ML20081L168

Proposed Tech Specs 1.1/2.1 Re Fuel Cladding Integrity & 1.2/2.2 Re RCS

ML20081L168
Person / Time
Site:	Quad Cities
Issue date:	06/27/1991
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20081L165	List: ML20081L163 ML20081L168
References
NUDOCS 9107020422
Download: ML20081L168 (70)

Text

--_ _ _ _ _ - _ _ _ _ _ . _ _ _ .

o 'N PROPOSED TECH SPEC TS 1.1/ 2.1 and 1.2 / 2.2

' FUEL CLADDING INTEGRITY" and

' REACTOR COOL ANT SYSTEM' 9107020422 910627 PDR ADOCK 05iOOO2 *A F PDR

o '%

QUAD CITIES UNITS 1&2 DPR-29 & DPR-30 1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMITS LIMITING SAFETY SYSTEM SETTING A. Thermal Power, High Pressure and A. Neutron Flux Trip Settings High Flow The LIMITING SAFETY SYSTEM The MINIMUM CRITICAL POWER RATIO SETTINGS for neutron flux shall (MCPR) shall not be less than 1.06 be as specified below:

with the P.EACTOR VESSEL PRESSURE greater than 800 psig and core flow 1. The APRM High Flux (flow greater than 10% of rated flow. biased) trip setting shall be as shown in Figure 2.1-1 and shall be :

APPLICABILIIX:

S s (.58Wo + 62)

OPERATIONAL MODES 1 and 2.

with a maximum setpcint of 120%

ACTION: for core flow equal to 98 x 106 lb/hr and greater.

With MCPR less than 1.06 and the REACTOR VESSEL PRESSURE greater Where:

than 800 psig and core flow greater than 10% of rated flow, be S = setting in percent of rated in at least HOT SHUTDOWN within 2 power hours and comply with the require-ments of Specification 6.4. Wo = percent of drive flow required to produce a rated core flow of 98 million lb/hr.

In the event of operation with a MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) greater than the FRACTION OF RATED POWER (FRP) , the setting shall be modified as follows:

EEP S s (.58Wo + 62) [ MFLPD )

Where:

l FRP = FRACTION OF RATED POWER

( 2 511 MW,)

MFLPD = MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.1/2.1-1 l

l l

t h QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 where the limiting power density for each bundle is the design LINEAR HEAT GENERAT10h RATE for that bundle.

The ratio of FRP/MPLPD shall be set equal to 1.0 unless the actual operating value is less than 1.0 in which case the actual operating value will be used. This adjustment may also be performed by increasing the APRM gain by the inverse ratio, MFLPD/FRP, which accomplishes the same degree of protection as reducing the trip setting by FRP/MFLPD.

2. The APRM High Flux (15% scram) shall be set at less than or equal to 15% of RATED NEUTRON FLUX.

3. The IRM High Flux scram setting shall be set at less than or equal to 120/125 of full scale.

APPLICABILITY:

As shown in Table 3.1-1.

ACTION:

With a neutron flux trip setcing less conservative than the value above, declare the CHANNEL inoperable and apply the applicable ACTION statement requirement of Specification 3.1. A until the CHANNEL is restored to OPERABLE status with its setpoint adjusted consistent with the trip setting value above.

1.1/2.1-2

QUAD CITIES UNITS 1& 2 DPR-29 & DPR-30 B. Thermal Power, Low Pressttre or Low Flow B. APRM Rod Block Setting Thermal power shall not exceed 25% The LIMITING SAFETY SYSTEM SETTING of RATED THERMAL POWER with the for the APRM Rod Block shall be REACTOR VESSEL PRESSURE less than as shown in Figure 2.1-1 and shall be:

800 psig or core flow less than 10% of rated flow. S S (.58Wo + 50)

APPLICABILITY: The definitions used above for the APRM scram trip apply. In the event of operation with a MAXIMUM OPERATIONAL MODES 1 and 2. FRACTION OF LIMITING POWER DENSITY (MFLPD) greater than the FRACTION ACTION: OF RATED POWER (FRP) , the setting shall be modified as follows:

With thermal power exceeding 25% of RATED THERMAL POWER and the REACTOR VESSEL PRESSURE less than 800 psig ERE or core flow less than 10% of rated S S ( . 5 8 W, + 50) ( MFLPD ]

flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.4. The definitions used above for the APRM scram trip apply.

The ratio of FRP tc MFLPD shall be set equal to 1.0 unless the actual operating value is less than 1.0, in which case the actual operating value will be used.

This may also be performed by increasing the APRM gain by the inverse ratio, MFLPD/FRP, which accomplishes the same degree of protection as reducing the trip setting by FRP/MFLPD.

APPLICABILITX:

As shown in Table 3.2-3.

i ACTION:

4 With the APRM Rod Block setting less conservative than the valur shown above, declare the CHANNt.u inoperable and apply the applicable ACTION statement requirement of Specification 1.1/2.1-3

s is QUAD CITIES UNITS 1& 2 DPR-29 & DPR-30 3.2.C until the CHANNEL is restored to OPERABLE status with its setting adjusted consistent with the trip setting value above.

C. Power Transient C. Other Reactor Protection Systen Instrumentation Setpoints

1. Neutron flux sh.tl1 not exceed the scram setting established in The following reactor protection Spccification 2.1.A for longer system instrumentation-setpoints than 1.5 seconds as indisated by shall be maintained:

the process couputer.

1. The reactor low water level

2. When the process computer is out scram setting shall be 2 144 of service, this SAFETY LIMIT inches above the top of the shall be assumed to be exceeded active fuel at normal if the neutron flux exceeds the operating conditions. Top scram setting established by of active fuel is defined Specification 2.1.A and a control to be 360 inches above rod scram does not occur. vessel zero (Reference Bases 3.2).

2. Turbine stop valve closure scram shall be s 10% valve closure from full open.

APPLICABILITY: 3. The scram for turbine control valve fast closure due to OPERATIONAL MODES 1 and 2. actuation of the fast acting solenoid valve shall be 1460 psig EHC fluid pressure.

ACTION:

4. Main stramline isolation With the neu:ron flux in excess of valve closure scran shall the scram setting established in be $ 10% valve closure from Specificati(n 2.1.A and a scram does full open.

not occur within 1.5 seconds as indicated by the process computer or 5. Turbine EHC control fluid low immediately if the process computer pressure scram on loss of is not in service, manually scram control oil pressure shall be the reactor and be in at least HOT. . set at 2 900 psig.

SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specifi- 6. Condenser low vacuum scram cation 6.4. shall be set at 1 21 inches Hg vacuum.

1.1/2.1-4 ,

e ),

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 APP _LICABILITY:

As shown in Table 3.1-1.

ACTION:

With a reactor protection system instrumentation setpoint less conservative than the value shown above, declare the CHANNEL inoperable and apply the applicable ACTION s iment requirement of Specification 3.1. A until the CHANNEL is restored to OPERABLE status with its setpoint adjusted consistent with the trip sotpoint value above.

D. Reactor Vessel Water Level D. Reactor Low Water Level ECCS Initiation The reactor vessel water level shall be greater than a level correspond- The reactor low water level ECCS ing to 12 inches above the top of setpoint shall be 2 84 inches the active fuel when it is seated above the top of the active fuel in the core. Top cf active fuel is at normal operating conditions.

3 defined to be 360 inches above Top of active fuel is defined to j' vessel aero (Ruforence Bases 3.2), be 360 inches above vessel zero (Reference Bases 3.2).

APPLICA3ILITY: APPLICABILITY:

OPERATIONAL MODES 3, 4 and 5. As shown in Tarle 3.2-2.

ACTIQH: ACTION:

4 With the reactor vessel water level 17tth the reactor low water level less than that corresponding to 12 ECCS setpoint less conservative inches above the top of the active than the value shown above, fuel, ranually initiate the ECCS to declare the CHANNEL inoperable restore the water level. Comply and apply the applicable ACTION with the requirements of Specifica- statement requirement of tion 6.4. Specification 3.2.'B until the CHANNEL is restored to OPERABLE utatus with its setpoint adjusted consistent with the trip setting value above.

1.1/2.1-5

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QUAD CITIES UNITS 1 &2 LPR-29 & DPR-30 E. Main Steamline Low Pressure Isolation The main steamline low pressure initiation of main steamline isolation valve closure shall be 1 825 psig.

AEEhlCABILITY :

As shown in Table 3.2-1.

Agngl{:

With the main steamline low pressure setpoint less conservative than the value shown above, declare the CHANNEL inoperable and apply the applicable ACTION statement requirement of Specification 3.2. A until the CHANNEL is restored to OPERABLE status with its setpoint adjusted consistent with the trip setpoint value above.

l t

1.1/2.1-6

• b QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 APRM Flow Reference Scram and APRM Rod Block Settings

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FIGURE 2.1-1 1.1/2.1-7

a- k QUAD. CITIES UNITS 1 & 2 DPR-29 & DPR-30 14C _

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FIGURE 2.1-2 (SCHEMATIC)

APRM FLOW BIAS SCRAM RELATIONSHIP TO NORMAL OPERATING CONDITIONS 1.1/2.1-8

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QUAD CITIES UNITS 1&2 .

DPR-39 & DPR-30 1 1.1 SAFETY LIMIT BASES The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient. Because fuel damago is  ;

not directly observable, a etop-back approach is used to catablish a SAFETY LIMIT such that the MINIMUM CRITICAL -

POWER RATIO (MCPR) is not loss than the fuel cladding integrity SAFETY LIMIT. MCPR greater than the fuel

- cladding integrity SAFETY LIMIT represents a conservativo 3

margin - relativo to the conditions required to maintain '

fuel cladding integrity.

The fuel cladding is ona of the physical boundarios which separate radioactivo materials from the environs. The  !

intoority of the fuel cladding in related to its relativo froodom from perforations or cracking Although some corrosion or uso-relatcd cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulativo and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly abovo design conditions and the protection system safety settings. While fission product migration from cladding perforations is just as n aasurt ble as thet from use related crackino .'..

thermally caused cladding perforations signal a threshold  ?

beyond which atill greater thermal stresses - may cause gross rather than incremontal cladding datorioration.

Thorsfore, the fuel cladding integrity SAFETY LIMIT to defined with margin to the conditions which would product enset of TRANSITION BOILING (MCPR of 1.0). Thoso conditic,ns represent a significant departure from the

~ condition intended by design for planned operation.

Thorafore,- the fuel cladding integrity SAFETY LIMIT is established such that no calculated fuel damage shall result from an abnormal operational. transient. Basis of l

the values damage shall result from an abnormal operational transient. Basis of the. values derived for

this-SAFETY LIMIT for nach fuel type is documented in Reference 1.

I i A. Thermal Power, liigh Pressure and liigh Flow Onset of TRANSITION BOILING results in a decrease in heat transfer from the cladding and therefore l

clovated cladding temperature and the possibility of cladding failure. However, the existence of B 1.1/2.1-1 l

• D QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 critical power, or TRANSITION BOILING is not a directly observable parameter in an operating reactor.

Therefore, the margin to TRANSITION BOILING is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is enaracterized by the CRITICAL POWER RATIO (CPR), which is the ratio of the bu' idle power which would produce onset of TRANSITION BOILING divided by the actual bundle power. The minimum value of this ratio for any bundic in the core is the MINIMUM CRITICAL POWER RATIO (MCPR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables (Figure 2.1-2).

The M6PR fuel cladding integrity SAFETY LIMIT has suf ficient conservatism to assure that, in the event of an abnormal oporational transient initiated from the normal operation condition, more than 99.9% of the fuel rods in the core are expected to avoid TRANSITION BOILING. The margin between MCPR of 1.0 (onset of TRANSITION BOILING) and the SAFETY LIMIT, is derived from a detailed statistical analysis i considering all of the uncertainties in monitoring the-core operating state, including uncertainty in the TRANSITION BOILING correlation (see e.g.,

Reference 1). Because the TRANSITION BOILING ,

correlation is based on a large quantity of '

full-scale data, there is a very high confidence that operation of a fuel assembly at the condition of MCPR equals the fuel cladding integrity SAFETY LIMIT would not produce TRANSITION BOILING.

However, if TRANSITION BOILING were to occur, cladding perforation would not be expected. ,

cladding temperature would increase to approximately 1100'F, which is below the perforation temperature of the cladding material. This had been verified by tests in the General Electric Test Reactor (GFTR),

where similar fuel operated above the critical heat flux for a significant period of- time (30 minutes) without cladding perforation.

If reactor pressure should ever exceed 1400 psia during normal power operation (the limit of applicability of the TRANSITION BOILING B 1.1/2.1-2

_ _ . _ _ _ . _ _ _ _ _ _ _ . . . _ _ _ _ _ .. .~._ _ . _ _ . - _ _ _

e d QUAD CITIES UNITS 1&2 DPR-29 & DPR-30 correlation), it would be assumed that the fuel cladding integrity SAFETY LIMIT has been violated.

In addition to the TRA11SITIO!i BOILIllG limit (MCPR) operation is constrained to a maximum LilGR specified in the CORE OPERATING LIMITS REPORT for various fuel types. This constraint is established by Specification 3. 5. !! to provide adequate safety margin to 1% plastic strain for abnormal operating tranaients initiated from high pc,wer conditions.

Specification 2.1. A.1 provides for equivalent safety margin for transients initiated from lower power conditions by adjusting the APRM flow-biased scram setting by the ratio of FRP/MFLPD.

Specification 3 . 5 . 11 establishes the maximum LilGR which cannot be exceeded under steady state power operation.

B. Thermal Power, Low Pressure or Low Flow At pressures below 800 psia, the core elevation pressure drop (0 power, o flow) is greater than 4.56 psi. At low powers and flows, this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low powers and flows will always be greater then 4.56 psi. Analyses show that with a l flow of 28 x 103 lb/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi driving head will be greater than 28 x 103 lb/hr. Full scale ATLAS test data taken at pre s...u re s from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. At 25% of RATED TilERMAL POWER, the peak powered bundle woub have to be operating at 3.86 times the averago pewered bundle in order to achieve this bundle power. Thus, a core thermal power limit of 25% for reactor pressures below 800 osia is concorvative.

C. Power Transient During transient operation, the heat flux (thermal power- to water) would lag behind the neutron flux due to the inherent heat transfer time constant of the fuel, which is 8 to 9 seconds. Also, the B 1.1/2.1-3 1

l

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QUAD CITIES UllITS 1 & 2 DPR-29 & DPR-30 LIMITING SAFETY SYSTEM GETTI!1GS for reactor scram are at values which will not allow the reactor to be operated above the SAFETY LIMIT during normal operation or during other plant operating situations which have been analyzed in detail. In addition, control rod scrams are such that, for normal operating transients, the neutron flux transient is terminated before a significant increase in surface heat flux occurs. Control rod scram times are checked as required by Specification 4.3.D and the MCPR operating limit is modified as necessary por Specification 3.5.0.

Exceeding a neutron flux scram setting and a failure of the control rods to reduce flux to less than the scram setting within 1.5 seconds does not necessarily imply that fuel is damaged; however, for this specification, a SAFETY LIMIT violation will be assumed any time a neutron flux scram setting is exceeded for longer than 1.5 seconds.

If the scram occurs such that the neutron flux dwell time above the LIMITING SAFETY SYSTEM SETTING is less than 1.7 seconds, the SAFETY LIMIT will not be exceeded for normal turbine or generator trips, which are the most severe normal operating transients expected. These analyses show that, even if the bypass system fails to operate, the design limit of MCPR equals the fuel cladding integrity SAFETY LIMIT is not exceeded. Thus, use of a 1.5 second limit provides additional margin.

The computer provided has a sequence annunciation program which will indicate the sequence in which scrams occur, such as neutron flux, pressure, etc.

This program also indicates when the scram setpoint is cleared. Th L will provide information on how long a scram condition exists and thus, provide some measure of the energy added during a transient.

Thus, computer information normally will be available for analyzing scrams: however, if the computer information should not be available for any scram analysis, Specification 1.1.C.2 will be relied on to determine if a SAFETY LIMIT has been violated.

D. Reactor Vessel Water Level During periods when the reactor is shutdown, consideration must also be given to water level B 1.1/2.1-4

_ - - . . _ _ . ~ , -_.. _ __. - ._. _ .- - - -

. 4 QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 i

requirements due to the effect of decay heat. If ,

reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core-cooling capability could Icad to elevated cladding temperatures and cladding perforation. The core will be cooled sufficiently to prevent cladding nelting should the water -level be reduced to two-thirds core height. Establishment of the SAFETY LIMIT at 12 inches above the top of the fuel

• provides adequate margin. This icvel will be continuously monitored whenever the recirculation pumps are not operating.

• Top of the active fuel is defined to be 360 inches above vessel zero (Reference Bases 3.2).

References

1. " Generic Reload Fuel Applications", NEDE-24011-P-A*.

• Approved revision number at time reload fuel analyses are performed.

t 3

1 B 1.1/2.1-5 l-

+ 6 QUAD CITIES UNITS ; &2 DPR-29 & DPR-30 ,

2.3 LIMITING SAFETY SYSTEM SETTING BASES The abnormal operational transients applicable to operation of the units have been analyzed throughout the spectrum of planned operating conditions in accordance with Regulatory Guide 1.49. In addition, 2511 MWt is the licensed maximum steady-state power level of the units.

This maximum steady- state power level will never '

knowingly be exceeded.

conservatism incorporated into the transient analysis is documented in References 1 and 2. Transient analyses are initiated at the conditions given in these references.

The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptabic by technical '

specifications. The offects of scram worth, scram delay time, and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rapid insertion of negative reactivity is assured by the time requirements for 5% and 20% insertion. Dv the time the rods are 60%

inserted, approximately 4 dol'les of negative reactivity have been inserted, which strongly turns the transient and accomplishes the desired effect. The times for 50% and 90% insertion are given to-assure proper completion of the '_

expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.

