BIT removal are the steamline break transients. These transients are affected with respect to core integrity.

For the Zion Units, the analysis was performed assuming that the BIT remains installed, and boric acid 91710:10/102185 2

concentration is reduced to O ppm. This is the conservative configuration.

This case is considered because, if feasible, the absence of boric acid in the BIT will alleviate potential degradation due to stress corrosion, eliminate baron plateout and line plugging concerns, and eliminate concerns associated with recirculating boric acid between the BIT isolation valves. Furthermore, recovery time both for the RCS (dilution) and the BIT (reconstitution) following an inadvertent actuation of the Safety Injection System will be significantly reduced, which could improve overall plant availability.

In addition, all Technical Specifications concerning BIT concentrations, temperatures, and associated surveillance can be eliminated, exclusive of the refueling water storage tank.

Savings associated with not maintaining the heat tracing would also be realized.

The following cases, analyzed for the Zion Final Safety Analysis Report, assumed a BIT concentration of 0 ppm:

' Hypothetical" Steamline Break, with and without offsite power available, for the largest double ended rupture of a steam pipe. This is analyzed for both the main steam pipe rupture outside containment and the pipe rupture at the steam generator exit.

' Credible' Steamline Break, with of f site power available, f or the largest single failed open steam generator relief, safety or steam dump valve.

Criteria For the hypothetical break, the same criteria is applied to the BIT l

alternative analyses as is applied in the FSAR. That is, for the most severe Condition IV break, Westinghouse will show that the radiation releases are within the requirements of 10CFR Part 100 by demonstrating that the ONB design basis is met. The steamline break dose calculations performed for the FSAR use a conservative fuel f ailure level of 5 percent although the core analyses show that no consequential fuel f ailures are anticipated.

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91770:10/102185 3

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The credible steamline break analysis was performed using a new criterion whereby the plant may return to criticality but no damage may occur to the fuel. This constitutes a rglaxation of the conservative internal Westinghouse

'- criterion for Class II Yventsj3' Th{s new criterion is in compliance with the criteria used by the NRC and ANS, which require that releases during steamline break accidents remain within the limits set forth in 10CFR Part 20.

This limit could be met with a return to criticality if it is assured that there is no consequential fuel damage.

Analysis Methods In the Zion steamline break analysis, the system transient parametcrs, i.e.,

RCS pressure, temperatures, steam flow, core boron concentration and core power are calculated using the LOFTRAN (1) system transient analysis computer code, instead of MARVEL, which was used f or the previous FSAR analysis. This computer code includes models of the reactor core, steam generators, pressurizer, primary piping, protection systems and engineered saf eguards systems. The changes in safety injection system volumes, initial concentrations, and temperatures corresponding to the O ppm BIT are introduced into the analyses in the LOFTRAN code.

RESULTS Figures I and 2 depict the 4.6 square feet double ended rupture at steam generator exit with offsite power available, This is the largest double ended break that may be postulated to occur. The plant is initially assumed to be at hot zero power at the minimum required shutdown margin.

1.

WCAP-7907, T.W.T. Burnett, et. al., 'LOFTRAN Code Description,'

October,1972.

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Following the break, the RCS temperatures and pressures decrease rapidly, and in the presence of a large end of lif e (EOL) moderator coef ficient of reactivity, the reactor returns critical with the rods inserted, assuming the most reactive RCCA in the fully withdrawn position. Thereactorpowerincreasesatadecreasingrateuntilboron').,,..pgs f rom the safety injection system reaches the core and begins to of f set the positive reactivity insertion caused by the cooldown. The core is subsequently brought subtritical with boron injection, aided by the abatement and eventual termination of steam flow f rom the broken steam generator.

Figures 3 and 4 show the similar transient assuming no of f site power available. The reactor coolant pump coastdown reduces the thermal coupling between the secondary side and the primary side, reducing the cooldown experienced by the core.

The additional delay f or starting the

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emergency diesel generators clso accounts for a slightly longer time e

P delay'to deliver the boron to the core.

,M-The breaks postulated outside containment do not produce as severe a return to power as the cases presented above.

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For these cases, assuming a concentration of 0 ppm in the Bli, a DNB analysis was performed at selected time points to verify that the DNB design basis is met.

Power distributions were calculated using a multidimensional neutron dif f usion code and the thernal-hydraulic local conditions including DNB Ratio were calculated using a thermal-hydraulic computer code.

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I 91770:10/102185 5

Although preventing clad damage is not necessary for Condition IV events, the analysis shows that the DNB design basis is met, i.e., the DNB Ratio remains greater than the limit value. The dose evaluation, which is perfomed assuming 5 percent f ailed f uel, theref ore continues to demonstrate that the Condition IV accident criteria are satisfied.

Figures 5 and 6 depict transient parameters for the Condition II steamline break assuming the BIT at 0 ppm. For the previous FSAR case, the reactivity plot in Figure 14.2.5-5 shows that the reactor remains subtritical. This Westinghouse criterion assures that the DNB design basis is met in a very conservative manner. 4n order to allow substantial reduction in BIT boron concentration or elimination of the BIT, this Westinghouse criterion is relaxed to allow a subsequent return to power for the Condition II transients, but the DNB design basis must be met in order to meet the 10CFR Part 20 dose requirements. Figure 6 shows that criticality is attair.ed and sustained when the BIT is at 0 ppm. 'DNB analyses for this case shows that the DNB design basis is met and no fuel failures are predicted. This conclusion is also consistent with the conclusion drawn on the Condition IV breaks, since no violation of the DNB design basis was calculated for the more extreme Condition IV, double ended ruptures.

Conclusions In conclusion, calculations have been perfomed for Zion Units 1 and 2 to show that Commonwealth Edison may reduce the boron concentration in the BIT to_0.. ppm or remove the BIT entirely. This will meet the safety criteria for Condition II and Condition IV events analyzed in Chapter 14 of the FSAR.

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REVISED FINAL SAFETY ANALYSIS REPORT (FSAR) SECTION O

e.WW 91710:10/100885 13

14.2.5 DEPRESSURIZATION OF THE MAIN STEAM' SYSTEM 14.2.5.1 ACCIDENTAL DEPRESSURI2ATION OF THE MAIN STEAM SYSTEM Identification of Causes and Accident Description The most severe core conditions resulting from an accidental depressurization of the main steam system are associated with an inadvertent opening of a single steam dump, relief, or safety valve. The analyses performed assuming a rupture of a main steamline are given in Section 14.2.5.3.

The steam release as a consequence of this accident results in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the RCS causes a reduction of coolant temperature and pressure, in the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity.

The analysis is perforned to demonstrate that the following criterion is satisfied:

Assuming a stuck rod cluster control assembly, with of f site power available, and assuming a single failure in the Engineered Safety Features System there will be no consequential damage to the core or the reactor coolant system after reactor trip for a steam release equivalent to the spurious opening, with f ailure to close, of the largest of any single steam dump, relief, or safety valve.

The following systems provide the necessary protection against an accidental depressurization of the main steam system.

1.

Safety injection System actuation f rom any of the following:

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a.

Two-out-of-three coincident low pressurizer pressure signals b.

Two-out-of-three dif f erential pressure signals between a steamline and the remaining steamlines.

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91650:10/100385 l

2.

The overpower reactor trips (neutron flux and AT) and the reactor trip occurring in conjunction with receipt of the safety injection signal.

3.

Redundant isolation of the main feedwater lines.

Sustained high feedwater flow would cause additional cooldown. Therefore, in addition to the nornal control action which will close the main feedwater valves following reactor trip, a safety injection signal will rapidly close all feedwater control valves and feedwater isolation valves, trip the main f eedwater pumps, and close the f eedwater pump discharge valves.

Accidental depressurization of the secondary system is classified as an ANS Condition 11 event; a f ault of moderate f requency.

I Analysis of Effects and Consecuences Method of Analysis The following analyses of a secondary system steam release are perf orned for this section.

1.

A full piant digital computer simulation using the LOFTRAN Code (Reference

3) to determine RCS temperature and pressure during the transient, and the effect of safety injection.

2.

Analyses to determine that the DNB basis is met.

The following conditions are assumed to exist at the time of a secondary steam system release:

1.

End-of-lif e shutdown margin at no-load, equilibrium xenon conditions, and with the most reactive rod cluster control assembly stuck in its fully withdrawn position. Operation of rod cluster control assembly banks during core burnup is restricted in such a way that addition of positive reactivity in a secondary system steam release accident will not lead to a more adversy condition than the case analyzed, ne. voc4 cAwe PCL.

l 91650:10/100285

2.

A negative moderator coefficient corresponding to the end-of-life rodded core with the most reactive rod cluster control assembly in the fully withdrawn position. The variation of the coefficient with temperature and pressure is included. The K,77 versus temperature at 1000 psi corresponding to the negative moderator temperature coefficient used is shown in Figure 14.2.5-1.

3.

Minimum capability for injection of boric acid solution corresponding to the most restrictive single failure in the Safety injection System. This corresponds to the flow delivered by one charging pump delivering its full contents to the cold leg header. No credit is taken for the low concentration boric acid which must be swept from the safety injection lines downstream of the refueling water storage tank prior to the delivery of boric acid (2000 par.ts per million (ppm)) to the reactor coolant loops.

4.

The case studied is a steam flow of 247 pounds per second at 1100 pounds per square inch absolute (psia) with offsite power available. This is the maximum capacity of any single steam dump, relief, or safety valve.

Initial hot shutdown conditions at time zero are assumed since this represents the most conservative initial condition.

Should the reactor be just critical or operating at power at the time of a steam release, the reactor will be tripped by the normal overpower protection when power level reaches a trip point. Following a trip at power, the RCS contains more storer! energy than at no-load, the average coolant temperature is higher than at no-load and there is appreciable energy stored in the fuel. Thus, the additional stored energy is removed via the cooldown caused by the steam release before the no-load conditions of RCS temperature and shutdown margin assumed in the analyses are reached. After the additional stored energy has been removed, the cooldown and reactivity insertions proceed in the same manner as in the f

analysis which assumes no-load condition at time zero. However, since the

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initial steam generator water inventory is greatest at no-load, the magnitude and duration of the RCS cooldown are less for a steam release occurring at power.

91650:10/100285

5.

In computing the steam flow, the Moody Curve (Reference 2) for f(L/0) = 0 is used.

6.

Perfect moisture separation in the steam generator is assumed.

1.

The design value of the steam generator heat transfer coef ficient was used.

Parametric analysis indicates that variation of the coefficient over a range of 130% has little effect on the transient.

Results The calculated time sequence of events for this accident is listed in Table 14.2.5-1.

The results presented are a conservative indication of the events which would occur assuming a secondary system steam release, since it is postulated that all of the conditions described above occur simultaneously.

Figures 14.2.5-3 and 14.2.5-4 show the transiert results f or a steam flow of

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247 lb/see at 1100 psia.

The assumed steam release is the maximum capacity of any single steam dump, relief, or saf ety valve. Saf ety injection is initiated automatically by high steamline differential pressure. Operation of one centrifugal charging pump is assumed. Boron solution at 2000 ppm enters the RCS f rom the RWST providing sufficient negative reactivity to prevent core damage..The transient is quite conservative with respect to cooldown, since no credit is taken for the energy stored in the system metal other than that of the fuel elements or the energy stored in the other steam generators. Since the transient occurs over a period of about 5 minutes, the neglected stored energy is likely to have a significant effect in slowing the cooldown.

Conclusions The analysis shows that the criteria stated earlier in this section are satisfied.

For an accidental depressurization of the main steam system, the 91650:10/102185

i DNB design limits are not exceeded. This case is less limiting than the steamline rupture case described in Section 14.2.5.3.

14.2.5.2 MINOR SECONDARY SYSTEM PTPE BREAKS Identification of Causes and Accident Description

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Included in this grouping are ruptures of secondary system lines which would result in steam release rates equivalent to a 6-inch diameter break or smaller.

sis of Effects and Conseauences Minor secondary system pipe breaks must be accomodated with the f ailure of only a small f raction of th'e f uel elements in the reactor. Since the results of analysis presented in.Section 14.2.5.3 for a major secondary system pipe l

rupture also meet this criteria, a separate analysis for minor secondary c

system pipe bre,aks is not required.

The analysis of the more probable accidental opening of a secondary system steam dump, relief or safety valve is presented in Section 14.2.5.1.

This analysis is illustrative of a pipe break equivalent in size to a single valve opening.

Conclusions The analysis presented in Section 14.2.5.3 demonstrate that the consequences of a minor secondary system pipe break are acceptable since a DNBR of less than the limit value does not occur even for a more critical major secondary system pipe break.