The MCPR operating limit-is, however, adjusted to account for the statistical variation of measured scram times as discussed in Reference 2 and the bases of Specification 3.5.0.

Steady-state operation without forced recirculation will not be permitted except during startup testing. The analysis to support operation at various power and flow relationships has considered cperation with either one or two recirculation pumps.

The bases for individual trip settings are discussed in the following paragraphs.

For analyt,es of the thermal consequences of the transients, the MCPRs stated in the CORE OPERATING LIMITS REPORT as the LIMITING CONDITION FOR OPERATION bound those which are conservatively assumed to exist prior to B 1.1/2.1-6

e +

QUAD CITIES Ul11TS 1 & 2 DPR-29 & DPR-30 initiation of the transients.

A. Neutron Flux Trip Settings

1. APRM Flux Scram Trip Setting (RUN OPERATIONAL MODE)

The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of RATED THERMAL POWER.

Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel.

Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting. Analyses demonstrate that, with a 120% scram trip setting, none of the abnormal operational transients analyzed violates the fuel SAFETY LIMIT, and there is a substantial margin from fuel damage.

Therefore, the use of flow-referenced scram trip provides even additional margin.

An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity SAFETY LIMIT is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity SAFETY LIMIT, yet allows operating margin that reduces the possibility of unn( .' ssar/ sc"ams.

The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of MAXIMUM B 1.1/2.1-7

e w l

l QUAD CITIES Ul11TS 1&2 DPR-29 & DPR-30 FRACTION OF LIMITI!1G POWER DEllSITY (MFLPD) and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1.A.1 when the MFLPD is grator than the FRACTIOli OF RATED POWER (FRP). The adjustment may be accomplished by increasing the APRM gain by the reciprocal of FRP/MFLPD.

This provides the same degree of protection ao reducing the trip settings by FRP/MFLPD by raising the initial APRM readings closer to the trip settings such that a scram would be received at the same point in a transient as if the trip settings had been reduced by FRP/MPLPD.

2. APRM Flux Scram Trip Setting (REFUEL or STARTUP/IIOT STAllDBY OPERATIO!JAL MODE)

For operation in the STARTUP OPERATIOliAL MODE while the reactor is at low pressure, the APRM scram setting of 15% of rated power provides adequate thermal margin between the setpoint and the SAFETY LIMIT, 25% of rated. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup.

Effects of increasing pressure at zero or low void content are minor; cold water from sources available during startup are not much colder than that already in the system; temperature coefficients are small; and, control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer. Of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5% of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the SAFETY LIMIT. The 15% APRM scram remains active until B 1.1/2.1-8

e v QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 the modo switch is placed in the RUN position.

This switch occurs when reactor pressure is greater than 825 psig.

3. IRM Flux Scram Trip Setting The IRM system consists of eight chambers, four in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are broken down into 10 ranges, each being cne-half a decade in size.

The IRM scram trip setting of 120 divisions is l

active in each range of the IRM. For example, if the instrument were on Range 1, the scram setting would be 120 divisions for that range; likewise, if the instrument were on Range 5, the scram would be 120 divisions on that range.

l Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up.

l The most significant sources of reactivity I change during the power increase are due to control rod withdrawal. In order to ensure that the IRM provides adequate protection against the single rod withdrawal error, a

, range of rod withdrawal accidents was analyzed.

This analysis included starting the accident at l various power levels. The most severe case I involves an initial condition in which the l

reactor is just subcritical and the IRM system l

is not yet on scale.

Additional conservatism was taken in this f analysis by assuming that the IRM channel l

l closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power is limited to 1% of rated power, thus maintaining MCPR above the i fuel cladding integrity SAFETY LIMIT. Based on the above analysis, the IRM provides protection i against local contro; rod withdrawal errors and continuous withdrawal of control rods in the sequence and provides backup protection for the l

APRM.

I B 1.1/2.1-9

e *v QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 j D. APRM Rod Block Setting Reactor power level may be varied by moving control l rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent gross rod withdrawal at constant recirculation flow l rate to protect against grossly exceeding the MCPR l fuel cladding integrity SAFETY LIMIT. This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation ,

flow range. The margin to the SAFETY LIMIT l increases as the flow decreases for the specified  ;

trip setting versus flow relationship; therefore, the worst-case MCPR which could occur during steady-state operation is a 108% of rated thermal power because of the APRM rod block trip setting.

The actual power distribution in the core is established by specified control rod sequences, and is monitored continuously by the incore LPRM system.

As with APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the MAXIMUM FRACTION OF LIMITING POWER DENSITY exceeds the FRACTION OF RATED POWER, thus preserving the APRM rod block safety margin. As with the scram setting, this may be accomplished by adjusting the APRM gains.

C. Other Reactor Protection System Instrumentation Setpoints

1. Reactor Low Water Level Scram The reactor low water level scram is set at a point which will assure that the water level used in the bases for the SAFE 7Y LIMIT is maintained. The scram setpoint is based on normal operating temperature and pressure conditions because the level instrumentation is density compensated.

2. Turbine Stop Valve Scram The turbine stop valve closure scram trip anticipates the pressure, neutron flux, and B 1.1/2.1-10

. 4 QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of 10% of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the MCPR fuel cladding integrity SAFETY LIMIT, even during the worst-case transient that assumes the turbine bypass is closed.

3. Turbine Control Volve Fast Closure Scram The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass; i.e., it prevents MCPR from becoming less than the MCPR fuel cladding integrity SAFETY LIMIT for this transient. For the load rejection without bypass transient form 100%

power, the peak heat flux (and therefore LHGR) increases on the order of 15% which provides wide margin to the value corresponding to 1%

plastic strain of the cladding.

The trip setpoint of greater than or equal to 460 psig -EHC fluid pressure was developed to ensure that the pressure switch is actuated prior to the closure of the turbine control valves (at approximately 400 psig EHC fluid pressure), yet assure that the system is not actuated unnecessarily, due to EHC system pressure transients which may cause EHC system pressure to momentarily decrease.

4. Main Steamline Isolation Valve Closure Scram The low pressure isolation of the main steamlines at 825 psig was provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature in the RUN OPERATIONAL MODE, which occurs when the main steamline isolation

! valves are closed, to provide for reactor shutdown so that high power operation at low i

reactor pressures does not occur, thus providing protection for the fuel cladding

! integrity SAFETY LIMIT. Operation of the 1

l B 1.1/2.1-11

• 4 QUAD CITIES UNITS 1& 2 DPR-29 & DPR-30 reactor at pressures lower than 825 psig requires that the reactor modo switch be in the STARTUP/ HOT Standby position, where protection of the fuel cladding integrity SAFETY LIMIT is provided by the IRM And APRM high neutron flux scrams. Thus, the combination of main steamline low pressure isolation and isolation valve closure scram in the RUN OPERATIONAL MODE assures the availability of the neutron flux scram protection over the entire range of applicability of fuel cladding integrity SAFETY LIMIT. In addition, the isolation valve closure scram in the RUN OPERATIONAL MODE anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure. With the scrams set at 10% valve closure in the RUN OPERATIONAL MODE, there is no increase in neutron flux.

5. Turbine EHC Control Fluid Low Pressure Scram The turbine EHC control system operates using high pressure oil. There are several points in this oil system where a loss of oil pressure could result in a fast closure of the turbi.no control valves. This fast closure of the turbine control valves is not protected by the turbine control valve fast closure scram, since

, failure of the oil system would not result in I

the fast closure solenoid valves being actuated. For a turbine control valve fast j closure, the core would be protected by the

(

APRM and reactor high pressure scrams.

However, to provide the same margins as provided for the generator load rejection on fast closure of the turbine control valves, a scram has been added to the reactor protection system which senses failure of control oil pressure to the turbine control system. This is an anticipatory scram and results in reactor shutdown before any significant increase in neutron flux occurs. The transient response is very similar to that resulting from the turbine control valve fast closure scram. The scram setpoint of 900 psig is set high enough to l provide the necessary anticipatory function and l low enough to minimize the number of spurious scrams. Normal operating pressere for this system is 1250 psig. Finelly, the control B 1.1/2.1-12

. . ___ _ _. . .~- -_. - - - - -

i QUAD CITIES UNITS 1 &2 DPR-29 & DPR-30 valves will not start to close until the fluid pressure is 600 psig. Therefore, the scram occurs well before valve closure begins.

6. Condenser Low Vacuum Scram Ioss of condenser vacuum occurs when the condenser can no longer handle the heat input.

Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat inpttt to the condenser. Closure of the turbine stop and bypass valves causes a cressure transient, neutron flux rise and an increase in surface heat flux. To prevent the cladding SAFETY LIMIT from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure in the RUN OPERATIONAL MODE. The Turbine stop valve closure scram function alone is adequate to prevent the cladding SAFETY LIMIT from being exceeded, in the event of a turbine trip transient with bypass closure.

The condenser low vacuum scram is anticipatory to the stop valve closure scram and causes a scram before the stop valves are closed and thus, the resulting transient is less severe.

Scram occurs in the RUN OPERATIONAL MODE at 21-inches Hg vacuum, stop valve closure occurs at 20-inches Hg vacuum, and bypass closure at 7-inches Hg vacuum.

D. Reactor Low Water Level ECCS Initiation The emergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident, and to limit fuel cladding temperature to well below the cladding molting temperature to assure that core geometry remains intact, and to limit any cladding metal-water reactor to less than 1%. To accomplish their intended function, the capacity of each emergency core cooling system component was established based on the reactor low water level scram setpoint. To lower the retpoint of the low water level scram would increase the capacity requirement for each of the ECCS components. Thus, the reactor vessel, low water level, scram was set low enough to permit margin for operation, yet will not be set lower B 1.1/2.1-13

P 4 j l

l 1

QUAD CITIES UNITS 1& 2 DPR-29 & DPR-30 because of ECCS capacity requirements.

l The design of the ECCS components to meet the above I criteria was dependent on three previously set parameters: the maximum break size, the low water level scram setpoint, and the ECCS initiation  !

setpoint. To lower the setpoint for initiation of i the ECCS could lead to a loss of effective core cooling. To raise the ECCS initiation setpoint j would be in a safe direction, but it would reduce i the margin established to prevent actuation of the  !

ECCS during normal operation or during normally )

expected transients. l E. Main Steamline Low Pressure Isolation i

The low pressure isoJation at 825 psig was provided to give protection against fast reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature, which occurs in the RUN OPERATIONAL MODE, when the main steamline isolation valves are closed to provide for reactor shutdown so that operation at pressures lower than those specified in the thermal I hydraulic SAFETY LIMIT does not occur, although operation at a pressure lower than 825 psig would not necessarily constitute an unsafe condition.

i l

l l'

REFERENCES

1. " Generic Reload Fuel Application," NEDE-24011-P-A.*

2. " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors", General Electric Co.

Licensing Topical Report NEDO 24154 Vols. I and II and NEDE-24154 Volume III as supplemented by letter dated September 5, 1980 from R.H. Buchholz (GE) to P.S. Check ,

(NRC).

' Approved revision number at time reload analyses are

! performed.

B 1.1/2.1-14

e w QUAD CITIES UNITS 1&2 DPR-29 & DPR-30 1.2/2.2 REACTOR COOLANT SYSTEM SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING A. Reactor Coolant System Pressure A. Reactor Coolant System liigh-Setting Pressure Scram Setting The reactor coolant system Reactor coolant high-presourc pressure as measured by the scram shall be not at 1 10(0 vessel steam space pressure psig.

indicator shall not exceed 1345 psig. APPLICABILITY :

AEPLICABILITY: As shown in Table 3.1-1.

OPERATIONAL MODES 1, 2, 3 and 4. ACTION:

ACTION: With the reactor coolant high-precoure scram setting less With the reactor coolant system conservative than the valuo pressure, as measured in the shown above, declare the CilANNEL reactor vessel steam space, inoperable and apply the above 1345 psig, be in at least applicable ACTION statement ilOT SifUTDOWN with reactor requirement of Specification coolant system pressure less 3.1.A until the CllANNEL is than or equal to 1345 psig restored to OPERABLE status with within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with its setpoint adjusted consistent tha requiremento of with the trip setting value specification 6.4. above.

B. Primary System Safety valve Settings Primary system safety valve nominal settings shall be as follows:

1 valve at 1135 psig d) 2 valves at 1240 psig 2 valves at 1250 psig 4 valvos at 1260 psig U) Target Rock combination safety / relief valve.

1.2/2.2-1 l

e e 1

i QUAD CITIES UNITS 1 & 2 i DPR-29 & DPR-30 I The allowable setpoint error for each valve shall be i 1%.

APPLICABILITY:

As shown in Specification 3.6.G.

b.CT1911i With a primary system safety valvo setting loss conservativo ,

than the valuo shown above, '

apply the applicable ACTION-  :

statomont requiremont of t specification 3.6.G.

t 5

T Y

1.2/2.2-2 i

r, , ... ~ . . . .-.. .--.-,=,.,,,,._,,ms..m. . , , ,,,,ym_,.__v__., ,,.ww,. , r,- - . _ , , , - ,

_ _ m . ___ _ _ . . . _ _ _ . _ . _ . _ _ _ _ _ _ _ - __ _ _ _

e o QUAP CITIES UNITS 1 & 2 DPR-29 & DPR-30 1.2 SAFETY LIMIT BASES The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products. It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

The pressure SAFETY LIMIT of 1345 psig, as measured by the vessel steam space pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor vessel.

The 1375 psig value is derived from the design pressures of the reactor pressure vessel and coolant system pipinrj.

The respective design pressures are 1250 psig at 575'T t nd 1175 psig at 560*F. The pressure SAFETY LIMIT was chosen as the lower of the pressure transients permitted by the applicable design codes, '.SME Boller and Pressure Vessel Code Section III for the pressure vessel, and USASI B31.1 Code for the reactor coolant system piping. The ASME Boiler and Pressure vessel Code permits pressure transients up to 10% over design pressure (110% x 1250 =

1375 poig), and the USASI Code permits pressure transients up to 20% over design pressure (120% x 1175 = 1410 paig).

The SAFETY LIMIT pressure of 1375 psig is referenced to the lowest elevation of the reactor vessel. The design pressure for the recirculation suction line piping (1175 psig) was chosen relative to the reactor vessel design pressure. Demonstrating compliance of peak vessel pressure with the ASME overpressure protection limit (1375 psig) assures compliance of the suction piping with the USASI limit (1410 psig) . Evaluation methodology to assure that this SAFETY LIMIT pressure is not exceeded for any reload is documented in Reference 1. The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety l

pressure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig; this is a factor of 1.5 below the yield strength of 40,100 psi at 575'F. At o the pressure limit of 1375 psig, the general membrane stresa will only be 29,400 psi, still safely below the yield strength.

The relationships of stress levels to yield strength are comparable for the primary system piping and provides l similar margin of protection at the established pressure SAFETY LIMIT.

B 1.2/2.2-1 ,

. _ _ _ __. _ .______._....m._. _ ._.m . _ . _ _ _ . _ _ ._ _ _ _ . _ _ _ . . . . _ . __

e o QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 The normal operating pressure of the reactor coolant system is 1000 psig. For a turbine trip or loss of electrical load transient, the turbine trip scram or generator load reject scram, together with the turbine bypass system, limit pressure to approximately 1100 psig (References 2, 3, and 4). In addition, pressure relief valves have been provided to reduce the probability of the safety valves operating in the event that the turbine bypass should fail. Finally, the safety valves are sized to keep the reactor vessel peak pressure below 1375 psig with no credit taken for relief valves during the postulated full closure of all MSIVs without direct (valve position switch) scram. Credit, however, is taken for the neutron flux scram. The indirect flux scram and safety valve actuation provide a3 equate margin below the

' Allowable peak vessel pressure of 1375 psig.

Reactor pressure is continuously monitored in the control room, during operation, on a 1500 psi, full-scale, pressure recorder.

References

1. " Generic Reload Fuel Application", NEDE-24011-P-A*

2. SAR, Section 11.22 l 3. Quad Cities 1 Nuclear Power Station first reload license submittal, Section 6.2.4.2, February 1974.

4. GE Topical Report NEDO-20693, General Electric Bolling Water Reactor Reload No. 1 Licensing l submittal for Quad Cities Nuclear Power Station Unit 2, December 1974.

• Approved revision number at time reload analyses are performed.

B 1.2/2.2-2

, O I

I QUAD CITIES UNITS 1&2 DPR-29 & DPR-30 2.2 LIMITING SAFETY SYSTEM SETTING BASES In compliance with Section III of the ASME Code, the safety valves must be set to opers at no higher than 103%

of design pressure, and they must limit the reactor pressure to no more than 110% of design preseure. Both the high neutron flux scram and safety valve actuation are required to prevent overpressurizing the reactor pressure vessel and thus, exceeding the pressure SAFETY LIMIT. The pressure scram is available as backup protection to the high flux scram. Analyses are performed as described in the " Generic Reload Fuel Application," HEDE-24011-P-A (approved revision number at time reload analyses are performed) for each reload to assure that the pressure SAFETY LIMIT is not exceeded. If the high-flux scram were to fail, a high-pressure scram would occur at 1060 psig.

B 1.2/2.2-3

, O i

EXISTING TECH SPEC l

TS 1.1/2.1 and 1.2 / 2.2 l

' FUEL CLADDING INTEGRITY' and l

' REACTOR COOLANT SYSTEM'

o e l

l QUAD CITIES DPR-29 1.1/2.1 FUEL CLADDING -INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING

-Applicebilitf App 44cobili ty:

The-safety-44mits-establ4thed te pr+sen Th: limit 4ng-safety-tystem-settings apply the-fuel-c4 adding-4ntegrity-apply-te to-tri p-s e ttingt-of-the-ins truments-a n d those-variables +hich-monit+o-the fuel devices-which-are provided-to prevent-the t+erme4-behav4er. fuci-cl a dd ing-integrity-s a fe ty-Hei tt-frem being-exceeded.