14.2.5.3 MAJOR SECONDARY SYSTEM PIPE RUPTURE Ruoture of a Main Steamline Identification of Causes and Accident Description The steam release arising f rom a rupture of a main steamline would result in an initial increase in steam flow which decreases during the accident as the i

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91650:10/102185

steam pressure falls. The energy removal from the RCS causes a reduction of coolant temperature and pressure, in the presence of a negative umderator temperature coefficient, the cooldown results in an insertion of posit've reactivity. If the most reactive rod cluster control assembly (RCCA) is assumed stuck in its fully withdrawn position after reactor trip, there is an increased possibility that the core will become critical and return to power.

A return to power following a steamline rupture is a potential problem mainly because of the high power peaking factors which exist assuming the most reactive RCCA to be stuck in its fully withdrawn position. The core is ultimately shut down by the boric acid injection delivered by the Safety injection System.

The analysis of a main steamline rupture is performed to demonstrate that the following criteria are satisfied:

~

Assuming a stuck RCCA with or without offsite power, and assuming a single failure in,the engineered safety features, the core remains in place and intact.

~

Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable, the following analysis, in fact, shows that no DNB occurs for any rupture assuming the most reactive assembly stuck in its fully withdrawn position.

A major steamline rupture is classified as an ANS Condition IV event.

The najor rupture of a steamline is the most limiting cooldown transient and is analyzed at zero power with no decay heat. becay heat would retard the cooldown thereby reducing the return to power. A detailed analysis of this 1

l transient with the most limiting break size, a double-ended rupture, is presented here.

The following functions provide the protection for a steamline rupture:

1.

Saf ety Injection System actuation f rom any of the following:

i 91650:10/100385

a.

Two-out-of-three differential pressure signals between a steamline and the remaining steamlines, b.

High steamline flow in two main steamlines (one-out-of-two per line) in coincidence with either low-low reactor coolant system average temperature or low steamline pressure in any two lines, c.

Two-out-of-four high containment pressure.

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2.

The overpower reactor trips (neutron flux and AT) and the reactor trip occurring in conjunction with receipt of the safety injection signal, 3.

Redundant isolation of the main feedwater lines.

Sustained high feedwater flow would cause additional cooldown. Therefore,

~

in addition to the normal control action which will close the main feed-water valves, a safety injection signal will rapidly close all feedwater control valves and feedwater isolation valves, trip the main feedwater pumps, and close the feedwater pump discharge valves.

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4. '~ Trip of tee ~ f a'st ~atting~steamline stop valves-(designed to T1ose-in-less than 5 seconds) on:

a.

High steam flow in two main steamlines in coincidence with either low-low reactor coolant system average temperature or low steamline pressure in any two lines.

b.

High-high containment pressure.

Fast-acting isolation valves are provided in each steamline; these valves will fully close within 10 seconds of a large break in the steamline. For breaks

~

downstream of the isolation valves, closure of all valves would completely terminate the blowdown. For any break, in any location, no more than one steam generator would experience an uncontrolled blowdown even if one of the isolation valves fails to close.

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i 91650:10/101185 i

Nozzles inside the steam pipes which are of considerably smaller diameter (16 inches) than the main steam pipe are located inside the containment and serve to limit the maximum steam flow f or any break f urther downstream.

Analysis of Effects and Consecuences Method of Analysis The analysis of the steam pipe rupture has been performed to determine:

1.

The core heat flux and RCS temperature and pressure resulting from the cooldom following the steamline break. The LOFTRAN code (Reference 3) has been used.

2.

The thermal and hydraulic behavior of the core following a steamline break. A detailed thermal and hydraulic digital-computer code, THINC, has been used to determine if DNB occurs for the core conditions computed in p

item 1 above.

The following conditions were assumed to exist at the time of a main steam break accident:

1.

End-of-life shutdown margin at no-load, equilibrium xenon conditions, and the most reactive RCCA stuck in its fully withdrawn position. Operation of the control rod banks during core burnup is restricted in such a way that addition of positive reactivity in a steamline break accident will not lead to a more adverse condition than the case analyzed.

2.

A negative moderator coef ficient corresponding to the end-of-lif e rodded core with the most reactive RCCA in the fully withdrawn position. The variation of the coef ficient with temperature and pressure has been

~

included. The K versus temperature at 1000 psi corresponding to the g7 negative moderator temperature coefficient used is shown in Figure 14.2.5-1.

The ef fect of power generation in the core on overall reactivity is shown in Figure 14.2.5-5.

91650:10/100385

The core properties associated with the sector nearest the affected stean generator and those associated with the remaining sector were conservatively combined to obtain average core properties for reactivity feedback calculations. Further, it was conservatively assumed that the core power distribution was unifom. These two condicions cause underprediction of the reactivity feedback in the high power region near the stuck rod. To verify the conservatism of this method, the reactivity as well as the power distribution was checked for the limiting statepoints

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for the cases analyzed.

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This core analysis considered the Doppler reactivity from the high fuel m '-/' '

temperature near the stuck RCCA, moderator f eedback f rom the high water enthalpy near the stuck RCCA, power redistribution and non-uniform core inlet temperature effects. For cases in which steam generation occurs in the high flux regions of the core, the effect of void fomation was also included.

It was determined that the reactivity employed in the kinetics analysis was always larger than the reactivity calculated including the above local effects for the statepoints. These results verify conservatism; i.e., underprediction of negative reactivity f eedback f rom power generation.

3.

Minimum capability for injection of boric acid (2000 ppm from the RWST) solution corresponding to the most restrictive single failure in the Safety injection System. The Emergency Core Cooling System consists of three systems: 1) the passive accumulators, 2) the Residual Heat Removal System, and 3) the Saf ety Injection System. Only the Safety Injection System and the accumulators are modeled for the steamline break accident analysis.

The actual modeling of the Safety Injection System in LOFTRAN is described in Reference 3.

The flow corresponds to that delivered by one charging l

pump delivering its full flow to the cold leg header. No credit has been taken for the low concentration borated water, which must be swept f rom the lines downstream of the RWST prior to the delivery of high concentration boric acid to the Reactor Coolant loops.

91650:10/100385

For the cases where of f site power is assumed, the sequence of events in the Safety Injection System is the following. After the generation of the safety injection signal (appropriate delays for instrumentation, logic, and signal transport included), the appropriate valves begin to operate and the high head safety injection pump starts. In 10 seconds, the valves are assumed to be in their final position and the pump is assumed to be at full speed. The volume containing the low concentration borated water is swept;before the 2000 ppm reaches the core. This delay, described above, is inherently included in the modeling.

P' In cases where of f site power is not available, an additional 12 second t'

,c delay is assumed to start the diesels and to load the necessary safety injection equipment onto them.

4.

Four combinations of break sites and initial plant conditions have been

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considered in detemining the core power and reactor coolant system transients:

Complete severance of a pipe outside the containment, downstream of a.

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the steam flow measuring nozzle, with the plant initially at no 1oad

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conditions, f ull reactor coolant flow with of f site power available.

Completeseveranceofpipeinskecontainmentattheoutletofthe b.

steam generator with the plant initially at no load conditions with offsite power available.

Case (a) above with loss-of-offsite power simultaneous with the c.

initiation of the safety injection signal. Loss-of-offsite power results in coolant pump coastdown.

d.

Case (b) above with the loss-of-offsite power simultaneous with the initiatinn of the safety injection signal.

9 91650:10/100385

5.

Power peaking factors corresponding to one stuck RCCA and non-uniform core inlet coolant temperatures are determined at end of core life. The coldest core inlet temperatures are assumed to occur in the sector with the stuck rod. The power peaking factors account for the effect of the local void in the region of the stuck control asseebly during the return to power phase following the steamline break. This void in conjunction with the large negative moderator coef ficient partially of f sets the ef f ect of the stuck assembly. The power peaking f actors depend upon the core power, temperature, pressure, and flow, and, thus, are dif ferent f or each case studied. The core parameters used for each of the four cases correspond to values determined f rom the respective transient analysis.

All cases above assume initial hot shutdown conditions at time zero since this represents the most conservative initial condition. Should the reactor be just critical or operating at power at the time of a steamline break, the reactor will be tripped by the normal overpower protection system when power level reaches a trip point. Following a trip at power, the RCS contains more stored energy than at no-load, the average coolant 9

temperature is higher than at no-load.and there is appreciable energy stored in the fuel. Thus, the additional stored energy is removed via the cooldown caused by the steamline break before the no-load conditions of RCS temperature and shutdown margin assumed in the analyses are reached.

Af ter the additional stored energy has been removed, the cooldown and reactivity insertions proceed in the same manner as in the analysis which assumes no-load condition at time zero.

6.

In computing the steam flow during a steamline break, the Moody Curve (Reference 2) for f(L/0) = 0 is used.

7.

Perfect moisture separation in the steam generator is assumed. This assumption leads to conservative results since, in f act, considerable water wculd be discharged. Water carryover would reduce the nagnitude of

~

the temperature decrease in the core and the pressure increase in the containment.

91654:10/101185 4

1

Results The calculated sequence of events for both cases analyzed is shown on Table 14.2.5-2.

The results presented are a conservative indication of the events which would occur assuming a steamline rupture since it is postulated that all of the conditions described above occur simultaneously.

Core Power and Reactor Coolant System Transient Figures 14.2.5-6, and 14.2.5-7 shows the reactor coolant system transient and core heat flux following a main steam pipe rupture (complete severance of a pipe) outside the containment, downstream of the flow measuring nozzle at initial no load condition (case a).

The break assumed is the largest break which can occur anywhere outside the containment either upstream or downstream of the isolation valves. Offsite power is assumed available such that full reactor coolant flow exists. The transient shown assumes an uncontrolled steam release from only one steam

~

generator. Should the core be critical at near zero power when the rupture occurs, the initiation of safety injection by high differential pressure between any steamline and the remaining steamlines or by high steam flow signals in coincidence with either low-low reactor coolant system temperature or low steamline pressure will trip the reactor. Steam release f rom more than one steam generator will be prevented by automatic trip of the f ast action isolation valves in the steamlines by the high steam flow signals in coincidence with either low reactor coolant system average temperature or is limited to no more than 10 seconds for the other steam generators while the one generator blows down. The steamline isolation valves are designed to be fully closed in less than 7 seconds after receipt of closure signal with no flow through them. With the high flow existing during a steamline rupture, the valves will close considerably f aster. The steam flow on Figures 14.2.5-7, 9,11, and 13 represent steam flow f rom the f ault.ed steam generator only. In addition, all steam generators were assumed to discharge through the break for the first 10 seconds.

l 91650:10/101185

l The core attains criticality with the rod cluster control assemblies inserted (with the design shutdown assuming one stuck assembly) before boron solution at 2000 ppm enters the reactor coolant sy: tem f rom the safety injection system. The delay time consists of the time to receive and actuate the safety injection signal and the time to completely open valve trains in the safety injection lines. The safety injection pumps are then ready to deliver flow.

At this stage, a further delay time is incurred before 2000 ppm boron solution can be injected to the reactor coolant system due to low boron concentration solution being swept from the safety injection lines. A peak core power well below the nominal full power value is attained.

The calculation assumes the boric acid is mixed with, and diluted by, the water flowing in the RCS prior to entering the reactor core. The concentration after mixing depends upon the relative flow rates in the RCS, and f rom the Saf ety Injection System. The variation of mass flow rate in the

~

RCS due to water density changes is included in the calculation as is the variation of flow rate in the Safety Injection System due to changes in the RCS pressure. The Safety Injection System flow calculation includes the line losses in the system as well as the pump head curve.

Figures 14.2.5-8 and 14.2.5-9 show case b, a steamline rupture at the exit of a steam generator at no load. The sequence of events is similar to that described above for the rupture outside the containment except that criticality is attained earlier due to a more rapid cooldown and a higher peak core average power is attained.

Figures brough how the salient parameters for cases e + d, which correspond to the :ases discussed above with additional loss of of f site power at the time the saf ety injection signal is generated. The Saf ety Injection System delay tim; includes 12 seconds to start the diesel in addition to 10 seconds to start the safety injection pump and open the valves. Criticality is achieved later and the core power increase is slower than in the similar case with of f site power available. The ability of the emptying steam generator to extract heat f rom the RCS is reduced by the decreased flow in the RCS.

For both cases, the peak power remains well below the nominal full power value.

i l

l !