Object 4+e: Otrjective---

The-obj ect4ve-ef-the-+a fety-14* it+-46-10 The-obj ect4+e-o f-the-14mi t4 ng-s a f+ty-sy s -

e s te b44 s h-H m i-ts-belev-*h i ch-the-i n t e9- tem-settings-is-to-define-the-leve4-ef - 1 ri ty-of-the- -fue4-c44 dd ing-45-p reserved . the process veriabies 6t wiiicii automat-ic l prettetive-eet4cn-41-4nitfeted-to pre-vent-the-fuel cladding-int +9r4ty-safety 14m4ts-from-be4ng-exceeded.

1 5f M FICATICits

'ihermal Pewer, High Prenare. and.

A. Reseter-Pressurv i 800 prty NUT 57e A. Neutron Flux Trip Settings j flow 2 10% of--RatTU 4/4h Flow _ l Ghawnct be 43 _

__ )

The ex-1: ten : f- Tbe UMi/M fMNb p'Hyft-t-rir//t- i ffylpf /Ath .(MCPRg) less-than 1.06#Mt# l

/111Hpy JJfy shall be as specified below:

s ha 44-<e nst4tu te-v Iole tion-o f-th e fuel cledding-integrity-safety-Hmit. -h - APRM +1ux-Scram-Trip-Set-t4ng ,

with the .REMTOR VGSSEL. PR2550RG {-Run-Hgfe) '

greder mn sccpi9 and core grea.fer- than to% op mfed pcm.ficw 1. Wher, the reaete ade witch-ts 4 in-the Lr. posi-t4em the APRM Hi h u seeaa setting shall be as s own in Figure 2.1-1 and shall be:

,. (flow ,blased) S 1 (.58WD + 62) l *I with a maximum setpoint of 120%

for core flow equal to 98 x 108 lb/hr and greater.

[here S = setting in percent of rated power 1.1/2.1-1 ' Amendment No.127

o 4' QUAD CITIES DPR-29 W = percent of drive flow thquired to produce a rat core flow of 98 million lb/hr. In the event of operation with a lN NN $$5l1 y//// (MFLPD) eea th the p (FRP), the settings // e modified as follows:

FRP Ss (.ssWD + 62) [ MTEFD )

Where:

/f/// 5 MFLPD = J' l , f' /

where the limiting power ,

l density for each bundle is the design l

l for i.

t at cundle.

. The ratio of FRP/MFLPD shall be set equal to 1.gniess the actual operat ng value is less than 1.0 in which case the actual operatirg value will be.

used.

This adjustment may also be L performed by increasing the APRM

! gain by the _ inverse ratio, -

l=

MFLPD/FRP, which accomplishes -

i- . the same degree of protection as ,

L reducing the trip setting by FRP/MFLPD.

tHgh Os*/. Scram)

2. APRM Flux Scram Trip-Setting (RefveHog-w-Stettsp-and-Hot 54edby Model o c reac4or-mode-sw44eh i1rthe-ftefuel or Stectgrtiet Sitndb.y pesition, Uni APRfheram shall be set at 10;; ther er

-equei to 15% of rated neutro flux.

1.1/2.1-2 Amendment No. 114 1

, , . , - - ...m., . . _ -. ., _ ..,.-,, ,-, -,_ . .,,..._i_...___.._________.______.______________.___.______._...m__m, . . ~ . . ,

. a QUAD CITIES DPR-29 3 4M-f4vx Scram-hip-Setthe govj Hon The IRM flux scra etting shall y 35I be set at-less then-cr equd to 120/125 of full scale.

4. When the reactor mode switch is in the Startup or Run position, l the reactor shall not be /

operated in the natural J Therrnal Pcwer, l.ow P/'essac t circulation flow mode. J cr LOLU FlouJ B. Cece-Tfemei-Pc er timi-tHReseter B. APRM Rod Block Setting m--__ _ ,

enn a'os r i t; a a u s w ; vvv r shown in Figure 2.1-1 and shall be :

When-the--ceeeter pressttre-i@1F4-of ps49 _op-coee-f4e: H 1e3Hhan--

catedr the core thermai power shall S < (.58WD + 50) not exceed 25% of /A%d M///// ffMl-with the REAtJc4. VE5sEl p,2EMuR.E- The definitions used above for the ess -than 600 pig or cue flow less thai APRM scram trip apply. In the event to% of nited flod of operation with a /////// ////f//X Y?[PD) jm;fjg(g SAFErf YsrEH 6E'IT/AQ Yp hb5NllJlNll$$bhbb/)}(1fff//$//

greater than the fffff p'ffff (FRP), the setting shall be modified as fallows:

FRP 51 (.58WD + 50) MFLPD_

The definitions used above for the APRM r. cram trip apply.

The ratio of FRP to MFLPD shall be s9t equal to 1.0 unless the actual operating value is less than 1.0, in which case the actual operat{ng value l will be used.

This may also be performed by increasing the APRM gain by the inverse ratio, MFLP0/FRP, which accomplishes the same degree of protection as reducing the trip setting by FRP/MFLPD.

l l

1 1.1/2.1-3 Amendment No. 114

.y

e. CfherRenelcr FYefec}ien Wiem Instrantentaffen

$ctpoinis.

The followinq reaclor palec/lon (SAD CITIES InsfraMenMfion J;elpoinfs DPR-29 shah be mainicined_.'

s C. Power Transient 1.y. eactor low water level scram setting shall be 144 inches above the top of

1. The neutron flux shall not the active uelX at nomal operating exceed the scram setting estab- conditions.

lished in Specification 2.1. A for longer than 1.6 seconds as 1 indicated by the process computer. ,

2. When the process computer is out of service, this safety limit l shall be assumed to be exceeded i l

if the neutron flux exceeds the scram setting established by D. Pe:udw loM Water trVed l Specification 2.1.A and a con- EU_S (nitiation trol rod scram does not occur.

Th D. Reactor Water Level (Shutdown '!KA[e eactor low water level ECCS Condition) A initiation shall be > 84 inches above the top of the active fue1N  !

Whenever the reactor is in the at normal operating conditions. I 1

?>

shutdown condition with irradiated l

-fuel in the reactor vessel, the watei 1evel shall not be less than that corresponding to 12 inches above the

( gggi l top of the active fuel

• when it is seated in the core.

• Top of active fuel is defined to be 360 inches above vessel zero (See i Bases 3.2).

f.2 Turbine stop valve scram shall be <

10% valve closure from full open. ~

h t X3 The scram for tudine control valve fast closure duc to actuation of the fast acting solenoid valve shall be > 460 psig EHC fluid pressure.

'{t % Main steamline isolation valve closure scram shall be < 10% valve closure from full open.~

XTop of active fuel is defined to be 360 inches above vessel zero (See.

Bases 3.2). Reference j 1.1/2.1-4 Amendment No. 129

- - -.,.- --.~. - - -.__- -_.-.--_-_.~ - -_-.-_-_ _

o a N

QUADDPR-29 ClllES E. Maih 5fcuh1hne low 8'e55urc gggictp;en gt;@A i g ( 9 ' @ Main steamline lo4 pressure ini-t tiation of main steamline isolation Qvalve closure shall be > 825 psig. ]

6 % Turbine pressure EHC control scram on lossfluid 106" oil of control 6gc.fW pressure shall be set at greater than c ,

or equal to 900 psig.

L 6 %' Conder.ser low vacuum scram shall be set at > 21 inches Hg vacuum.

.a.

i 1.1/ 2.1- 5 Amendment No.114

o  % 1 1

QUAD CITIES DPR-29 6 gr 1.1 SAFETY LIMIT BASJS grC l The fuel cladding integrity limit is set such that o calculated fuel damage would occur as a result of an abnormal operationalf transient. Because fuel damage is not directly observable, a step-back approach is used to tablish a

$hl/ff JJf.lf such that the /fyift$ ffff ffQ ffffflffisib (HCPR) is not ess than the fuel cladding integrity /Aftyy 7/yfy. MCPR % the fuel cladding integrity I 46///f ///1/1 represents a conservative margin relative to the conditions required l to maintain fuel cladding integrity. l The fuel cladding is one of the physical boundaries which separate radioactive materials from the environs. The integrity of the fuel cladding is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the protection system safety settings. While fission l product migration f rom cladding perforation is just as treasurable as that f rom l use-related cracking, the thermally caused cladding perforations signal a  !

threshold beyond which still greater thermal stresses may cause gross rather than '

incremental cladding deterioration. Therefore, the fuel cladding M//f/ ff/ff is  !

defined with margin to the conditions which would produce onset of //Mtf/Jpg '

Eff//pg (MCPR of 1.0). These conditions represent a significant departure from the condition intended by design for planned operation. Therefore, the fuel cladding integrity f//p// Al/JY is established such that no calculated fuel i damage shall result from an abnormal operational transient. Basis of the values

. derived for this ////f/ ///h f or each fuel type is documented in Reference 1. ,

Thermal iYwm High f'reMarc. ancL. Ho h Flo w '

A. Recetet4eeuw+-+-404-ps4< ped-Cw w+-lhf-htM l

Onset of //#ffff// V/M/6l results in a decrease in heat transfer from I the cladding andytherefore1 elevated cladding temperature and the possibility of cladding failure. However, the existence of critical power, or'NJJ)ftLf/_L_/fr/fvJ / is not a directly observable para:teter in an cperr ing reactor. Therefore, the margin to'ff;/J//Mgfffffp6is

. calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is characterized by the //////// ft//t y//J6 (CPR), which is the ratio of the bundle power which would produce onset of/// of value //ffff/

this///////

ratio divided for anybybundle the actual in the bundle corepower.

The minimum is the ///f///, //fff///

/////- //fJ p' (MCPR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables (Figure 2.1-/).

A.

1.1/2.1-6 Amendment No. 114

a '%

QUAD CITIES  :

OPR-29 The MCPR fuel cladding integrity /g/J/ ///// has suf ficient conservatism to assure that91n the event of an abnormal operational ,

transient initiated from the normal operation condition, more than i 99.9% of the fuel rods in the core are expected to avoid boi & g i soluN4 JuM/Yfyh The margih Letree MCPR of 1.0 (onset of //A41May yp#/4) and the /ggy J/M ', is derived from a detailed statistical analysis considering all of tie uncertainties in monitoring the core -

operating state, incluc correlation Occ c.g. ,(jing uncertainty Reference 1). in the '(AtJ/$Ljffitt/ffsBecause the l correlation is based on a large quantity of full-scale data, there is a l

_ verv high confidence that operation of a fuel assembly at the condition l of MCPf)( the fuel cladding integrity // fry / ////A would not produce

_93 1 boiling transition.

However, if'yf////gLffgf' were to occur, cladding perforation would not be expected. C add ng temperature would increase to approximately 1100'F, which is below the perforation temperature of the cladding material. This had been verified by tests in the General Electric Test Reactor (GETR), where similar fuel operated above the critical heat flux for a significant period of time (30 minutes)  !

without cladding perforation, j If reactor pressure should ever exceed 1400 psia during normal power i operation (the limit of applicability of the'pp//#g[f/Mdffm correlation), it would be assumed that the fuel cladding integrity ftf/H llHJ. has been violated.

In' addition to the 7//pMIW/v///X461irnit (MCPR) constrainedtoamaximumLHGRspecifiedintheCORkoperationis OPERATING LIMITS ,

REPORT for various fuel types. This constraint is established by l h Specification 3.S'X to provide adequate safety margin to 1% plastic strain for abnormal-operating transients initiated from high power 1

conditions. Specification 2.1. A.1 provides for equivalent safety margin for transients initiated from lower power conditions by adjusting the APRM flow-blased scram setting by the ratio of FRP/MFLPD.

y (-s tu te)

Specification 3.5./efablishefthe aximum]which cannot be exceeded under steady power operation.

Thermal (twer, LOLU Prusure. ce Lou) Flou)

B. C$re-ThermM-Powee4imit (Reacter Drass'ir. < Ann p a) _

At pressures below 800 psia, the core elevation pressure drop (0 power, O flow) is greater than 4.56 psi. At low powers and flows (,y.his pressure-differential is maintained in the bypass region oT the core.

Since the pressure drop in the bypass region is-essentially all elevation head, the core pressure drop at low powers and flows will l

I 1.1/2.1-7 Amendment No. 120 i I

W c'Y QUAD CITIES DPR 29 alwags be bundle greater flow, than 4.56 Analyses show that with a flow of 28 psi.pressure drop is nearly independent of x 10 lb/hr bundle bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi driving head will be greater than 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 M/t. At 25% of //M4 XN//6 /////, the peak powered bundle would have to be operating at 3.86 times the average powered bundle in order to achieve this bundle power. Thus, a core thermal power limit of 25%

for reactor pressures below 800 psia is conservative.

C. Power Transient During transient operatiorgthe heat flux (thermal power-to-water) would lag behind the neutron flux due to the inherent heat transfer time constant of the fuel, which is 8 to 9 seconds. Alsuthe %4/J/dtgj

///yff fyg// (g/M IN///hj/ are at values which wirl not allow the reactor to be operated above the /4f//g limit during normal operation or during other plant operating situations which have been analyzed in detail. In addition, control rod scrams are such thagor normal operating transients, the neutron flux transient is terminated before a

' lificant increase in surface heat flux occurs. Coftrol rod scram

aes are checked as required by Specification 4.3./ and the MCPRD operating limit is modified as necessary per Specification 3.5g Eve. ding a neutron flux scram setting and a failure of the control rt ; to reduce flux to less than the scram setting within 1.5 seconds that fuel is daniaged; however, for this does specification, a /Af//// Jf /W. violation will be assumed any time a not necessarily imply neutron flux scram setting is exceeded for longer than 1.5 seconds.

If the scram occurs such that the neutron flux dwell time above the 7/Pff/d/4 /A(/// /////A hMJ/fg is less than 1.7 seconds, the /4(/,t'/

JMAA will not be exceeded for normal turbine or generator trips, which are the most severe normal operating transients expected. These analyses show that $ even if the bypass system fails to operate, the design limit of MCPR the fuel cladding integrity f//pW ///// is not exceeded. Thus, use of 1.5 second limit provides additional margin.

cW46 gn The computer provided has a sequence annuciation program which will indicate the sequence in which scrams occur, such as neutron flux, pressure, etc. This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition existsandthugrovidesomemeasureoftheenergyaddedduringa transient. Thus, computer information normally will be available for analyzing scrams; however, if the e mputer information should not be sis, Leecification 1.1.C.2 will be relied available for any on to determine if a /////y') scram analy$/f has been violated.

1.1/2.1-8 Amendment No. 114

> 6 QUAD u TIES DPR-29 o ReaeJor Ves. sci %fer Level y During periods when the reactor is shut 6down, consideration must also be given to water level requirements due to the ef fect of decay heat.

If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core-cooling capability could lead to elevated cladding i

temperatures and cladd'ag perforation. The core will be cooled sufficiently to prevent cladding melting should the water level be reduced to two-thirds the core height. Establishment of the g4(///

////f at 12 inches above the top of the fuel

• provides adequate margin. This levtl will be continuously monitored whenever the recirculation pumps are not operating.

• Topoftheactivefuelisdefinedtobe360inchesabovevesselzero(Id Bases 3.2).

References

1. "GenericReloadfuelApplicationsgNEDE-24011-P-Ah

@

• Approved revision number at time reload fuel analyses are performed.

1.1/2.1-9 Amendment No. 114

. O QUAD ClllES OPR-29 2.1 LIMITING SAFETY SYSTEM SETTING BASES The abnormal operational transients applic t- to operation of the units have been analyzed t.hroughout the spectrum of pia .J operating conditions in accordance with Regulatory Guide 1.49. In Jution, 2511 MWt is the licensed maximum steady-state power level of the units. This maximum steady-state power level will never knowingly be exceeded.

Conservatism incorporated into the transient analysis is documented in References 1 and 2. Transient analyses are initiated at the conditions given in thesefeferences.

The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by technical specifications. The ef fects of scram worth, scram delay time, and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rapid irsertion of negative reactivity is assured by the time requirements for 5% and 20% insertion. By the time the rods are 60% inserted, approximately 4 (ollars of negative reactivity have been inserted, which strongly turns the transient and accomplishes the desired effect. The times for 50% and 90%

insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady state condition.

The MCPR operating limit is, however, aajusted to account for the statistical variation of measured scram times as discussed in Reference 2 and the bases of Specification 3.5 Steady-state operation without forced recirculation will not be permitted except during startup testing. The analysis to support operation at various power and flow relationships has considered operation with either one or twn recirculation pumps.

The bases for individual trip settings are discussed in the following paragraphs.

For analyses of the thermal consequences of the transients, the MCP8s stated in the CORE OPERATING LIMITS REPORT as the /MI///g /# tim /M f- ////4W/i bound l those which are conservatively assumed to exist prior to in tiation of the transients.

Pcx 1.1/2.1-10 Amendment No. 120

, 6 QUAD CITIES DPR-29 A. Neutron flux Trip Settings

1. APRMfluxScramTripSetting(R/AMMp) IucanrmAIn 4 bolc The average power range monitor ng (APRM) system, which is calibrated using heat balance ata taken during steMy-state conditions, reads in percent 9, /Afd IWfrM pW//. Because fission chambers provide the caH+ input signals, the APRM system responds directly to average neutron flux. During transients (gLhe instantaneous rate of heat transfer f rom the fuel (reactor thermal power) is less than the instantaneous . neutron flux due to the time constant of the fuel.

Therefore, during abnormal operath nai transients, the thermal power of the fuel will be less M 2n that indicated by the neutron flux at the scram setting. Ar lyscs demons +rute that@ith a 120%

scram trip setting, none of the cbnore il operational Transients analyzed violates the fuel y'/#fft Afru, anu there is a substantial margin from fuel <'anage. Theref re, the use of flow-referenced rcram trip proek ;s et a additional margin.

An increase in the APRM scram tr19 tetting would decrease the margin present before the full c.ladd ng integrity (#fff J7V// is reached. The APRM scram trip se'ti.g was determined by an analysis of margins required tu ;rNida a reasonable range for maneuvering during operation. Reat. .ag this operating manpn would increase the frequency of spurious scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected beu.ust; it provides adequate margin for the fuel cladding integrity sa'ety limitgyet allows operating margin that reduces the possibility of unnet'Cesary scrams.

The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of fg /#96

///f r/# // )W/ffM ffh/ p'v#ff/ (MFLPD) and reactor care thermal power. The scram setting is adjusted in accordence with

-a - the formula in Specification 2.1. A.1 when the MFLPD is greater than the ////ffM // /M/d (//// (FRP). The adjustment may be accomplished by increasing the APRM gain by the reciprocal of FRP/MFLPD. This provides the same degree of protection as reducing the trip settings by FRP/MfLPD by raising the initial  !

APRM readings closer to the trip settings such that a scram would be received at the same point in a transient as if the trip settings had been reduced by FF _. FRP/riftro MFA.hD Q 1.1/2.1-11 Amendment No. 114

... . - . . . . - - - - - . . - . . -~ - - . . . . - . - . - _ . .

,. 3-QUAD C111ES

.DPR-29 I c r e u rio d R L l

- 2. APRM Flux Scram Trip Settinc W/#/ or ' '.#p/H// SJ pS'd cecu r' s for operation in the-SM/#p(## v' the rt actor is at low -

pressure, . the APRM scram setti.- - 25% of rated power provides adequate thermal margin between , setpoint and the ,5l4(//,f //,4ff, j 25% of rated. The margin is adequate to accomodate anticipated maneuvers associated with-power plant startup. Effects of

• increasing-pressure at zero or low void content are minoQ, cold q water from-sources available-dt ing startup_is not much colder than at already in the syste temperature coefficients are smal angontrol rod pattern are constrained to be uniform by opera ng procedures backed up by_ the rod worth minimizer. Of all possible sources of reactivity input, uniform control rod

_ withdrawal is the most probable cause of significant power rise.

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods:must be moved to change-power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an

- assumed uniform rod. withdrawal: approach to the scram level,- the rate of-power riseMo more than 5% of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the //,f/)y //AfA. The 15% APRM

- scrao remains active until the code switch is placed in the RN L position. This switch occurs when reactor pressure is greater b +han 825 psig, g

l

(

l-l l

l' L

L L

l 1.1/2.1-12 Amendment No. 114

QUAD CITIES DPR-29

3. IRM Flux Scram Trip Setting logic The IRM system consists of ei t chambers, four in each of the reactor protection system V+ channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are broken down into 10 ranges, each being one-half a decade in size.

The IRM scram trip setting of 120 divisions is active in each range of the IRM. For example, if the instrument were on Range 1, the scram setting would be 120 divisions for that range; likewise, if the instrument were on Range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accomodate the increase in power level, the scram trip setting is also ranged up.

The most significant sources of reactivity change during the power increase are due to control rod withdrawal. In order to ensure that the IRM provides adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed.

This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor 's just subtritical and the IRM system is not yet on scale.

Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod i' bypassed.

The results of this analysit show that the reactor is scrammed and peak power limited to 1% of rated povar, tht.s matntaining MCPR above the fuel cladding integrity /N#f MrM. . Based on the above analysis, the IRM predes protection against local control rod withdrawal errors and continuous withdrawal sf control rods in sequent ovides backup protection fer the APRM.

Setting B. APRM Rod Bloc Reactor power level may be varied by cemoving control rods or by varying the recirculation flow rate. T/e APRM system provides a control rod block to present gross rod withdrawal at constant recirculation flow rate to protect against grossly exceeding the MCPR /uel fladding /nteg-rity S///// L//#. This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal.

The flow variable trip setting provides substantial margin from fuel setting, over the damage, assuming aflow entire recirculation steady range. state operation The margin at tothe thetrip / #4/4/J' /)'y increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst-case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences, and is moni-tored continuously by the incore LPRM system. As with APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the

//N/M ffNJfM $$ lif$$$M $U4/ WNN}) exceeds the l'f)/1,%h 6l ///f/

fpff/, thus preserving the APRM rod block safety margin. As w//4 jhe scram . setting , this may be ace.omphshed. by adjustahg th e.

R PIU1 gains .

1.1/2.1-13 Amendment No. 114

, ~4 QUAD CITIES DPR-29 .

t.. Other Reaclor Profec.fior) Systern InskamenfafNn SelfulnN l p'. Reactor Low Water Level Scram The reactor low water level scram is set at a point which will assure that the water level used in the bases for the fgAy 7//M is maintained. The scram setpoint is based on normal operating temperature and pressure conditions because the level instrumentation is density compensated.

O p'. Reactor Low Low Water Level ECCS Initiation Trip Point The emergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident ({pnd to limit fuel cladding temperature to well below the cladding meltTng temperature to assure that core geometry remains intact @nd to limit any cladding metal-water reaction to less than 11 To accomplish their intended function, the capacity of each emergency core cooling system component was established based on the reactor low water level scram setpoint. To lower the setpoint of the low water levei scram would increase the capacity requirement for each of the ECCS components. Thus,thereactorvesseGlowwater levehscram was set low enough to permit margin for operatTon, yet will not tW set lower because of ECCS capacity requirements.

The design of the ECCS components to meet the above criteria was dependent on three previously set parameters: the maximum break size,

' the low water level scram setpoint, and the ECCS initiation setpoint.

To lower the setpoint for initiation of the ECCS could lead to a loss of effective core cooling. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or during normally expected transients.

.L E'. Turbine Stop Valve Scram The turbine stop valve closure scram trip anticipates the pressure, neutron flux, and t4+e heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of 10%

of valve closure from full open, the resultant increase in surf ace heat flux is limited such that MCPR remains above the MCPR fuel cladding integrity /#gff //% ven during the worst-case transient that assumes the turbine ass is closed.

1.1/2.1-14 Amendment No. 114

QUAD CITIES DPR-29 3 /. Turbine Control Valve Fast Closure Scram The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control y lves due to a load rejection and subsequent f ailur.e, of the bypas

• i.e. , it prevents MCPR from becoming less than the MCPR fuel cladding integrity Mf/// T/Aff for this transient. For the load rejection without bypass transient from 100%

power, the peak heat flux (and therefore LHGR) increases on the order of 15% which provides wide margin to the value corresponding to 1%

plastic strain of the cladding.

gn akr- ftun or eqaal +o Thetripsetpointoff460psigEHCfluidpressurewasdevelopedto ensure that the pressure switch is actuated prior to the closure of the turbine control valves (at approximately 400 psig EHC fluid pressure yetassurethatthesystemisnotactuatedunnecessarilguetoEHCg system pressure transients which may cause EHC system pressure to momentarily decrease.

. 6 A. Reactor Coolant low Pressure Initiates Main Steam Isolation Valve Closure The low pressure isolation at 825 psig was provided to give protection against fast reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram featur6phich occurs OPERATICNRL"in the RM+#dijwhen the main steamline isolation valveFare closed to provide for reactor shutdown so that operation at pressures lower than those specified in the thermal hydraulic M(/f/ /Wff does not occur, although operation at a pressure lower than 825 psig would not necessarily constitute an unsafe condition.

4 K. Main Steamline Isolation Valve Closure Scram cecan rionA L.

The low pressure isolation o the main steamlines at 825 psig was provided to give protection ainst rapid reactor depressurization and the resulting rapid cooldow f the vessel. Advantage was taken of the scram feature in the RM fMpjwhich occurs when the main steamline isolation valves are closed j to provide for reactor shutdown so that high power operation at low reactor pressures does not occur, thus providing protection for the fuel cladding integrity fM/M J/' ///.

Operation of the reactor at pressures lower than 825 psig requires that the reactor mode switch be in the S#ff# positior,, where protection of the fuel cladding integrity /gf,tf 7/AfA is providad by the IRM and APRM high neutron flux scrams. Thus, the combination of main steamline low pressure isolation and isolation valve closure scram in the RM %

///hassurestheavailabilityofneutronfluxscramprotectionoverthe entire range of applicability of fuel cladding integrity /g/M ////jt.

In addition, the isolation valve, closure scram in the RM ff anticipates the pressure and fit x transients which occur uring normal l or inadvertent isolation valve losure. With the scrams . set at 10%

valve closure in the RM //, here is no increase in njutron flux.

[the 4fennonm l

1.1/2.1-15 Amendment No. 129

l' QUAD CITIES DPR-29 f/. Turbine EHC Control Fluid Lodessure Scram The turbine EHC control system operates using high pressure oil. There are several points in this oil system where a loss of oil pressure could result in a fast closure of thu turbine control valves. This fast closure of the turbine control valves is not protected by the turbine control valve fast closure scram, since f ailure of the oil system would not result in the fast closure solenoid valves being actuated. For a turbine c valve fast closure, the core would be protected by the APRM and i pressure scrams. However, to provide the same margins as provi ed for the generator load rejection on fast closure of the turbine control valves, a scram has been added to the reactor protection system which senses failure of control oil pressure to the turbine control system. This is an anticipatory scram and results in reactor shutdown before any significant increase in neutron flux occurs. The transient response is very similar to that resulting from the turbine control valve fast closure scram. The scram setpoint of 900 psig is set high enough to provide the necessary anticipatory function and low enough to minimize the number of spurious scrams. Normal operating pressure 'or this system is 1250 psig.

Finally, the control valves will not start until the fluid pressure is 600 psig. Therefore, the scram occurs well before valve closure begins.

4, /. Condenser Low Vacuum Scram Loss of condenser vacuum occurs when the condenser can no longer handle the heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise and an increase in surface heat flux. To prevent the cladding /Af/f/ //4f4 from being exceeded if this occurs, a reactor scram occurs on turbine stop valve Kwimagclosure in the Pg/Mp. The turbine stop valve closure scram function alone is adequate to prevent the cladding /Afgty llAf4 from being exceedecgin the event of- a turbine trip transient with bypass closure.

The concenser low vacuum scram is anticipatory to the stop valve closure scram and causes a scram before the stop valves are closed and thu Scram occurs in t RM a

1. 1. qthe resul}j.pg transient is less severe.at 21-incMig vacuum, stop valve closu vacuum, and bypass closure at 7-incgg vacuum, cPes.nricdRL 1.1/2.1-16 Amendment No. 114

QUAD CITIES DPR-29 References

1. " Generic Reload fuel Applicatio @ NEDE-24011-P-A.*

• Approved revision number at time reload analyses are performed)

2. " Qualification of the One-Dimensional Core Transient Model for Boiling V!ater Reactors", General Electric Co. Licensing Topical Report NED0 24154 Vols. I L and II and NEDE-24154 Volume III as supplemented by letter dated September 5,1980 from R.H. Buchholz (GE) to P.S. Check (NRC).

l .s' -

1.1/2.1-17 Amendment No. 114

QUA0 CITI ,

DPR-29 APRH Flow Reference Scram and APRH Rod Block Settings 4

L til

' ^

lil- q 111-lil-i- -

3 + R0 200 Itt

t li-T + R0 EDn 5 3

~

j 51; - DLO KUn w

j Sl; --- DLO E00 LL I al-11-li-llJ l

g , , , i > > > >

g gg ;g 3g al il 61 11 Il 11 til ill 121 RECIRCULATION LOOP FLOW (1 of rated)

FIGURE 2.1-1 Amendment No. II4

1 A. .

Figure 2.1-2 has been deleted.

c Amendment No.114

e a QUAD CITIES ,

DPR-29 )

I 140 -

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/ 100/100 PCV11/Pt0V LINE to / g Opetatty Regtes $vppetted j g / Sy N.I.D.C. - 26167 and f N.I.8.0. 2219:

$ *0titattes sa $tatte 14ep or R . o ,,, Natural Citewastles is

. Lisited per Test. Spets.

L t.R.3 sad 2.1. A.6 203 PL'MP 3PI D LINI **0peration at greater

' than rated core flow is supported by N CC-31&&9 1AfD C00171CNS P0vt1 2311 MWth COLI FLOW ll K1be/51 0 8 ' ' i 0 20 60 60 80 100 120 Hi CORE FLOW RATE (% of RATED)

Figure 2.1-3 (SCHEMATIC)

APRM FLOH BIAS SCRAM RELATIONSHIP TO NORMAL OPERATING CONDITIONS Amendment No. 114 i

t

l QUAD-CITIES OPR-29 1.2/2.2 REACTOR COOLANT SYSTEM SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING AppHeebi44ty. Appi b iaiiity.

Applies-to-44mi ts-on-teac tor-coolant Appli es-to-trip-s ettings-o f-the-ins ttts-system-pressuee. m ents-a n&dev ices-which-a re-p rovide&t o p rev en t-the-re eetomys ttmrTa f vty-timits

-f rom-being1xceeded.

Objective. Objective.

To-es te bMsh-a-Mst-bel o w-which-tAe-in- Tc define-the-4evel-of-the-process-vads teg r4ty--of--the-reactor-coMant-+y+ts 4 s emes-e t-wMeh-ets tome tie-prettet-ive-et -

not-th re a te ne d 4ue-to-auve r-Pre & &w - sion-i s-i n iMa ted-to-peevent-the--safe ty

-condi tion. Iimitt-frem-beisg-exceeded SPE41fICATIONS-

n. Rea&r t. octan} .54 skm Presure .SeHing n, w&r ecclant 54km Wesm Mng The reactor coolant system pressure Reactor coolant high pressure scram X. as measured by the vessel steam space

')K.

shall be at <1060 psig.

pressure indicator shall not exceed set -

1345 //J$.et ar.y time when i-r+ad4ted.

f.a une

-is -y ,- + u. . . +. s. e.

i % ., s o . .

r e a r_ +

n ,-

,,m e ssa

% Primary system safety valve nominal settings shall be as follows:

1 valve at 1135 psig II) 2 valves at 1240 psig 2 valves at 1250 psig 4 valves at 1260 psig (1) Target Rock combination safety / relief valve The allowable setpoint error for each valve shall be % 11%.

1.2/2.2-1 Amendment No. 114

- a QUAD CITIES DPR-29 1.2 SAFETY LIMIT BASES The reactor coolant system ir,tegrity is an important barrier in the prevention of uncontrolled release of fission products. It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

The pressure (4//f/ //Aff 45 psigDas measured by the vessel steam space pressure indicaton)is equivalent to 1375 psig at the lowest elevation of the reactor vessel. The 1375 psig value is derived from the design pressures of the reactor pressure vessel and coolant system pipina. The respecti Q design The pressure

///pff/pressuresare1250psigat575*Fand11756t560F.ffff was chosen the applicable design codeyp) ASME Boiler and Pressure Vessel Code Section 111 for the pressure vessel", and USASI B31.1 Code for the reactor coolant system piping. The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10% over design pressure (110% x 1250 = 1375 psig), and the USASI Code permits pressure transients up to 20% over design pressure (120%

x 1175 = 1410 psig). The /////f ///J/. pressure of 1375 psig is referenced to the icwest elevation of the reactor vessel. The design pressure for the edne &n rec 4r+. suction line piping (1175 psig) was chosen relative to the reactor vessel design pressure. Demonstrating compliance of peak vessel pressure with the ASME overpressure protection limit (1375 psig) assures compliance of the suction piping with the USASI limit (1410 psig). Evaluation methodologytoassurethatthis////,t//////pressureisnotexceededfor any reload is documented in Reference 1. The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig; this is a factor of 1.5 below the yield strength of 40,100 psi at 575 F. At the pressure limit of 1375 psig, the general membrane stress will only be 29,400 psi, still safely below the .

yield strength.

The relationships of stress levels to yield strength are comparable for the primary system pipina and provide similar margin of protection at the

.- establishedpff,t/ fressure)///d of the For 4;he@

The normal operating pressure reactor coolant system is 1000 psig.

turbine trip or loss of electrical load transientX, the turbine trip scram or generator load rejection scrarGtogether with the turbine bypass systeg limitX pressure to approximately 1100 psig (References 2, 3, and 4). In addition, pressure relief valves have been provided to reduce the probability of the safety valves operating in the event that the turbine bypass should fail.

1.2/2.2-2 Amendment No. 114

QUAD CITIES i

DPR-29 Finally, the safety valves are sized to keep the reactor vessel peak pressure below 1375 psig with no credit taken for relief valves during the postulated ful1 J ospre of all MSIVs without direct (valve position switch) scram. Creditfjs taken for the neutron flux scram,,(howeverj The indirect fluxscramandsafetyvalveactuation[provideadequatemarginbelowthe allowable peak vessel pressure of 1375 psig.

Reactor pressure is continuously monitored in the control rooguring operatiogon a 1500 psgfull-scalgressure recorder.

References

1. " Generic Reload fuel Applicatio h NEDE-24011-P-A*

2. SAR, Section 11.22

3. Quad Cities 1 Nuclear Power Station first reload license submittal, Section 6.2.4.2, February 1974.

4. GE Topical Report NEDO-20693, General Electric Boiling Water Reactor $cloael)

No.1 Licensing submittal for Quad Cities Nuclear Power Station Unit 2, December 1974.

• Approved revision number at time reload analyses are performed.

-O' -

l 1.2/2.2-3 Amendment No. 114 1

-O e QUAD CITIES DPR-29 2.2 LIMITING SAFETY SYSTEM SETTING BASES In compliance with Section 111 of the ASME Code, the safety valves must be set to open at no higher than 103% of design pressure, and they must limit the reactor pressure to no more than 110% of design pressure. Both the high neutron flux scram and safety valve actuation are required to prevent overpressurizing the reactor pressure vessel and thu@xceeding the pressure

/4fff//////.. The pressure scram is available as bacTup protection to the high flux scram. Analyses are performed as described in the " Generic Reload fuel Application," NEDE-24011-P-A (approved revision number at time reload analyses are performed) for each reload to assure that the pressure safety limit is not exceeded. If the high-flux scram were to fail, a high pressure scram would occur at 1060 psig.