91650:10/100385

e ;-

f It should be noted that following a steaaline break only cne s:eam generator

.1< blows down completely. Thus, the remaining steam generators are still ~ .'lJ available for dissipation of decay heat after the initial transient is over. In the case of loss of of f site power this heat is removed to the atmosphere o'~- i._ via the steamline safety valves. n 4 Marcin to Critical Heat Flux A DNB analysis was performed for the three cases mast critical to DNB. It was found that all cases nave a minimum DNBR greater than the limit value. T-Conclusions T Although DNB and possible cladding perforation following a steam pipe rupture 7, is are not necessarily unacceptable and not precluded by the criteria, the above ~ analyses, in fact, show that the DNB design basis is met. f s we4 h. 't g4 s REFERENCF.S f e. &E ede d. //oW ~ ueb7 1. F. H. Moody, ' Maximum Flow Rate of Single Component, Two-Phase Mixture,' ~ ~~ Paper No. 64-HT-35, and ASME Publication. 2. J. M. Geets, ' MARVEL - A Digital Computer Code for Transients Analysis of a Multiloop PWR Systems,' WCAP-7909, June 1972. 'l 3. Burnett, T. W. T., et al., 'LOFTRAN Code Description,' WCAP-7907-P-A (Proprietary) WCAP-7907-A (Non-Proprietary), April 1984. l 3 i c 91650:10/100785

b W MST FGL NPS GET WEM IMJ June 10, 1988 SCH CMA KAA JAS Messrs: J. O'Connor D. Galle RJF W. Behnke G. P11m1 GtD B. Thomas R. Pleniewicz LEH C. Reed D. Bax PCZ N. Wandke J. Bitel 88R J. Abel R. Querlo TOSS E. Eenigenburg D. Scott H. Bliss / W. Showski L. De1 George W. Kalivianakis ~ ' ~ ~ ~ ~ ~ G. Diederich G. Wagner R. Flessner W. Worden A. Roberts R. Hennigan B. Shelton T. Maiman K. Graesser

Subject:

Department of Nuclear Safety Activities Report - May, 1988 This report includes a separate line chart of equipment failures (mechanical, electrical, and instrument f ailures). Also included are graphs of events classified by Offsite Review as personnel related errors. ~, ps .CN / J. S. Bitel ' 7(anager of Nuclear Safety JSB/AJS/emb Att. 0000m/0014m Page 1 l l t i

h.' Page 2 Department of Nuclear Safety Achivi_ ties Report - May. 1988 Part I Safety Issues - None Part II Itemt of Interest Reviewed During The Month Part III : Significant Events

Trend Analysis

Graphs Number of Personnel Related Deficiencies Percentage of Personnel Related Deficiencies / Total Number of Events per Month Flectrical Equipment Failures Instrument Equipment Failures Mechanical Equipment Failures Personnel Errors By Responsible Department Part IV :

Station Related Activities Part V Special Activities

page 3 Department of Nuclear Safety Activities Report - May, 1988 part 1 - Safety Issues - None part II - Items of Interest Reviewed Durina the Month DRESDEN Following shutdown on May 15, 1988, the MSIV's were tested and found not to close by spring force alone (total loss of air, including the accumulator). The NRC convened on AIT to investigate the problem. The station did an excellent job in testing and identifying the cause which appears to be excessive packing friction. The problem only manifests itself during gradual loss of air, not supply line rupture or rapid depressurization. The necessity (design basis) for the valves to close under those conditions is being evaluated by BWRED and G.E. BRAIDWOOD A event reported and subsequently cancelled by the station was judged significant by Offsite Re"lew. The event involves an unmonitored release path created from containment atmosphere to the environment. This path was created by the simultaneous conduct of two separate maintenance activities on the IB steam generator. Specifically, while a steam generator manway was removed for gasket repairs, the Shift authorized a safety valve to be removed on the same generator. The grounds used by the Shift to justify the safety valve removal coincident with the gasket repair was the containment pressure was negative with respect to outside atmosphere. This type of reasoning is questionable because no assurance exists that the negative pressure can be maintained throughout the valve repair. (i.e., HVAC failures or loss-of-offsite power.) These two maintenance activities should have been performed sequentially rather than simultaneously. Offsite Review is continuing to investigate from the perspective of practices used to prevent simultaneous conduct of separate, unrelated maintenance activities that by virtue of being performed together interact to degrade plant safety or reduce regulatory compliance. This concern is of generic sense and is not specifically for unmonitored release paths. BYRON Byron Quarterly Trend Review Meeting - The Byron Quarterly Trend Review meeting was held May 26, 1988. Offsite Review expressed concern about occassions where the station found it necessary to isolate more than 1 Steam Generator PORY (power operator relief valve). These valves are important for recovery from a steam generator tube rupture accident. The station shares this concern and have informed plant operators to quickly unisolate these valves if plant conditions warrant this action.

pago 4 Department of Nuclear Safety Activities Report - May. 1988 part II - Items of Interest Reviewed Durinx the Month (Continued) BYRON Also, at the meeting, the performance of the process computer and sequence of event computer was discussed. After several reactor trips this equipment has failed to record valuable information. The station described several corrective actions which should help the situation. However, it is believed that the equipment needs to be replaced with up-to-date equipment. Offsite Review expressed a concern that computer performance af ter transients may not be well documented in management performance report. Offsite Review will look into this question. ZION The Station agreed to have Nuclear Safety perform an independent review of their surveillance program this summer. A charter is beigg prepared to define the scope of the review. The scope will include the lessons learned from the Dresden Station surveillance program review and the Company's commitments to the NRC. The review is scheduled to be completed by August 31, 1988 and a report will be prepared by September 30, 1988. This schedule will provide the Zion Station with the result of the review prior to this fall refueling outage for Unit 2. part III - 1988 Sinnificant Events The following is a listing of 1988 Deviation Reports (DVR) which Offsite Review has classified as being significant using the IKp0 screening guide. No events have been classified by Offsite Review as significant in May, 1988. Significant events per unit per year (3 year average *) Byron Unit 1 0.66 Dresden Unit 2 2.00

• Byron Unit 2 1.00 (10/30/86)

Dresden Unit 3 1.33

• Braidwood Unit 1 1.00 (10/18/86)

LaSalle Unit 1 0.00

• Braidwor* Unit 2 0.00 (12/28/87)

LaSalle Unit 2 0.33 Zion Unit 1 2.00 Quad Cities Unit 1 0.33 Zion Unit 2 0.66 Quad Cities Unit 2 0.00

• Significent events rate at Byron Unit 2 & Braidwood do not represent 3-year average.

page 5 Department _pf Nuclear Safety Activities Report - May. 1988 port III - TREND ANALYSIS AND GRAPHS All graphs are based on event dates. 1. Graph III A illustrates the cumulative number of personnel related srrors for the period of March,1987 through February,1988. 2. Graph III B illustrates the cumulative percentage of personnel related errors of all deviations for the period of March,1987 through February, 1988. 3. Graph III C illustrates the cumulative number of electrical failures for the period of March,1987 through February,1988. 4. Graph III D illustrates the cumulative number of instrument failures for the period March,1987 through February,1988. 5. Graph III E illustrates the cumulative number of mechanical failures for the period of March,1987 through February,1988. 6. Graph III F illustrates the cumulative number of personnel errors for Operations Department for the period of March,1987 through February,1988. 7. Graph III G illustrates the cumulative number of personnel errors for Technical Services for the period of March,1987 through February,1988. 8. Graph III H illustrates the cumulative number of personnel errors for maintenance for the period March,1987 through February,1988. 9. Graph III I illustrates the cumulative number of personnel errors for non-station personnel for the period of March,1987 through February,1988 cumulative. TREND ANALYSIS DRESDEN No change in trends. QUAD CITIES No change in trends LASALLE COUNTY Although there has been a small rash of personnel related error DVRs since 2/29/88, early data for such errors in 1988 has been encouraging. The number of total DVRs continues to be low.