1.2/2.2-4 Amendment No. 114

. -c i

SIGNIFICANT HAZARDS CONSIDERATIONS AND ENVIRONMENTAL ASSESSMENT EVALUATION PROPOSED TS 1.1/ 2.1 and 1.2 / 2.2

" FUEL CLA.DDING INTEGRITY" and

' REACTOR COOLANT SYSTEM"

EVALUATION EQE flIG_NIylgANJ llAZARDS CONSID.KEATION PROPOSED SPECIFICATION 1.1/2.1 AND 1.2/2.2 FUEL CLADDING INTEGRITY / REACTOR COOLANT SYSTEM The proposed changes provided in this amendment request are made in order to provide a more user friendly document, incorporate desired technical improvements, and to incorporate some improvements from later operating BWRs. These changes have been reviewed by Commonwealth Edison and we believe that they do not present a significant Hazards Consideration. The basis for our determination is documented as follows:

BASIS f_OB Q ILQ SIGNIFICANT HAZARDS CONSIDE3ATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards consideration. In accordance with the criteria of 10 CFR 50.92(c) a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility, in accordance with the proposed amendment, would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated, because:

The proposed changes to Specifications 1.1/2.1 and 1.2/2.2 to delete the present Applicability and Objective sections represent administrative changes to format and presentation of material. The proposed changes provide the user with a format that will allow quicker access to needed information and provides concise Safety Limit, Limiting Safety System settings, Applicability and Action requirements. The additions of Applicability and Action requirements represent clarification of intended requirements that do not presently state all required conditions of operability or provide clearly stated Action statements if the requirements are not met. The added requirements follow STS guidelines that are in use at many operating BWRs with similar design and operating configurations as Quad Cities Units 1 and 2. Operability requirements for Safety Limits have been chosen to reflect only those Operational Modes where the Safety Limits apply.

Operability requirements for Limiting Safety System Settings are already stated in other sections of the Quad Cities Technical Specifications, thus reference to the appropriate operability requirement is made rather than repeating the requirement in the Limiting Safety System Settlng Specification. The requirement prohibiting, operation in the natural circulation flow mode is not a Limiting Safety System Setting and has teen moved to Specification 3.6 J on Recirculation Pump Flow Requirements. The change to the setpoint for the reactor low water level scram is made to provide consistency with other Technical Specifications and to l

e .

i provide a setpoint tolerance, in a conservative direction, to allow for instrument drift and accuracy.

The proposed changes do not alter the intent of existing setpoints or accident assumptions and follow existing requirements at other operating BWRs for operability and Action statements. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2) Create the possibility of a new or different kind of accident from any previously evaluated because:

The proposed administrative changes to the format and arrangement of material do not affect technical requirements or assumptions of any potential accident and; therefore, cannot create the possibility of a new or different kind of accident from any previously evaluated.

The proposed addition of Applicability and Action requirements enhance the understanding and usability of the Technical Specifications and thus represent an improvement over present specifications. New requirements are modeled after those in use at operating BWRs and do not represent requirements that will adversely affect potential accident analyses or assumptions. The proposed move of the prohibition on operation in the natural circulation flow mode from the Limiting Safety System Settings to a Limiting Condition for Operation, maintains the present restriction on operation while clarifying the nature of the requirement. Therefore, the proposed changes do not create the possibility of a newThe or different kind of accident from any previously evaluated.

proposed change to the reactor low water level scram setting only allows movement of the cetpoint in the conservative direction which is presently allowed in Section 3.1/4.1 for this same setpoint.

3) Involve a significant reduction in the margin of safety because:

The proposed administrative changes to format, arrangement of material, clarification of requirements and other non-technical changes do not affect any safety aspects of the plant and as such can not involve a significant reduction in the margin of safety.

The proposed Applicability statements require availability of Safety Limits and Limiting Safety System Settings when required to perform their respective functions. Proposed Actions for Safety Limits allow only 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to be in Hot Shutdown and then reference Specification 6.4 to ensure that proper reports are made and restart is prohibited until l

e s approved by the NRC. These provisions help ensure that present margins are not significantly reduced.

The proposed change to the reactor low level scram setting does not reduce the margin of safety since the setpoint is only allowed to move in the conservative direction and the change is made to provide consistency with allowances for this same setpoint in Section 3.1/4.1.

e i ENVIRONMENTA_h MLS_Eg8]iENT EVALUATIQH PROPOSED SPECIFICATION SECTION 1.1/2.1 FUEL CLADDING INTEGRITY AND SECTION 1.2/2.2 REACTOR COOLANT SYSTEM Commonwealth Edison has evaluated the proposed amendment in accordance with the requirements of 10 CFR 51.21 and has determined that the amendment meets the requirements for categorical exclusion as specified by 10 CFR 51.22 (c) (9) .

Commonwealth Edison has determined that the amendment i involves no significant hazards consideration, there are no significant change in the types or significant increase in the amounts of any effluent that may be released offsite, and there is no significant increase in individual or cumulative l occupational radiation exposure.

The proposed amendment does not modify the existing facility and does not involve any new operation of the plant, i As such, the proposed amendment does not involve any change  ;

in the type or significant increases in effluents, or increase individual or cumulative occupational radiation  !

exposure. The proposed amendment to Section 1.1/2.1, " Fuel Cladding Integrity" and 1.2/2.2, " Reactor Coolant System" contains administrative changes such as including appropriate applicability statements within the specifications to define the applicability of operating mode and the required actions to be implemented in the event that the specification cannot be met. The added requirements are based on Standard Technical Specifications and later operating plant requirements. The proposed specification also relocates one requirement-to provide for better operator usage of the document.

1

i 1

QC-lL/ QC-2 -DIFFERENCES- '

TS . l .1/ 2,1 and 1,2 / 2.2

' FUEL' CLADDING INTEGRITY" and REACTOR' COOLANT SYSTEM' e

l 9

• W w ve- *r -p e--

f 1 COMPARISON OF UNIT 1 AND UNIT : TECHNICAL SPECIFICATIONS FOR THE IDENTIFICATION OF TECHHICAL DIFFERENCES SECTION 1.1/2.1 FUEL CLADDING INTEGRITY Commonwealth Edison has conducted u comparison reviGW of

- Unit 1 and Unit 2 Technical Specifications to identify ~

technical differences in support of combining the

.hnical Specifications into one document. The intent of

.4e review was not to identify any differences in presentation style (e.g. table formats, use of capital 16tters, etc.), punctuation or spelling errors, but rather to identify areas which the Technical Specifications are technically or administratively different.

The review of Section 1.1/2.1 " Fuel Cladding Integrity" revealed the following technical differences:

The third paragraph, last sentence on page 1.1/2.1-6 (DPR-29) states " Basis of the values derived for this safety limit for each fuel type is documented in Reference 1." The Unit 2 Technical Speclfications states, " ...is documented in References 1 and 2." NEDO-24259-A (Reference 2) contained information concerning the use of barrier fuel. The latest revisionlof NEDE-24011-P-A contains the information regarding barrier fuel which was previously-only contained in NEDO-24259-A. As a result, reference 2 can be deleted.

The last sentence of paragraph B on page 1.1/2.1-9 (DPR-

30) states, "As with the scram setting, this may be accomplished by adjusting the APRM gains." This information is not contained in the Unit 1 Technical Specifications. The Unit 2 Technical Specification information will be retained in the combination since the information is consistent with the requirements of Limiting Safety System Setting 2.1.B.

The turbine control valve fast closure scram which is contained in section 2.2.F for both units' Technical Specifications are different due to a current design difference. A Technical Specification amendment to the Unit 2 Technical Specifications has been submitted to be reflect an upcoming modification which will result in a consistent design to both units.

e %

Several administrative differences were identified as follows:

Pace 1.1/2.1-3 2.1.B Unit 1: The APRM rod block setting shall be shown...

Unit 2: The APRM rod block setting shall be as shown...

Unit 1: This may also be performed by...

Unit 2: This adjustment may also be performed by...

Page 1 1/2.1-6 Paragraph 1 Unit 1: the fuel cladding integrity safety limit. MCPR...

Unit 2: the fuel cladding integrity safety limit MCPR Paragraph "A: Unit 1: (CPR), which is the ratio of the bundle...

Unit 2: (CPR), which is the ratio for the bundle...

Pace 1.1/2.L-11 1 rate of pow'ar rises no more...

Paragraph Unit 1:

Unit 2: rate of poter rise is no more...

Pace 1.1/2.1-13 Paragrap. Unit 1: reactor protection system low channels.

Unit 2: reactor protection system logic channels.

Unit 1: sequence provides backup...

Unit 2: sequence and provides backup Paragraph "B" Unit 1: Reactor power level may be varied by removing control rods Unit 2: Reactor power level may be varied by I moving control rods Pace 1.1/2.1-14 Paragraph "E" Unit 1: neutron flux, and the heat flux increase...

Unit 2: neutron flux, and heat flux

! increase...

l l

l

, - . . - ,,, 1 .

e %

Pace 1,1/2.1-15 Paragraph "H" Unit 1: Main Steamline Isolation Valve closure Scram Unit 2: Main Steamline Isolation to Valve Closure Scram U.mit 1: entire range of applicability of fuel cladding...

Unit 2: entire rand of applicability of the fuel cladding...

Eggg 1.1/2.1-16 Paragraph "I" Unit 1: are several points in this oil system where a loss of oil...

Unit 2: are ceveral points in this oil system where a lost of oil...

r a l

l I

COMPARISON OF UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS FOR THE-IDENTIFICATION OF-TECHNICAL DIFFERENCES-SECTION.1.2/2.2 REACTOR COOLANT SYSTEM Commonwealth Edison has conducted a comparison review-of the Unit 1 and Unit:2 Technical Specifications to identify any technical differences in support of combining the-Technical' Specifications into one document. The intent of the review was not to identify any differences in  !

presentation style (e.g. table formats,.use of-capital '

letters,_etc.), punctuation, or spelling errors but rather to

-identify-areas which the Technical Specifications are technically or administratively _different.

The review of Section-1.2/2.2 " Reactor Coolant System" q did not reveal any technical differences. .Several j administrative differences were identified as follows; i Pace 1,2/.2.2-1 Title Unit 1:-1.2/2.2 Reactor Coolant System-Unit 2: _1.2/2.1 Reactor Coolant System

2. 2. A~ Unit'1: Reactor coolant high-pressure scram shall be at < 1060 psig Unit 2: . Reactor coolant high-pressure scram shall be < 1060 psig i

L 2.2.B Unit 1: The allowable setpoint error for each l valve shall be at + 1%

Unit 2: The allowable setpoint error for each valve shall be + 1%

D Unit 1: Target Rock combination safety relief valve is designated by (1)

Unit 2: Designation _(1) is defined but not indicated on-the valve listing Page 1.2/2.2-2 Paragraph _2: Unit 1: ... design pressures are 1250 psig and 575 F and 1175 at 560 F.

Unit 2: ... design pressures are 1250 psig and 575 F and 1175 psig at 560 F L ' Paragraph 4 Unit 1: _ The normal operating pressure reactor

coolant system...

1. ' Unit 2: The normal operating pressure of the E reactor coolant system...

l:

u l-

e s

• ' 4 MAR 17N MARIE 7"FA f 'll 0/ ')/]

MARTIN MARIETTA ENERGY SYSTEMS, INC. yf"CEgy,Q3t 37m June 28,1991 Distribution Draft Task 3 Technical letter Report Enclosed is the Draft Task 3 technicalletter report for FIN L-1647-0, National Profile on Commercially Generated Low-Level Radioactive Mixed Waste. If you have any comments, please contact Chad Glenn, Nuclear Regulatory Commission by telephone (301-492-0567) or by facsimile [(301) 492-0260] by July 12.

We thank you for your continued interest in this project.

Sincerely,

,q. \ '

J. A. Klein JAK:sim Enclosure Distribution M. Alissi, EEI R. Alvarado, TLLRWD W. Dornsife, PDER C. Glenn, NRC L Hendricks, NUMARC R. LaShier, EPA C. Owens, EG&G T. Plummer, DOE /EM C.Pugh L Tripoli, Afron

,L;)h,fN m-,c ,

v .:. a u q J h )& './1Jl '

\

/\lL1l

I NATIONAL PROFILE ON COMMERCIALLY GENERATED LOW-LEVEL RADIOACrlVE MIXED WASTE TECllNICAL IEITER REPORT FOR TASK TIIREE FIN L-1647-0 June 30,1991 Contractor: Oak Ridge National Laboratory

• NRC Program Manager: C. J. Glenn EPA Program Managers: R. LaShier J. Ekx)d

• Managed by Martin Marietta Energy Systems, Inc., for the U.S. Department of Energy under contract DE AC05-840R21400.

1 1

1 CONTENTS f.!!EE E X ECUTIV E S U ht ht ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v

1. I NTR O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1 OVERALL OlUECTIVE OF hilXED WASTE STUDY . . . . . . . . . . . . . . . . . . 1 1.2 WO R K TO D E PE R FO R h1 E D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.3 W'1RK TO DE PERFORMED IN TASK TilREE . . . . . . . . . . . . . . . . . . . . . . . 2 1.4 D EFI N ITI O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.5 PURPOSE OF TECilNICAL LE* ITER REPORT . . . . . . . . . . . . . . . . . . . . . . . 3

2. TAS R Ti l R E E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.1 DEVF. LOP A PLAN TO COLLECP AND ANALYZE MIXED WASTE DATA...................................................... 3 2.1.1 Obicejives of S t ud y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.1.2 Study Desien Snefi fica t io n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.1.3 Soccification of Collection Methods ................. .......... 4 2.1.4 Da t a R ea uire rnen t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.1.5 Ouality Centrol of Data Collection . . . . . . . . . . . . . . . . . . . . . . . , . . . . 6 2.1.6 Ouality Control of Data Processinc . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.1.7 D a t a Descript ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2.1.8 D a t a Ta bula t io n a nd An n ivsis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 P

2.2 MIXED WASTE SURVEY PRETEST . . .......................... 13 2.2.1 P re t es t Oues t ion n a ire . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 2.2.2 Plan for Ad ministerin e Pretest . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 APPENDIX A. STUDY DESIGN SPECIFICATIONS APPENDIX B. QUESTIONNAIRE - NATIONAL PROFILE ON MIXED WASTE 11i

v

• EXECUTIVE

SUMMARY

This technicalletter report details the findings resulting from Task 3 of the U.S. Nuclear Regulatory Comrnission and U.S. Environmental Protection Agency-sponsored project entitled National Profile on Commercially Generated Im Level Itadioactive Mixed Waste. The overall objective of the work is to compile a national profile on the volumes, characteristics, and treatability of commercially generated mixed waste.

The Tash 3 report presents the detailed plan necessary to collect and analpe the required mixed waste data that will allow the compilation of the national mixed waste profile. This report includes a detailed statement of the study's objectives, specification of the survey design, and a description of the mixed waste data base that will be developed. A description of the revised data gathering instrument and a plan for pretesting the data gathering instrument are also included in this report.

v

1 NNI10NAL PROIM.E ON COMhiliRCIA11Y OliNERNFED  :

LOW. LEVEL RADIDAC11VE hi!XED WAFill TECIINICAL IRI'lTiR RIIPORT FOR TASK TilREll FIN I 16174)

June 30,1991 I

1. INTRODUCTION I 1.1 OVERALL OBJECTIVE OF h!!XED WASTE STUDY

• the overall objective of this U. S. Nuclear Regulatory Commission (NRC)- and U. S.

Environmental Protection Agency (EPA)-sp(msored project is to compile a national profile on the volumes, characteristics, and treatabilhy of commercially generated low level radioactive mixed waste (htW) by cicarly defined generator categories.

1.2 WORK TO BE PERFORh1ED The information collected and assembled in this project will be used by NRC and EPA, with  !

assistance from the U. S. Department of Energy (DOE) and the iiut State Technical l Coordinating Committee for low level Radioactive Waste Disposal (TCC), to compile a national profile on the volumes, characteristics, and treatability of commercially generated mixed waste.

In general, the study consists of the following eight tasks:

1. Evaluate existing available htW information from past surveys conducted by llost States, Compacts or other parties, summarize the results and summarize the lessons learned from past survey reports (See Technice' Letter Report for Task One, November 16, 1990).

2. Determine the adequacy of existing data to estimate and project the volumes, characteristics and treatability of htW (See Technical letter Report for Task Two, hiarch 31,1991).

3. Develop a plan to collect and analyze htW data, develop a pretest questionnaire.

4. Administer the pretest, develop the Gnal survey questionnaire, and Gnalize the overall survey design.

5. Collect and analyze htW data,

6. Compile a national htW profile,

7. Identify available treatment technologies to reduce to Best Demonstrated Available Technology (BDAT) levels or climinate the hazardous component of specific hiW streams.

1

2

8. Document the study results in a NRC NUREG report.

13 WORK TO DE PERFORh1ED IN TASK TilREE Task Three consists of the development of the detailed plan necessary to collect and analyze the htW data that will allow for the compilation of the national h1W profile. This task will include a detailed description of the study's objectives, specification of the survey design, specification of the data collection methods, as well as speciGention of the data requirements and a description of the mixed waste data base that will be developed. A description of the revised data gathering instrument and a plan for pretesting the data gathering instrument will also be detailed in this task.

1.4 DEFINITIONS For purposes of this project, mixed waste (h1W) is defined as " waste that satisGes the dennition of low-level radioactive waste (LLRW) in the low-Lesel Radioactive Waste Policy Amendments Act of 1985 (LLRWPAA) and contains hazardous waste that (~) is listed as hazardous waste in subpart D of 40 CFR Part 261 or (2) causes the LLRW to exhibit any of the hazardous waste characteristics identified in Subpart C of 40 CFR Part 261".

The LLRWPAA defines LLRW as " radioactive material that (a) is not high-level radioactive waste, spent nuclear fuel, or byproduct material as defined in section lle. (2) of the AEA (i.e.,

uranium or thorium mill tamngs) and (b) the NRC classiGes as LLRW consistent with existing law and in accordance with tM".

In addition, the following are included in the definition of hazardous materials for the purpose of this study:

~

o Oils and oil sludges.

e Other materials classiGed as hazardous by the Resource Conservation Recovery Act (RCRA) authorized states.

Commercially generated MW, for the purposes of this report, includes all MW generated by facilities that would normally send any LLRW to one of the three existing LLRW disposal facilities. This dennition would therefore include all generators of htW except the DOE facilities.

3 1.5 PURPOSE OF TECilNICAL LE1TER REPORT j The purpose of this technical report is to present and document the detailed data collection I and analysis plan resulting from Task *Ihree. In addition, the draft questionnaire used to collect the data is presented. This is the same questionnaire that will be used in the pretest.

2. TASK TliREE 2.1 DEVELOP A PLAN TO COLLECT AND ANALYZE hilXED WASTE DATA  !

1he principal objective of this subtask is to develop a plan that will be used to collect and j analyze the data acquired in support of the development of a national profile of MW. l i

2.1.1 Obiectives of Study l

The overall objective of this study is to determine the volumes and characteristics of h1W l l

typically generated for facilities in major generator categories. The specific parameters for the study are as follows:

• National MW volumes are to be determined within a factor of two for tmth 1990 annual h1W generation rates and, if possib!c, the total quantity of MW in storage at the end of 1990.

• This factor of;wo will also apply to MW volumes for each of the major generator categories.

The categories will include utilities, medical facilities, industry, academic institutions, and government.

• The radiological characteristics that need to be acquired will include the LLRW Class (A. U, C, etc.), as defined in 10 CFR 61.55, and a listing of the major nuclides present.

• The hazardous characteristics will include the EPA code (D. F, K, P, or U series), if applicable, and a common name descriptor.

• In addition, information will be acquired that allows for a determination of how MW streams are generated and any plans for reducing or climinating the various streams.

• Information en how the various MW streams are presently being treated, stored, and/or disposed of will also be acquired.

2.1.2 Study Design Specifications The information that was to be included in this section has been provided by David C. Cox &

Associates and is attached as Appendix A.

e a 4

2.13 Speci0 cation of Collection Methods e Sample collection methodology. The mixed waste survey will be a mailed questionnaire with telephone follow up. Although respondents are requested to return the completed questionnaire by mail, the follow-up telephone calls are added to increase the res[onse rate and to resolve any questiens the respondents may have in filling out the questionnaise. The questionnaire is presented in Appendix B. It is hoped that such a procedure will produce a return rate of at least 75%

o Time to administer survey. It is anticipated that the survey will take one week to mail out.

Initial follow-up phone calls will begin three weeks later and take approximately one month.

The second cycle of phone calls should also take about a month adding up to a minimal estimate of three months to complete the survey. Data tabulation and analysis (see Sect. 2.1.8) will take additional time, e ConGdentiality. NRC and EPA will draft a statement of intent that will indicate the purpose of the study and also that it is not the federal intent to kiok for noncomplinnce areas. Data responses will be submitted to ORNL, and only ORNL will have access to the raw data.

Survey data will be provided to the NRC and the EPA stripped of any facility identification.

The available information will only be published m aggregate form. No individual facilities will be identiGed.

• Coordination with states / compacts. A preliminary letter (drafted by and cosigned by NRC and EPA), accompanied by a copy of the survey questionnaire, will be mailed to all the compacts, the states within the compacts, and to each of the unaligned states. The letter will inform them of the purpose of the survey and enlist their aid in helping to ensure that the compilation of the national survey will be a meaningful and creditable undertakir4

• Advance notification to prospective survey participants. No formal notification of those selected for the pretest or the actual survey is planned. It is anticipated that a number of facilities will be aware of this survey and its purpose from the various initial contacts we will have made with the states / compacts. Prenotification would entail at least two more weeks of time and an additional mailing of approximately 1600 letters. (De mailing of 1600 letters is tentative subject to findings in the pretest and the final construction of the sampling frame.)

e Follow-up. Three weeks after mailing to the individual facilities, follow-up telephone calls will begin by ORNL to: (a) encourage the rapid return of the questionnaire, (b) answer any questions, and (c) either acquire the needed information over the telephone or arrange for a

. o  !

$ I future time to acquire the data over the telephone. If, after two additional weeks, no response is received from the above procedure, an additional telephone call will be made in an attempt to obtain the data or to arrange for a time when the information can be collected. ,

No additional telephone calls will be made except on a case-by-case basis.  !

• Format for questionnaire. The draft questionnaire, which is identical to the oa- being used for the pretest,is included as Appendix B.

e Data rights. Information supplied by the respondents will only be used for statistical tabulations, and individual replies will be retained by ORNL 1t is anticipated that summary data will be used by a number of agencies for a wide variety of purposes, but that the interests of the respondents will be protected to the maximum extent possible.

2.1.4 Data Reacirements The following is a listing of all the data required to meet the objectives of this study, in a few cases, additional information is requested (such as LLRW volumes) to serve as a point of reference, to judge the accuracy of the response, to aid in the data analysis, or to provide a useful piece of information that will be relatively easy to collect. Data requirements can be broken down into three areas: (1) general information, (2) h!W volumes and characteristics, and (3) treatment and disposal. The survey instrument (questionnaire - see Appendix D) has been developed to ensure that all the following data requirements are met.

e General Information

• Facility name and address.

• Name, title, and phone number of person completing form.

• Date completed.

• Facility category (nuclear power plant, medical facility, academic institution, industry, and government). Finer category breakdowns will also be included where applicable.

• NRC/ agreement state license number.

• EPA id fication number. EPA facility classification (large quantity generator, small quantity, conditionally exempt, etc.).

• Volume of LLRW generated in 1990.

• hiixed Waste (hiW)

• htW stream description.

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6

• Annual amount of htW generated in IVA)(before and after on site treatment).

• Cumulative amount of hiW stored as of 12/31fAl.

• Generation of waste streams (how htW streams are generated).

e llazardous characteristics EPA hazardous waste codes.

Chemical or common description of hazardous waste and concentration if appropriate.

• Radiological characteristics LLRW class.

hiajor radionuclides and activities.

• Waste oils and materials declared hazardous by RCRA authorized states to be specifically included.

• Treatment and Disposal e Plans for reducing and/or climinating htW streams.

• Present treatment rnethods On-site.

Off site.

• Untreatable htW requiring ultimate disposal, in addition, although not part of the data requirements, the survey form will include extensive instructions, definitions. and tables containing the EPA hazardous waste codes as well as an accompanying letter explaining the purpose of the survey.

2.1.5 Ouality Control of Data Collection Quality centrcel of the entire spectrum of data collection activities will be essential to ensure a product that meets the requirements placed on the National Pro 6te resulting from this study.

Data collection quality control elements will include:

• Specification of collection methods. The complete specification of the collection methods that will be used to collect all the required information necessary to meet the objectives of this study (as listed under section 2.1 A Data Requirements) is a necessary element in the maintenance of a high quality data collection effort. Prior and complete specification allows for the timely review of the proposed collection process prior to the actual data collection.

The complete specification is included in section 2.1.3 of this document.

e e 7

e Training of data collection specialists. Individuals handling the follow up telephone calls will need to le briefed as to the various acceptable options that can result from both the first and 1 second follow-up telephone call, how to collect the required technical information if it is to be received over the phone, and how to make preliminary quality control checks c f the data.

It is anticipated that the number of indisiduals requiring this training will be limited in number (1 to 4).

e Preliminary checking of 'as received data'. All received data will be initially checked by a principal investigator for obvious errors and inconsistencies, misinterpretation of instructions, incomplete data, and clarifications. Unacceptable responses will require a follow up telephone call for resolution.

• Possible use of computer assisted telephone inteniewing (CATI) packages. Although CATI is a highly recommended technique for acquiring data during a telephone inteniew, this survey will consist of a questionnaire with a request to return by mail. It is hoped that only a minority of respondents will choose to supply the required data in the follow-up telephone calls. It is not anticipated that CATI will be used.

e Technical evaluation of data by principal investigators. All data will be reviewed by the principal investigators to determine if the data "makes sense" at a technical level. Questions include:

• Are the indicated waste streams consistent with the type of operations performed by that facility?

• Are the reported waste streams volumes consistent with other data we may have on that particular facility from past surveys?

e Spot checking of data for completeness and validity. This quality control element is very similar to other elements discussed here. All returned survey questionnaires will have both a preliminary check for completeness and any obvious errors, and a more thorough technical evaluation. Both of these checks will be made by a principal investigator.

• Follow.up data collection for incomplete or missing data. Any unacceptable responses due to incomplete or missing data or the misinterpretation of directions will require a follow-up call to acquire the missing data or to rectify any obvious errors.

2.1.6 Ouality Control of Data Processing The various processes involved in the coding of data from the survey instruments to the data processing media (personal computer, main frame computer, etc.) need to be defined and

. v l

8 l followed to ensure citor. free data processing. Data processing quality control elements include:

• Doubic checking of all data coding. Two data entry personnel will check all data as coded into the data processing media to ensure correct data entry.

• Spot checking of coded data by principal investigators. Coded data will be checked by a principal investigator for obvious errors. Depending on the quantity of data, there will be at least a spot check and hopefully all encoded data will be checked.

• Computer error checking on input data field and range definitions. Data fields will have format type and range definition restrictions to ensure that " nonsense" numbers do not get encoded in the data base. l 2.1.7 Data Description Data will be provided on personal computer (PC) floppy disks of NRC an't EPA speciiications and in hard copy. The PC format will include a compiled executable program with menu-driven data and output selection routines. Determination of output seicction options will be finalized in consultation with NRC and EPA. In addition, all data will be provided in ASCII format that will allow the user to employ any data base management system of their choice.

2.1.8 Data Tabulation and Anahsis Data collected during this survey will be presented in two formats. Where feasibic, information will be presented as tabular listings. The shells for a number of such listings are illustrated in Tables 2.1-2.4. When it is not practical to present the information in tabular form (e.g., information on how various was:e streams are generated), the data will be incorporated into specific sections in the various survey reports.

Tables 2.1 and and 2.2 will be completed for all major generator categories. The need for more detailed tabulations by subcategories of generators or hazardous streams will depend on the amount of data received.

Table 2.3 will be used to estimate the amount of mixed waste generated within each compact and/or state. For those compacts in which the estimated standard error is unacceptable, other techniques (such as modeling based on the amount of LLRW or the number of LLRW shippers) will be used to determine the predicted amount of mixed waste generation.

Other tables (including Table 2.4) will be developed, as necessary, to show the pertinent survey results. Information from these table = will be necessary to demonstrate the completeness of the acquired oata and to extrapolate the data to the national profile.

i' Table 2.1. National mixed waste profile

, (generation and storage) 4 Generator category Academic Institution -

4 Estimated Estimated Estimated annual amount annual amount cumulative generated in 1990 generated in 1990 nmount stored before on-site treatment after on-site treatment as of 12B1/30

(cu ft/ year) (cu ft$ tar) (cu ft) i-IIazardous stream a

i Organics Liquid scintillation fluids i

Waste oils and oil sludges Other organic solvents Chlorinated fluorocarbons (CFCs)

Other organics M etals l Lead Chromates j

Mercury Other metals i

Aqueous corrosives Acids Bases TOTAL i

I

T l t t

t t

~

I. f i

j Table 2.2. National mixed waste profile l (treatment) l ' Generator category i Acadcrcie Institution  !

l I i

l Estimated Estimated Estimated '

! annual amount annual amount annual amount j generated in 1990 generated in 1990 generated in 1990 I l treatabte on-site treatabic off-site requiring ultimate disposal (cu ftlycar) (cu It% car) (cu ftlycar)

Hazardous stream i i

Orgames -

! Liquid scintillation fluids l

. Waste oils and oil sludges -

i' Other organic sohents j

Chlorinated fluorocarbons (CFCs) ,

l Other organics t i Metals  !

Ixad l Chromates j Mercury l Other metals i I

i 9

Aqueous corrosives l Acids Bases t 1

TOTAL <

4 i

t

(

e d

i I

.i'

m _ - s - - - - . - _ . _r . m,e

. s I1 i

Table 2.3 Estimated annual amount of htW generated in 1990 (cu ftlycar)

[ Estimated standard error]

Generator category Power plant hiedical Academic industrial Government Compact / State Northeast Appalachian Southeast Central States hiidwest Central hiidwest Rock) hicuntains Southwest Northwert Texas New York Others

. t l

12 I

Table 2.4 National mixed waste sursg Ocncrator category Power plant Medical Academic Industrial Government Nurnber in sampling frame Known or potential generators LLRW shipper lists Other Number recching questionnaires Known or potential generators LLRW shipper lists Other Number returning questionnaires Known or potential generators LLRW shipper lists Other Number reporting MW Known or potential generators LLRW shipp(r lists Other

. t e ,

13 2.2 hilXED WAS PE SURVEY PRETEST 2.2.1 Pretest Ouestionnaire The pretest questionnaire to be used in this mixed waste study is identical to the draft questionnaire in Appendix D. It is possible (based on the results frorn the pretest to be conducted in Task Four) that some modifications may be made to the final questionnaire, especially in the area of instructions and/or definitions.

2.2.2 Phn.for Administerint Pretest The pretest candidates will be 20 25 facilities selected from a presently available list of 47 known or potential htW generators in the Appalachian Compat. Appalachian Compact Users of l I

Radioactive Isotopes (ACURI), the association of radioactive licensees within the compact, has agreed to cooperate in the initial testing phase of the national MW survey. Due to the wide variety of facility types within the Appalachian Compact, this group offers an appropriate forum for a valid pretest. Six nuclear power facilities are located in the Appalachian Compact, and these will all be included in the pretest. The remaining 15-20 pretest facilities will be selected at random from the known or potential htW generator list with allowances made for obtaining a broad mix within each of the four remaining major generator categories.

The pretest collection methodology will follow, as much as possible, the methodology te be used for the actual s'arvey, A preliminaiv letter (drafted by and cosigned by NRC/ EPA),

accompanied by a copy of the pretest survey, will be mailed to the Appalachian Compact, ACUR1, and each of the Appalachian Compact member states to inform them and enlist their aid in the survey. ORNL will mail individua! surveys, along with the preliminary letter mentioned above, to each of the selected pretest facilities. Two weeks after the mailing to the individual facilities, telephone follow up calls will begin by ORNL to either encourage the rapid return of the questionnaire, to answer any questions, to acquire the necessary data over the phone, or to arrange for a subsequent telephone call to acquire the data.

The administering of the pretest is on a tight schedule. As such, the initial mailing of the

, preliminary letter is planned for the latter half of July, the package of individual pretests during early August, and the follow-up phone calls during mid August. All pretest surveys need to be returned by the end of August.

I APPENDIX A. STUDY DESIGN SPECTII~1 CATIONS Provided by David C Cox & Associates

. i.

NATIONAL PROFILE OF MIXED WABTE GENERATORS SURVEY DESIGN DOCUMENT DRAFT July 1, 1991 Prepared by:

Arnold Greenland David C. Cox & Associates 1620 22nd Street, N.W.

Washington, DC 20008 Prepared for:

U.S. Environmental Protection Agency Permits and State Programs Division Office of Solid Wasto U.S. Nuclear Regulatory Commission Decommissioning and Regulatory 1ssues Branch Division of Low-Level Waste Management and Decommissioning Under Contract No. 68-DO-0061, Task 1-10 Exposure Evaluation Division Office of-Toxic Substances EPA Project Officer: Edith Sterrett EPA Task Managers: Richard Lashier and Jared Flood NRC Task Manager: Chad Glenn

e ts l

1. Introduction This document is intended to discuss the issues relating to the statirtical design of a national survey of commercially generated mixed waste. The objective of the survey is to compile a profile both at the national level and by certain broad classes of establishments the volumes, characteristics and treatability of commercially generated nixed waste. Because of the technical nature of the definition of mixed waste, the reader is directed to the " Technical Letter Report for Task Three" developed by Oak Ridge National Laboratory (ORNL) for a full definition. In brief, mixed waste is material which is both Low-Level Radioactive Waste (LLRW) under the Atomic Energy Act and its amendmenes and a hazardous waste under the Resource Conservation Recovery Act (RCRA).

The key goals of this document are:

+ to characterize the target population;

. to write down the specifics of a sampling plan;

• to describe the details of the data collection plan;

. to describe plans for dealing with survey and sampling errors; and

+ to lay the foundation for the estimation process which will follow the data collection process.

The sections which follow will uddress each of the goals in turn.

This document is labeled " DRAFT" because certain data files required to estimate sample sizes and complete descriptions of some of the sampling frames were not available as of this writing. When that information is available, revised estimates of the population and sample will be made and a final draft of this report will be issued.

2. Characterization of the Target Population The unit of investigation for this study is defined as an establishment in the United States which has a potential to generate mixed waste. Since an establishment could have mixed waste on its site only if it was licensed by either the NRC or one of the Agreement States, we can certainly restrict the target population to such establishments. It is reasonable to suggest also that we further limit the target population to establishments who also have permits or interim status under RCRA. We do not make this limitation because it is possible that an establishment generates mixed waste, for example the emission of a hazardous substance from a piece of equipment on the premises, but is not required to have a permit under RCRA.

Many of the establishments having licenses from the NRC or an 1

. -. -- - - - -_~.-_- - . - _ _ - - - - - - - - - -. .- - .-

e s l i

Agreement State could not, by the nature of their business, be i generators of mixed waste. For example, from the NRC list of I approximately 8,000 establishments, only about 1,700 could  !

reasonably generate mixed waste. This group was determined by I eliminating any establishment on the list which had a Material License Program Code (a field available on the NRC data base) which indicated a type of establishment which, according to the judgment j of cognizant technical personnel, should be excluded. A table j showing the specific Material License Program Codes which were excluded will be included in the final draft of this report.