Page 6 Department of Nuclear Safety Activities Report - May. 1988 Part III - TREND ANALYSIS AND GRAPHS (Continued) ZION No change in trends. BYRON No change in trends BRAIDWOOD The improvement in personnel error trends exhibited in late 1987 has not continued into 1988. Particularly noteworthy are personnei errors by non-station personnel (Curve III.I). This trend is interestin5 from the standpoint of the Breatly reduced number of non-station personnel currently involved at Braidwood. I

page 7 Department of Nuclear Safety Activities Report - May. 1988 part IV - Station Related Activities Quad LaSalle Item Reviewed Dresden Cities Zion County Byron Braidwood Tech Spec Chanaes ? 2 2 procedures and 50.59 Evaluations 6 46 121 40 Onsite Reviews 1 1 1 Safety-Related Modifications 2 2 2 4 DVR Reports 25 11 15 12 25 99 Onsite Nuclear Safety Activities The LaSalle ONSG observed OAD verifying the calibration of a transducer on a diesel generator wattmeter. ONSG observed that the station's lifted lead procedure, lap-240-6, was not being followed during the process of isolating the instrument. Following a discussion with OAD, retraining was provided to all onsite OAD personnel. Also, procedure LES-GM-123 (used for the calibration of wattmeters) will be revised to include the transducer involved. The LaSalle ONSG participated with Quality Control and Tech Staff in a partial walkdown of a Drywell Cooling modification. Cooler E and its associated supports were inspected. ONSG was concerned that the cooler was making contact with the drywell liner. Tech Staff agreed to resolve the issue. The LaSalle ONSG questioned the drywell equipment qualification temperature setpoint for TE-Vp215. Tech Staff had raised the setpoint valve without replacing the necessary solenoid valves per the S&L letter. The oversight was quickly corrected. Tech Staff appreciated the concern and started tracking the temperature appropriately until the replacement of the SRV solenoids. The Byron ONSG reviewed NETWORK OE 2477, "Hydrogen Gas Accumulation in Charging Pump Suction piping at Farley", and issued letter BNS 88-12 to the station with their conclusions. The specific piping arrangement and enndition which caused the event at Farley do not exist at Byron. The Byron ONSG reviewed NETWORK OE's 2458 and 2483, which discuss a Catawba event when Auxiliary Feedwater flow was severely degraded due to the presence of asiatic clams. ONSG concluded that the specific event is unlikely at Byron because of the absence of clams, but it should be evaluated at braidwood, where clams have been found. ONSG is concerned that the nearby stagnant Essential Service Water supply piping may contain sediment from the Rock R3ver. ONSG recommended that when the planned modification to this piping is installed, the piping should be flushed and the effluent filtered to detect accumulation of river sediment.

Pago 8 popartment of Nuclear Safety Activities Report - May. 1988 Part IV - Station Related Activities (Continued) The Byron ONSG completed a study of Byron Station deviations that occurred during the first quarter of 1988. The rate of deviation occurrences is the same as in 1987; however, the rate of reportable events (LER's) and personnel error have decreased from 1987. On May 16, the Byron ONSC turned over to the Byron Station Regulatory Assurnace Department the responsibility to distribute and track INPO NETWORK Operating Experience (CE), Significant Event (IS) Operation and Haintenance Reminder (OR) and Human Performance Reports.

O Page 9 Department of Wuclear Safety Activities Report - May. 1988 Part VI - Special Activities Quarterly Meetings - Zion Station May 23, 1988 10:00 A.M. LaSalle Station May 24, 1988 9:00 A.M. Braidwood Station May 24, 1988 2:30 P.M. Quad Cities Station May 25, 1988 10:00 A.M. Byron Station May 26, 1988 10:00 A.M. Dresden Station May 27, 1988 10:00 A.M.

III.A PERSONNEL RELATED ERROR DVRS 03/01/87 To 02/29/88 CUMULATIVE 150.00 pAi mo00 140.00 / Th ese are devic tions ipdependently classi ' led by Offsity Review. These are the summat on of 120.00 pe[d management deficiencit

s. sonnei errorp, procedurci deficiencies, an f-100.00

/ / 80.00 f / / ,. BYRON [ - #.ORESDEN / / 40.00 M I ON [ UAD cITi A ASALLE / -~' --+--_, 20.00 c. . --.r.'<" ",,,,, Es 7.. m.e. ,..: g,........... 0.00 MAR APR MAY JUN JUL AUG SEP OCT NOV OEC JAN FEB MONTH OF OCCURRENCE III.B PERCENT OF PERSONNEL ERROR DVRS 03/01/87 To 02/29/88 CUMULATIVE 100.00 These are deviation reports (DVR's) Independently classified 90 00 by Off site Review as personnel errors, procedural deficiencies, _or management errors. The graph is computed by the following: Cumulative oersonnel related error DVR's of the end of the month'

• 100 70'00 Cumulative total DVR's at the end of the month 60.00 50.00 40 00 BRAIDA00D 30.00

's 20.00 YRON / ION ASALLE ,---:,,*~..-----...L:-:----,,,,,.,,,,,, j0RESDEN 121 t 10.00 - l.... T '-..l......,, j... *........,.,.

. a. i. - ~ -. v. r... - -

l l l l l l UAy ClTIES 0.00 MAR APR VAY JUN JUL AUG SEP OCT NOV OEC JAN FEB MONTH OF CCCURRENCE

III.C ELECTRICAL FAILURES 03/01 87 TO 02/29 88 CUMULATIVE 50.C; / l UUU LL I'U UURVL U IL I U)) U T~ l 3 LIL U]) U~Il V IL b y r classific 1 by Cffsite Review as elec tt~it at failures. .~ 40.N OFlESDEN @Y RON \\.....$ 35.00 ORWD 30.00 [ QUA ) CTY 25.00 l-20.00 SiON ,~ 'b*00 ~,. - - -+ _- ----m / _ASALLE 10.O,u 7?,,,e ~ W 5.00 ..a-0.00 MAR APR MAY JUN JUL AUG SEP OCT NOV OEC JAN FEB MONTH OF OCCURRENCE

E:I.D ::NSBUk3NT :?AILURES 03/01/87 TO 02/29/88 CUMULATIVE 80.00 b

l Tibesc at e devkation repor\\ s independt ntly 70.C0 ~~ ,,,6YRCN . ' N RE5 DEN 60.00, 50.00 mpAigwagg f a l< p 40.0: e' qiAD Cl" lES / N -,0N 30'c3 , '/ .-......>f'_ ,' / w AS AuE 20.00 y,, 10.CC ',.' f.,,.s ^*,,. T.-- ,-.0... ,e g'..- 0.03 MAR APR VAY JUN JUL AUG SEP OCT NOV DEC JAN FEB MONTH OF OCCURRENCE

r III.E MECHANICAL FAI~ LURES 03 01 87 TO 02 29 88 CUMULATIVE 60.00 Th ese ar$e detkation s incil 'ndently report clynific :i by Cffsite Review as vrechanical failures. 50.00 , N YRON i / s 40.00

• N RESDEN s

/ 'Z10t p y 30.00 / ,- '[# QRAID.0000 ... ~; 20.00 f-

• ~ ~ ~ '

,c LASALLE-- ,...,./ -,um r r,es 10.00

• ~~~---~.

/ c:... 0'00 MAR APR WAY JUN JUL AUG SEP OCT Nov DEC JAN FEB MONTH OF OCCURRENCE III.F PERSONNEL ERRORS OPERATIONS 03/01/87 To 02/29/88 CUMULATIVE 50.00 feee an Set h tion qepovfe-t'miepemiephy-ehfidj offuite fut'iew as personnef error by Ogeratic ns. 0heratic ns i nci udes fuel Handling, Stationrr en,ahd / ubricaiibn Pro' pram. I L ens sow 00c 40.00 / 30.00 / 20.00 =- BYRO* Z*Ct ...,..p....-- -'ESDEN / 10.00 ,,g. g