Although analogous data files to that used for NRC states could not be obtained for the Agreement States, the definition for the target populatiori remains "all establishments on either the NRC or Agreement States-lists which, because of the nature of their business, have a chance of generating, either by design or by accident, any mixed waste." We will describe this population as the " potential generators of mixed waste."

Within the population of potential generators of mixed waste, there will be wide variation regarding the likelihood of generating mixed waste. In particular, utilities are very likely candidates  !

to generate such wastes because of the volumes of LLRW which are l generated on such sites. Therefore we must segment the larger population of establishments into smaller groups from which to 1 select the establishments which will ultimately be included in the sample. Since etAimating the volume of mixed waste is the primary goal of the survey, groupo which have greater potential to generate substantial volumes will be more likely to be included ir. the survey.

~Along these lines, there are several key breakdowns of the target population that require discussion. First there exists a '

list of establishments which was developed by oak Ridge National Laboratory from an earlier stage of this study which contains "likely" generators of mixed waste. This list contains many major generators of mixed waste including all utilities holding NRC licenses, and is planned to be sampled with certainly for this survey. More will be said about this list later.

A second key breakdown of the population of interest is by type of establishment. The main establishment types are:

Utility Medical

. Academic Industrial Government These are specific broad classes of establishments which were mentioned in the introduction to this document for which accuracy requirements will be made.

The population of interest, the potential generators of mixed waste, can be viewed now as broken down by the two variables just 2

. _ = _ - __ - .. -- . _ . . - . . . _ . _ _ _ - - ~ _ _ _ _ - . . . . . ~-

, 6 9 4 Not on ORNL On ORNL List List TOTAL Uti2ities 57 0 57 Medical 54 942 996 Academio 91 782 873 Industrial 138 2323 2461 Government 30 376 406 Other 60 -

60 TOTAL 430 4423 4853 i

Exhibit 1. Potential Mixed Waste Generators  ;

mentioned. Exhibit 1 contains a tabular display of the population of interest. The different establishment types are shown as row )

headings and whether the establishment is on the ORNL list or not  ;

is indicated as shown as the column heading. As will be discussed in the next section, this breakdown of the population will define the stratification of the population for purposes of sampling and estimation.

l

3. Sampling Plan This section will discuss several components of the design of the survey. They include:

. stratification; l

. sample size determination;  ;

. sampling frame; and '

. sampling procedure.

3.1 StratificatiqD There are two basic reasons for stratification. The first is to fulfill the requirement for producing estimates within subgroups of the population at a predetermined level of accuracy. This requirement is present in the Mixed Waste Survey with respect to the types of establishments shown in Exhibit 1 above. This high level stratification constitutes the set of estimation cells or l estimation groups which will be used in this survey.

The second reason to introduce stratification into a survey design is to optimize the accuracy of the estimates ultimately l produced by the survey. This is accomplished by selecting subgroups within the population which are similar with respect to their characteristics and with respect to the quantities being estimated. For example, a critical group of establishments is the set of all shippers of mixed waste who do not already appear on the 3

. 6 2

ORNL list. Because this group of establishments already ships j LLRW, they are considered to be much more likely to be generators of mixed waste than other groups in the population. Also, because  :

they are more likely to generate mixed waste, the variability of

• the amount of mixed waste generated in this group (measured by the standard deviation) is likely to be higher than other groups. ,

Sampling practice dictates that survey resources should be concentrated in those segments of the population in which the variability of the key estimates (total volume of mixed waste in this case) is the highest. Following that practice will accomplish two important goals. First it will result in overall estimates of the total mixed waste which are more accurate. Second it will use the financial resources of the survey project in the most cost ef fective manner by concentrating the surveys among establishmonts which are most likely to provide the NRC and EPA with useful information.

EXHIBfT 2. MW SURVEY TARGET POPULATION ORNL Shippor's Other NRC Potontial Other Agroomont TOTAL List Ust Mixed Wasto State Potential Excluding Generators Mixed Wasto ORNL List Generators With EPA Wahout With Withou i Permit EPA EPA tEPA Permit Permit Permit Utilities S7 - - - - - 57 Modcal 54 342 120 80 240 160 996 j Academe 91 -182 1 20 80 240 60 873

• Industrial 138 823 300 200 600 400 2461 l Govemment 30 76 69 40 120 80 406 Other 60 - . - - - 60 TOTAL 430 1,423 600 400 1200 800 4853 Exhibit 2. Breakdown of the Population for Sample Stratification See Exhibit 2 for a breakdown of the population of interest into the primary estimation cells and, within that, a further breakdown into cells which will be used for producing the most accurate national estimates in the most cost effective manner. The substratification cells defined in that exhibit are the following:

a Shippers List Excluding the ORNL list. This list I contains all shippers of LLRW who do not already appear 4

. w on the ORNL list. Outside of the ORNL list, this group is considered to be the next most likely group to generate mixed waste, a Other NRC Potential Mixed Waste Generators. This is the group of establishments having NRC licenses and Material License Codes which are considered to be potential generators of mixed waste. This group is further broken down into those with and wi':heut a EPA Permit to treat, store or dispose of hazardoas waste. The EPA list also includes generators of hazardouc waste.

u Other Agreement State Potential Mixed Waste Generators.

This group is the analogous group to NRC category above.

Although there is no list which categorizes these establishments by material license codes (like the NRC list), they are defined as part of the population for completeness. It is expected that estimates representing this part of the population will be produced in the final results of the survey. They are also shown in Exhibit 2 as broken down by having and not having an EPA permit.

3.2 Samole Size A sample size determination is made using several key facts about the survey. Those include the number of sample units (establishments with potential for generating mixed was+.e) which are in each of the population strata, estimates for the variances of the total volume of mixed waste within each stratum, and the accuracy requirements which are being imposed on the survey.

Target sample size estimates are generally related to ensuring that the sampling error experienced in the survey is within the accuracy requirements imposed by the designers of the survey. The impact on the sample size estimates of non-sampling errors (to be discussed below) is factored in generally as a multiplier of the base sample size estimate.

For the purposes of this computation we will use a tampling error requirement of plus or minus 10% of the survey estimate. The remaining roughly 50% to 100% of estimates, which we are using as the accuracy requirement for this survey (to remain within a factor of two of the actual mixed waste volume), will be required to insure that the impact of non-sampling error is restrained to acceptable levels.

Consider first the table of means and standard deviations which are shown in Exhibit 3. We show in the first two columns of this exhibit, the mean and standard deviation obtained from a small

" subgroup" of six states included in an earlier Technical Letter Report produced the Oak Ridge National Laboratory for this project.

The volumes of mixed waste included in this subgroup represents different surveys with different selection criteria. However, they are the best data we have from which to make mean and standard 5

. 6 Subgroup Subgroup Estimated Estimated Mean STD DEV Mean STD DEV Utilities 76 140 75 150 Medical 78 85 75 100 Academic 107 402 100 250 Industrial 15 34 75 75 Government - -

75 50 Other 0.14 .52 - -

Exhibit 3. Estimates of Population Means and Standard Deviations deviation estimates for the total volume. The last two columns of the table are the " estimated" mean and standard deviations. These estimates are an adjustrent of the numbers obtained from the six state subgroup. The most dramatic deviation from the subgroup parameters was the use of a standard deviation of 250 cubic feet of mixed waste rather than 407 cubic feet. This was done because using such a large standard deviation had a dramatic impact on sample allocation, and represents a "best guess" of a more realistic estimate.

In order to allocate the sample to the various strata in a way which accomodates the different levels of variability that are present in the population, we have made the following assumptions.

Within the shipper's list, the standard errors for each of the estimation cells are those which appear in the latter two columns of Exhibit 3. The other substrata within the population are all assumed to have a standard deviation which is a multiple of that number. They differ from the base figure because experts in this field indicate that these groups have very dif ferent likelihoods of generating mixed waste. Our assumed fractions from the base are the following:

ORNL List 125%

NRC or Agreement State within EPA List 40%

NRC of Agreement State off of EPA List 20%.

Using the means and standard deviations just described, we employed Neyman allocation methods to compute a sample size of 1,256 and to allocate that sample to the cells shown in Exhibit 4.

As Exhibit 4 demonstrates, Neyman allocation tends to concentrata the sample in those segments of the population in which the volume is the highest and, at the same time, the variability in that volume is also the highest. Exhibit 4 also shows more establishments in the academic category under the Shipper's List 6

. 6 ORNL Shippor's Other NRC Other Agroomeri TOTAL Ust Ust Potontial Mixod Stato Potontial Mixod Excluding Wasto Generators Wasto Gorxwators ORNLUst With Without With Without EPA EPA EPA EPA Permit Permit Permit Pomut Utilitios 57 - - - - -

57 Medical 54 186 21 7 42 14 324 Acadernic 91 201 42 14 85 28 461 Irxiustrial 138 97 11 4 23 8 281 Government 30 21 5 2 11 4 73 Other 60 - - - - -

GO TOTAL 430 505 79 27 161 54 1,256 Exhibit 4. Sample size allocation including Agreement States column than are available in the population numbers of Exhibit 2.

This discrepancy will be adjusted in subsequent Exhibits.

Exhibit 5 represents a reallocation of the sample into groups from which sampling is technically feasible for this survey. In particular, the sample allocated to the agreement state list cannot be accomplished as illustrated in the table. Therefore, those cases are allocated to other parts of the target population proportionately. The resulting target sample sizes by sampling stratum are shown in Exnibit 5. The total sample sizes are the same; however, no sample cases are shown under the Agreement State group because no lists to support that selection could be obtained.

As will be discussed in the section below on sampling frames, we will use results from the shipper's lists alone to characterize the Agreement State component of the target population. Note also that Exhibit 5 is adjusted so that the number of academic establishments in the Shipper's List column does not exceed the number available.

The excess sample was, again, allocated to other columns to preserve the total sample required for the estimation cell.

Exhibit 6 represents the estimates of the sample size taking into consideration the impact of non-response on the final set of completed interviews. Such non-response has two affects. One is that is lowers the sample of available cases to be used for estimation, thus lowering the accuracy of the estimates. Second, 7

e s ORNL Shipper's Ottu NRC Ottu Agroomont TOTAL Ust Lht Potontial Mixed Stato Potential Mixed Excluding Wasto Generators Wasto Gonorators ORNL Ust With Without With Witixmt EPA EPA EPA EPA Permit Perma Perma Permit Utilitios 57 - - - - - 57 Medical 54 235 26 9 - - 324 Academic 91 182 139 49 - - 461 Industrial 138 124 14 5 - - 281 Govemment 30 32 8 3 - - 73 Other 60 - - - -

,- 60 TOTAL 430 573 187 66 - - 1,256 Exhibit 5. Sample size allocations excluding Agreement States ORNL Shippor's Other NRC Ottu Agrurmont TOTAL ust Ust Potential Mixed Stato Potential Mixed Excluding Wasto Generators Wa::to Gonorators ORNL Ust Wah Wahout Wah Wahout EPA EPA EPA EPA Pormit Permit Pormit Permit Utilities 57 - - - - - 57 Medical 54 313 35 12 - - 414 Academic 91 182 230 81 - - 584 Industrial 138 165 19 7 - - 329 Govomment 30 43 11 4 - - 88 Ottu 60 - - - - - 60 TOTAL 430 703 295 104 - - 1,532 Exhibit 6. Non-response adjusted (Final) sample size allocation since those who participate essentially select themselves for participation or not, there could be a subtle bias creeping on the estimates representing the different between those who respond versus those who choose not to respond. Since we expect a 75%

response rate, the numbers in Exhibit 6 were obtained from Exhibit 5, by dividing each estimate by 0.75. This sample size adjustment 8

l l

l can compensate for the fact that the number of cases who respond would be too low, but it cannot. compensate for the bias due to non-response. We must assume here that the group of responders are similar to the non-responders with respect to volumes of mixed waste generated, so that a non-response adjustment is possible.

Again in the case of academic establishments in the Shipper's List column, the number of cases selected for inclusion in the survey exceeded the number in the population (see Exhibit 2). In that case, the sample to be selected was set equal to the number in the population and the excess cases were allocated into other cells of the survey.

3.3 S3molina frame The sampling frame is intended to be a complete physical list of the entire target population for this study. In practice, obtaining such lists, either in computer readable form or in a hard copy list, is often difficult; and this particular survey is not an exception to this rule. The definition of the target population is all " potential generators of mixed waste." The word potential was added to the definition to exclude establishments which, because of the nature of their operations, could not generate mixed waste. Although, work has been done with the NRC list to exclude such establishments, comparable work for the Agreement states was not able to be done because of the lack of a complete list.

The set of lists that were available for use as all or part of the sampling frame include the following:

W The Oak Ridge List of Likely MW Generators (the ORNL List). This list contains 430 establishments including all utilities. The list was formulated during the preliminary work done by ORNL on this project, and represents a group of establishments which are very likely to generate mixed waste. Since generating an estimate for total MW generated is the main goal of this survey, this frame will be sampled entirely, a The Shipper's Lists. These are actually three separate lists of establishments which ship LLRW to one the three sites licensed to handle such waste, one each in the states of South Carolina, Washington, and Nevada.

Computer readable lists for each of these sites has been requested from state authorities. At this writing, only the South Carolina list has been delivered, but the others are expected. As a check on the computer data bases that will be obtained, several hard copy lists are available, a The NRC Licensee Data Base. This list contains a complete accounting of all NRC Licensees, and as such is the most complete source for the population of mixed waste generators for NRC states. However, since it is 9

. o 1

believed that most of the roughly 8,000 licensees would not be potential mixed waste generators, other lists will be matched against the NRC list and used for the survey.

Also, there is information on the NRC data base (namely, the Material License Program Code) which allows the 8,000 cases on that file to be reduced to approximately 1,700 l potential mixed waste generators. From this point forward, reference to the NRC list will mean the list of 1,700 potential generators.

m The Agreement State Licensees. There does not exist a list analogous to the NRC list of licensees for Agrooment States. Even when specific states or compacts had lists, they proved to be only in hard copy, of different formats or dif ficult to obtain. Therefore, the idea of obtaining a complete such list to use as part of the frame for this survey was abandoned.

u Hazardous Waste Data Management System (HWDMS) and l Resource Conservation and Recovery Information System  ;

(RCRIB). These data bases contain information about i establishments which have permits to treat, store or  !

dispose of hazardous waste under RCRA. The HWDMS is an  ;

older data base which is being replaced by RCRIS. At the present time, only eight states are available in the RCRIS format, the remainder being obtained from HWDMS.

In either case, information relating to name, address, phone number, etc. will be available for both establishments permitted to treat, store or dispose of hazardous waste as well as those who are generators of

, such waste. These files will-be available in computer readable form and (including the generators) include some 150,000 establishments.

Based upon the availability of the frames mentioned and in congruence to the sampling stratification described above, we are proposing the following approach to creating a sampling frame for this survey. First, all establishments on the ORNL list and the Shipper's Lists will be matched against the NRC data base on the NRC side. On the Agreement State side only the ORNL list will be crossed with the Shipper's Lists. All remaining cases will be put into one the following unique groups:

(1) the ORNL list; (2) tha Shipper's Lists (excluding any cases on the ORNL list); and (3; the potential generators on the NRC list which are not on either the shipper's lists or the ORNL list.

This sampling frame does not include the Agreement State list.

However, it is known that a great many of the establishments on the ORNL List and the Shipper's lists are from Agreement States. The recommendation is that the Shipper's lists and ORNL list alone be

(

used for establishments which fall in Agreement States. At the l time of data analysis, it will be possible to compare the 10

-- . - . _ ~ . - - = .- .-. _-_ - _ - - . - - - ..

. o relationship between the group of establishments on the NRC side which are on or off of the Shipper's lists. Should there be any appreciable difference detected at that time, an adjustment can be made to estimates.

The selection of and construction of thia sampling frame is echoed in the sample allocation which is shown in Exhibits 5 and 6 of this document.

3.4 Samolina Drocedure Central to the sampling procedure is that each case included in the survey be selected with known probability. Such a sample is called a " probability. sample. " Without a probability sample, it is not possible to produce estimates of total volumes or other estimates from the survey which can ce i .flated to represent the entire population of interest. Therefore aperational activities relating to the sample selection enueavor to preserve the probabilities of selection.

First of all, each establishment (unit of sample selection for the survey) must appear in one and only one sector. Consider first the ORNL list. This list is planned to be a certainty sector.

This means that each establishment in this group will appear in the survey with probability one. However, these establishments must be purged from other segments of the survey. Therefore, the first task-in the sampling procedure (as described briefly in the section on sampling frames) is to match various segments of the population to other segments. The ORNL list must be identified and extracted from other lists.

Finally, for the purposes of segmentation and sub-stratification, we must cross the NRC file with the RCRIS or HWDMS filen to determine which establishments fall into which of the EPA-defined selection cells.

The second major component of sample selection is to select _a simple random sample within each of the strata according to the sample size numbers shown in Exhibit 6. With computer lists, this

! is generally done using the concept of systematic sampling. This l procedure is implemented by determining the samplitig interval (the number of cases between successive establishments to be included in the survey), selecting the first case at random within the first sampling interval, and picking each " nth" case from there forward, i where n, here, is the sampling interval.

4. Data collection methodology This data collection methodology selected for use in this survey is a mailed out survey with telephone follow-up. The surveys will be mailed out and respondents will be allowed approximately three weeks to respond before a telephone follow-up will be made.

11

i

, s t . .

The follow-up call will consist of two parts. The first part will be a reminder to fill out the survey form. The second part will be an offer either to collect the information over the phone at the time of the call or to schedule a call in the future to cellect the information by phone. Should those who promise to send the questionnaire in by mail not fulfill this promise within three weeks of the first call, a second call will be made to collect the data or schedule the collection by phone. Such a protocol has been shown to include a response rate that approaches 75% of the cases selected.