• • DUA0 CITIF.S

~~~_ ASALLE S...- ...A 4 ~ l 0.00-- MAR APR MAY JUN JUL AUG SEP OCT NOV DEC JAN FEB MONTH OF OCCURRENCE

III.G PERSONNEL ERRORS - TECH SERVICES 6 03/01/87 TO 02/29/88 CUMULATIVE '" 00 'These are Deviation Reports independentiy classified by enAsow000 Offsite Review as Personnel Error by Tech Services. Tech Services includes Rad / Chem, Central 12.00 Files and Tech Staff. 10.00 BYRON 8.00 s' ,a' 6.00 e' ASALLE / 4.00 ,-----l--------------------------r---------------'---------- ORESDEt / / 't0N i / 2.00

---

r-------

,......,................................................. 7. 9 U A D C I -- --- -{-- -- -[**- - -l l l l l l l l l 0.00 MAR APR MAY JUN JUL AUG SEP OCT NOV DEC JAN FEB MONTH OF OCCURRENCE III.H PERSONNEL ERRORS - MAINTENANCE 03/01/87 TO 02/29/88 CUMULATIVE 00 ~These are Deviation Reports independently classified by Offsite Review as Personnel Error by Maintenance. 10.00 _ Maintenance incl.udes Electrical, Instrument, and Mechanical Maintenance 9.00 8 00 BRAIDwCO 7.00 6.00 / \\ ZION ~ ,/ 4.00 e-CRESDEN i 3.00 gg j g, -- - - - -- - - - - - - -- --- ' - - - - - -- - - - - - - - - -[ r 2.00 s------------r-------- OUAD CtTIES 1.00 - --- ;'---------- l<..<'.....l.........{........[........l.........[,, l l l l l l o.00 MAR APR MAY JUN JUL AUG SEP OCT NOV DEC JAN FEB MONTH OF OCCURRENCE

III.I PERSONNEL ERRORS - NON STATION 03/01/87 TO 02/29/88 CUMULATIVE 22.00 -These are Deviation Reports htdependently classified by Offsite Review as Personnel Error by Non Station Personnel. O'00 3'on station incitides OAD, Contractors,. and Sttbstation Construction. 18.00 BRA 10 0 16.00 14.00 12.00 10.00 8.00 / 6.00 BYROM ' l--

---

k

^ 4.00 ...-......-............... < 1 '.....s c:'.- d ......... DRE S D C tl....... 2.00 - - ' - - - - - - - - - - - * ' ' * - - - - - - - - ' - - - - ' .....OUA G. C 4.T A E.S .g....... 7.... ..(........ T ~z'P" 'l 0.00 MAR APR MAY JUN JUL AUG SEP OCT NOV DEC JAN FEB MONTH OF OCCURRENCE t l l e,- -,.,y,

k dH2i%= Wf \\ WWW h H57 June 8, 1988 O W sidlo c4 o+rk pe e,<de9c4 3) To rego(ve -l careent9ex Add: TO: D. P. Galle

a. D C M

I'N M b8)db[W/ 4 WfJ H. L. Massin C/2 R. Spear

b. f7kl inhrp,%bf;pn y & p ff y n,j.9 gg,e n fj n y t

(Pack b b ack csce9Gf(,, Q4,,) p(* gef[f (g

• =dMB */f. rc h o f A f 'f b g,A g

SUBJECT:

Scope of L2RO2 b /ho() a/fers4en 6eivrg wen fo a,17e h

REFERENCE:

NOD-MA.13, Conduct of Work Planning, Paragraph .l.1.6 w il be a G M tuit(. NLA <% a face The station has completed a preliminary review of the approved surveillance b b h e4M deferral for L2RO2. This deferral will result in a significant increase in 0" Sub M the scope of this outage. 5/ya %. 1. The deferral waives the requirements of Technical Specification 4.0.2.b (3.25 surveillance interval) to permit the unit to operate until the scheduled start of the outage in October, 1988. However, by stretching this surveillance interval, we are shortening the subsequent interval such that many surveillances (requiring, unit shutdown) will require performance prior to L2RO3. 2. Approximately 80 tests requiring unit shutdown will have critical dates prior to L2RO3 at the end of L2RO2 if no action is taken. In the past our approach to this difficulty has been to repeat the required tests to .. crease the interval before the next test. The re-performance cf this many tests constitutes a significant increase in the scope of the outage. 3. The aforementioned surveillances can be broken down as follows: d. Valve Cycling / Timing / Indication Check 8 l b. Preventive Maintenance / Rebuilds 11 l c. Logic Tests 24 i d. Integrated System Tests 10 l e. Local Leak Rote Tests l'1 4. Several surveillances will be individually difficult to perform twice. a. Annual rebuild / inspection diesel generators 2A and 2B b. Local Leak Rate Testing of all 8 MSIV's c. Division III battery discharge test d. SBLC injection test 5. The schedule impact of this scope addition can not be properly evaluated until a more detailed schedule is available. These tests will fall on the critical path for return to service of the affected system since the test pre-requisites preclude concurrent maintenance activity. A one week outage duration increase is probably a very conservative estimate of the total impact. DOCUMENT ID 2438k

I s 6. One related issue could cause schedule impact. No guidance currently exists conce,r.ning the validity of performing an 18-month surveillance twice in a short period of time. The outage schedule will contain no minimum time between tests. Any change in this requirement in mid-outage will raise havoc with the schedule. 7. Since further licensing activity does not appear fruitful for L2RO2. Three alternatives seem to be available: a. Perform the surveillance twice in L2RO2. b. Perform the surveillance once in L2RO2 and accept a cycle 3 surveillance outage, c. Perform the surveillances once in L2RO2 and pursue a deferral for 1.2RO3. The station is presently pursuing option "a". 8. A detailed listing of the surveillances is attached. This letter meets the requirements for the referenced notification and also is a request for action on the standardized Technical Specification refuel surveillance frequency definition. It should be noted that Commonwealth Edison elected not to request in the subject deferral a resetting of the surveillance clocks based on someone in NRR informing Licensing that it would not be approved in time, h ,f&db v4 b G J. Diederich (/ h ation Manager GJD/DSB/kg xc: C. Allen l D. Berkman l R. Allen P. Thouvenin i T. Hammerich l J. Gieseker i l 1 I I DOCUMENT ID 2438k I

s ATTACHMENT A LISTING OF SURVEILLANCES WHICH MUST BE DONE TWICE IN L2RO2 Procedure Description LTS-100-3 MSIV LLRT (4) LES-GM-223 OAD Meter Calibrations LES-HP-202 HPCS System Logic Test LES-HP-203 HPCS System Thermal O/L Bypass LOS-MS-R4 SRV Downcommer Check Valve Cycling LES-DG-201 DG Annual Inspection (2) LES-DG-202 DG Annual Inspection LTS-600-6 ADS Accumulator Check Valve Test LTS-600-7 MSIV Accumulator Check Valve Test LMS-DG-01 DG Annual Inspection (2) LTS-900-10 HPCS System LLRT LTG-500-2 PC Vacuum Breaker Force Check (3) LES-MS-202 MSIV-LCS Thermal O/L Bypass LES-RI-202 RCIC Thermal O/L Bypass LES-RI-201 RCIC Logic Test LES-RI-203 RCIC Logic Test LES-PC-209 PCIS Group 2 Logic Test LES-PC-210 PCIS Manual Logic Test LTS-900-12 C RHR LLRT LTS-900-ll RCIC LLRT LTS-900-13 RR LLRT LES-RP-204 Neutron Monitoring Scram Logic LTS-600-15 Drywell Nitrogen Check Valve Exercise LES-PC-207 PCIS Group 7 Logic Test LES-RH-203 RHR Thermal O/L Bypass LTS-900-12 C RHR Suction LLRT LTS-500-13 MSIV LCS Functional Test LES-RR-201 ATVS System Logic Test LES-RR-202 ATVS System Functional Test LES-RP-206 RR Pump Trip System Functional Test LES-PC-203 PCIS Group 3 Logic Test LES-VQ-201 Vacuum Breaker Position Calibration LES-NB-201 ADS System Logic Test N/A SBGT System EQ Inspection LOS-SC-R1 SBLC Vessel Injection Test LES-PC-205 PCIS Group 5 Logic Test LTS-900-2 LPCS Injection Valve LLRT DOCUMENT ID 2438k

4 Procedure Description LOS-PC-R1 ~ PCIS Valve Cycle / PIT Test LES-AP-201 Primary Containment Over Current Breaker Calibration LTS-900-2 LPCS Injection Check LLRT LOS-SC-R2 SBLC Suction Valve Cycling LOS-RP-R1 Mode Switch Scram Functional Test LES-LP-201 LPCS System Logic Test LOS-SC-R3 SBLC Piping Flow Test LTS-700-8 Division III Battery Discharge Test LES-LP-202 LPCS System Thermal O/L Bypass LTS-900-4 B RHR Injection Check LLRT LTS-900-4 B RHR Injection Valve LLRT LTS-900-7 RHR Shutdown Cooling outboard Valve LLRT LTS-900-7 RHR Shutdown Cooling Inboard Valve LLRT LTS-900-12 B RHR System LLRT LTS-1000-31 Bus Duct Fire Seal Inspection LES-RH-205 RHR Division 1 O/L Bypass LES-RH-204 RHR Division 2 O/L Bypass LES-RH-201 Division II ECCS Logic Test LES-RH-200 Division I ECCS Logic Test LTS-800-1 Div. I DG Start and Load Acceptance LTS-800-4 Div. I DG 24 Hour Run LT'd-500-9 Div. I DG ECCS Response Time LTS-800-5 Div. II DG 24 Hour Run LTS-800-8 Div. II DG Trip Logic Test LTS-800-2 Div. II DG Start and Load Acceptance LTS-500-10 Div. II DG ECCS Response Time DOCUMENT ID 2438k

June 9, 1988 TO: General Office Department Heads RE: Safety-EAP Meetings i Since December of 1987, we have been having a series of meetings for General Office employes concerning safety and employe assistance related topics. When we began this program, it was to run approximately six months. This note is to announce the last two meet-ings scheduled as a part of this program. Please review the names of the employes from your department who have already attended these meetings so that those who have not attended may come to one of the last two meetings. The schedule is as follows: June 16, 1988, 10:00 a.m., Room 650-L June 16, 1988, 1:00 p.m., Room 650-L As with the other meetings, we exp,:t them to last approxi-mately one hour and fifteen minutes. If you have questions concerning the last meetings, please contact Dick Montgomery, Ext. 4063, Earl Hollins, Ext. 3999 or. Maryanne Kawczynski, Ext. 4010. I / / ,/' sp George P. Mitchell l i 1 l}}