5. Assessing and controlling errors one of the r.ost critical aspects to designing a survey is preparation for errors. The two main categories of error which cre9p into surveys (whether censusee or samples) are sampling and non-samplin.g errors. The former refers to the error in estimates which occurs because not all of the cases in the population were used in msking the estimate. This is the type of error which can be handled the easiest. Statistical methodology has been developed to the point where such errors are easily quantifiable and estjec&es of the impact of such errors can be made. Further, based upon assumpilons of the type used in the section on sample size compwation, we can plan for providing enough sample to make sampling errors within predetermined bounds.

The other type of error, non-sampling error, is much more difficult to estimate or control. Such errors include:

1. Non-response bias.

2. Frame bjas.

3. Response bias ' ! ying , misunderstanding, answering a different question).

The first of these, non-response bias, was discussed briefly above.

This type of error exists because not all of the cases selected it itially are willing to participate in the survey. The usual approach to handing this error is to plan the operations of the survey and the survey instrument to minimize the existence of this type of problem.

A second very important way to improve response relates to how the respondent is contracted and whether he can be convinced that it is in that establishments best interest to respond. Therefore, trade organizations and other industry groups that could have an influence on response will be contacted to provide supporting letters to be either mailed separately to the sample or will be included as an attachment to the main mailout.

The problem relating to frame bias results when the sampling frame does not match the target population exactly. The main problem occurs when cases which are potential generators of mixed waste are excluded from the frame. In that case the estimate for volume is underestimated. Other frame problems include errors in 12

. o the information on the lists (e.g., wrong address in a mail survey), duplicate entries on the file, definition of a unit on the frame not matching the definition in the target population (e.g.,

different uses of the term, " establishment"). Dealing with frame bias is generally very costly in person hours, but if it is not done, can r , ult in errors in the final estimates.

The issue of response bias relates to whether the respondents answer the questions intended. With respect to this survey, the.re is some concern that the respondents may not be fully cognizant cf the definition of mixed waste and could clain that they do not generate mixed waste (a situation that could exclude them from the survey) when in fact they do.

The impact of response bias is best mitigated by very careful design of the survey instrument. This includes including questions which require the respondent to provide answers to different questio , which would tend to indicate the same key information.

For exa._fle we can ask: "Does your establishment generate mixed waste? Yes or No?" Later in the questionnaire, questions about types of operations at facilities which tend to be correlated with the generation of mixed waste can be asked. Answers to such questions can be used at the analysis stage to judge the quality and accuracy of responses.

A second way to reduce response bias is to carefully pretest the questionnaire. Since OMB clearance has been obtained, there is a plan for a protest of at least 25 establishments of varying types and sizes. Results of the pretest will be used to optimize the survey instrument.

6. Estimation Estimates of total mixed waste generated and other quantities collected in the survey will be produced for each estimation cell.

All survey estirators will be self-weighting using weights attached to each respondent's data record at the time of the creation of the sample. Estimates of totals (for example the total mixed waste generated) will take the following form:

l*h,$,x Y.i.k*

i M .J.k*

1 M W1.1,k is the response of the kth establishment in the jth where stratumy,*('jkh column in Exhibit 2) of estimation cell i (the ith row in Exhibit 2). WGT is the sampling weight associated with the establishment, and the NRAF is the stratum's unit non-response adjustment factor.

The WGT for each stratum is defined as the reciprocal of the probability of selection. This number is the quotient of corresponding cells in Exhibit 2 to those in Exhibit 6 (the population number divided by the sample number) .

l 13

. a The NRAF is computed as follows:

L,u.wari.s.,

NRAF,y u Eusable WGT3 ,j,y where the term " viable" in the formula indicates that the sum should include all units (k) in stratum j and estimation cell i which are in scope for the survey. This would only exclude establishments which were found to be out of business or otherwise outside of the scope of the survey. The term " usable" refers to all establishments (k) in stratum j and estimation cell i which completed the survey.

Estimates of means or proportions can also be obtained from the survey using standard formulas. A mean would be computed as follows:

Y= Ei,$, & Y .3. x

• WGTj ,3, g

• NRAFj ,3, y i

E3,j,y WGT ,3, y

• NRAFj,3, y j

The proportion of establishments having some characteristic can be is computed intsrpreted using as the a 1 same formula as on or 0 depending thewhether mean where the value yb'd,c the characteris g

is present or not.

The final comment regarding estimation relates to estimation of sampling errors. It is generally accepted as good practice in sample surveys to compute sampling errors related to estimates produced. It is planned that such errors will be computed for at least the major estimates of the survey. These include the total volume _of mixed waste nationally and by major type of establishment. For this survey it is planned to select one of the following three commonly used methods to compute sampling errors:

=

balanced half sample replication;

• jackknife; or Taylor series approximation.

As a detailed discussion of these methods is beyond the scope of this design document, we provide a reference to the book by 1

Wolter.

l i

l l

l 'Volter, K. M. (1985). Introduction to Variance Estimation, Springer-Verlag, New York.

l 14

. a APPENDIX B. QUESTIONNAIRE - NATIONAL PROFILE ON MIXED WASTE

e c F.-

Recipients of the National Profile on Mixed Waste Questionnaire As described in the following hSer, Oak Ridge National Laboratory (ORNL) has been -

participating in a project to con;phe a national profile on the volumes, characteristics, and treatability of commercially generated low-level radioactive mixed waste. The project is jointly sponsored by the U.S. Nuclear Regulatory Commission (NRC) and U.S. Emironmental Protection Ay:ncy (EPA). ORNL realizes that a number of facilities have participated in recent state or compact surveys that have requested various amounts of information on mixed waste.

Although generally supportive of regional interests and requirements, these regional surveys are unsatisfactory for compiling a national survey. Therefore, ORNL is in the process of conducting a national mixed waste survey.

Your facility has been randomly selected to participate in this pretest of the national survey to be conducted this fall. This pretest is being conducted with the support of the Appalachian Compact Users of Radioisotopes (ACURI) and is critical in ensuring that meaningful results arise out of the full survey.

Please answer the questions on the attached questionnaire to the best of your ability and return the survey, if you do or do not generate and/or store mixed waste, no later than August 15,1991.

Any commenta on it questionnaire, instructions, or definitions would also be appreciated and would help us in fine-tuning the survey form for the national survey.

We are very appreciative of your support in this important national project. If you have any questions, please feel free to telephone collect:

John Mrochek (615/574 6840)

Jerry Klein (615/574-6823)

Andy Francis (615/576-8456)

Sincerely, Jerry A. Klein

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- - OMB No. 3150-0161 Expiration Date 6/3092 l

DRAFI' QUESTIONNAIRE NATIONAL PROFILE ON MIXED WASTE by Oak Ridge National Laboratory O/ /

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NOTIG-Public reporting burden for this collection of information is estimated to average 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per response, including the time for reviewing instructions, scarching existing data snurces, gathering and maintaining the data needed, and completing and reviewing the collection of information. Send comments regarding this burden estimate or any other aspect of this collection ofinformation,includmg suggestions for reducing this butden, to the Information and Records Managerrent Branch (MNBB 7714) U. S. Nuclear Regulatory Commission, Washington, DC 20555; and to the Office of Infoimation and Regulatory Affairs, Office of Management and lludget, Paperwork Reduction Project, #3150-0161, Washington, DC 2050T.

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QUESTIONNAIRE INTIRUCrlONS A. General Information )

• Facility Information - Name is the facility name as shown on the NRC/ Agreement State license or the name as shown on official facility stationary. Contact name is the person best able to answer questions concerning the questionnaire and its objectives.

• Facility Category - Please select the single, best match to your facility's category.

If the choice is between two possibilities, select the one most representative of your waste.

• NRC/ Agreement State license number - Self explanatory.

• EPA identification number - Self explanatory.

• Name and title...- This may or may not be the same as the previously requested Contact Name.

• Date survey...- Self explanatonj.

B. Iow-Level Radioactive Waste (LLRW) - Please enter the total, "as. shipped volume (in cubic feet) of LLRW shipped to a disposal site during 1990 in each of the three radioactive waste classifications Attachment I contains a list of 23 potential LLRW streams, and if none of these fit your waste stream, the last one can be employed together with your own description of the stream. Please use the selected waste stream numbers throughout the remainder of the questionnaire for those same streams. Use the single, most descriptive name for that waste stream as shown in Attachment 1 (this is the only place where it should appear in the Questionnaire). Use your best judgement in describing the Generating Practice which results in the indicated waste stream.

For the individual waste streams identified, use checkmarks to indicate their character:stics or contaminants; please enter the name, if possible, in the column under Other, if it is necessary C. Mixed Waste (MW) la. Self explanatory.

Ib. Self explanatory.

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W 2a. Use the same Waste Stream Number as employed in question 1 (listed in Attachment 1). Based upon the in rormation entered in question 1, the choice of whether or not the waste stream is a MW (MW - y/n). One of two responses is possible for Basis. A response of M is indicative of Measured and is dependent on whether a liquid waste or a leachate of a solidiGed waste (using the Toxicity Characteristic Iraching Procedure, TCLP procedures described in 40 CFR Part 261, Appendix 11) contains a hazardous material listed in Attachment 2, Table 1, at concentr,tions exceeding those in the Table by chemical analysis. A response of E, indicating " estimated' based on process knowledge is used where concentrations were not measured and judgement was based upon knowledge of the hazardous characteristics of the waste or actual listed hazardous chemicals present in the waste.

Activity requests the radioactivity of the waste stream in millicuries. Under llazardous Component, the Name and Concentration in mg/Kg (if known) is requested -- typical names are toluene, xylene, Icad, cadmium, chloro 0uorocarbons, trichloroethylene, tetrachloroethyk ~:, chromate, waste oil, organic chemical, etc. As Generated requests the total annual volume (in cubic feet) of the waste stream "As Generated no.t as shipped."

2b. Using the same Waste Streams enumerated in 2a, enter the Physical Form e.g.

aqueous, bulk liquid, adsorbed liquid, uncompacted or compacted solid, trash, etc.

Enter the Major Radionuclides and radioactive waste Classification (Class A, Class B, or Class C). Enter the EPA HW Codes as listed in Attachment 2, Tables 1 to 5, and whether the waste was Treated On- or Off-site. Indicate the Type of Treatment which the waste was subjected to; here some possible choices include Onsite Incineration undu Nternate Fuels storage for decay, compaction or super-compaction, offsite shipment of scintillation Guids containing "C or tritium to Quadrex, encapsulation and stabilization in a cement or bitumer., and wastewater treatment for aqueous. based process solutions and facility wash wastes (here neutralization, volume reduction, and contaminant removal may come into play).

2c. This question requests information on the effect of current treatment and storage practices on waste Volume . Again, use the same Waste Streams identified previously; indicate whether the MW is Stored On- or OiI-site. Indicate any Change in Volume, Activity, or IlW Concentration caused by your current treatment and storage practices and, finally, the Final Volume of MW requiring Ultimate Disposal.

This is the MW remaining after all current treatment techniques have been exhausted and is waste which, under current conditions, you cannot dispose of.

3a. Requested here is information on the Cumulative Volume of MW in storage as of 12/31/90. Waste Stream in this question may or may not match the streams listed in questions 2b and 2c depending on previous versus 1990 activities at your facility.

Also requested is the activity, and the name and concentration of the hazardous components of any MW in storage.

3b. The final question requests information on your plans regarding the reduction or elimination of your Waste Streams. This refers to ongoing practices which resulted in the generation of those streams enumerated in questions 2b and 2c.

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i Please complete the Questionnaire as accurately as possible within three weeks after receipt and return it in the enclosed envelope to:

OAK RIDGE NATIONAL LABORATORY ATIN: Dr. J. A. Klein Nuclear Waste Studies and Applications P. O. Box 2003 K 1037, MS-7358 Oak Ridge, TN 378317358 IF YOU IIAVE ANY OUESTIONS ABOUT COMPLETING tills OUFSTIONNAIRE, please call (615)S74 6823 or (615)S74-6840; M F,8:00 AM to 4:30 PM, EDT I

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QUESTIONNAIRE A. General Information

• Facility Information Name:

Address:

Contact Name:

Contact Tel No.: ( )

t e Facility Category Check ONE category which best describes your facility:

1. Nuclear Power Plant Boiling Water Reactor (BWR):

Pressurized Water Reactor (PWR):

Research & Test Reactors:

2. Medical (non-government)

, Hospital j <250 beds:

l 250 to 750 beds:

i >750 beds:

l Medical college / hospital:

Laboratory:

Research:

3. Academic

<10,000 students:

l 10,000 - 20,000 students:

l >20,000 students:

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4. Industrial Manufacturing:

<500 employees on site:

500 to 2000 employees on site:

>2000 employees on site:

Research and Development:

Decontamination facility & waste reduction:

Scaled sources / gauges / instruments:

Waste broker / processor:

Nuclear fuel cycle other than power reactors:

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5. Government Federal IIospital:

Research & Development:

Military:

State:

Other (describe):

• NRC/ Agreement State license number:

• EPA identification number:

EPA facility classification Large quantity generator (>1000 Kg/ month):

Small quantity generator (100-1000 Kg/ month):

Conditionally exempt small quantity generator (<100Kg/ month):

No EPA classification:

• Name and title of person completing form and telephone number:

Name:

Title:

Tel. No.: ( )

• Date survey form completed:

B. Iow-Level Radioactive Waste (Il.RW)

Enter total volume of LLRW (as defined in the Low-Level Radioactive Waste Policy Amendments Act of 1985) shipped for disposal in cubic feet (one 55-gal drum is equivalent to 7.5 ft 3and one 30-gal drum is equivalent to 4.0 ft)) and list the radioactivity classification (A, B, or C as described in 10 CFR 61.55) during 1990:

Class A (ft 3)

Class B (ft 3)

Class C (ft 3) 2

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List your LLRW streams by number and the name of the waste (as dermed in Attachment 1). Also list practices at your facility that generate this LLRW {c.g. laboratory counting pmcedures, waste from research or manufacturing, spent reagents, cleaning of laboratory equipment, cleaning of contaminated components, decontamination of lead shiciding, lead contaminated during process, backflush of resin filters and changeouts, equipment / tool decontamination, laundering garment waste, pump seal oil, etc.).

1.LRW WASTE STREAM NO. LLRW WASTE STREAM NAME 11RW GENERATING PRACTICE For each of the LLRW streams identified above, indicate which exhibit the following characteristics or contain the contaminants listed (continue using the same waste stream numbers as the identifier for the waste stream.

WASTE IGNIT- REAC- CORRO- TC CONTAMINANE STREAM ABLE TIVE SIVE *IDXIC SGLV. PAINT ORGANICS MFiTAIS OTHER l

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C. Mixed Wastc (MW)

Mixed waste contains radioactive material subject to the Atomic Energy Act (AEA) of 1954 and hazardous waste (HW) subject to the Resource Conservation and Recovery Act (RCRA) of 1976 and its amendments.

A dual regulatory framework exists for MW, with EPA or authorized states regulating the llW and NRC.

NRC agreement states, or DOE regulating the radioactive waste. NRC generally administers the AEA for commercial and non DOE federal facilities while DOE regulates radioactive materials at DOE facilities.

The dual regulatory framework for MW stems from the RCRA definition of solid wa:te; HW is defined as a subset of solid waste. RCRA specifically excludes from the definition of solid waste " source, special nuclear, or by-product material as denned by the AEA of 1954, as amended" [ RCRA Section 1004 (27)]. The AEA regulates source material, special nuclear material, and byproduct material. Classes of radioactive waste are transuranic waste, high- and low-level waste, and spent nuclear fuel. These wastes are subject to regulation by NRC. Any class of radioactive waste that contains a HW as listed in Subpart D,40 CFR 261.31.33 or which exhibits any of the HW characteristic.; (ignitability, corrosivity, reactivity, toxicity) as identified in Subpart C,40 CFR 261.20.24 is considered to be a MW. The radioactive component of commercial MW will generally be LLRW as defined in the Low-Level Radioactive Waste Policy Amendments Act of 1985 (LLRWPAA).

[

Reference:

' low Level Mixed Waste - A RCRA Perspective for NRC Licensees," EPN530 SW-90-057 (August 1990).]

la. Have you generated LLRW during 1990 that contains a listed hazardous chemical as defined in Attachment 2, Tables 1-5 (EPA hazardous chemical code series D, F. K, P, and U) or exhibit one of the hazard characteristics (ignitable, corrosiveness, reactivity, toxicity characteristic - see Attachment 1, Definitions - for further descriptions of these hazard characteristics), or that contains a hazardous chemical deGned by your state (RCRA authorized state)?

Check YES: or NO:

l b. As of 12/31/90, did you have any LLRW in storage as defined in D. Low-Level Radioactive Waste above?

Check YES: or NO:

If the answers to Question la and Ib are both NO,it will not be necessary for any other question of this survey to be answered. Please return the survey to the Oak Ridge National Laboratory in the enclosed envelope. If the answer to either or both of the above questions is yes, please complete the remaining survey questions.

2a. In order for a waste stream to be a mixed waste, it must contain radioactive waste as well as a hazardous waste. The hazardous components of mixed waste typically are organic solvents, metallic lead, mercury, chromate, cadmium wastes, halogenated cleaning / degreasing wastes, aqueous corrosive liquids, or waste oils. For each waste stream identified in Ib, indicate if it is a mixed waste, the basis for that judgement e.g.

M measured or E-estimated from process knowledge, and if it is, how much was generated (not shipped)in 1990 and its activity in millicuries. Also indicate the name of the hazardous component and its concentration.  !

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I TOTAL VOLUME WASTE MW BASIS ACTIVITY H AZ. COMPONENT GENERATED DURING 1990 STREAM (Y/N) (M or E) (mci) NAME (mgEg) (ft 3/yr) 2b. For each mixed waste listed in 2a. indicate the physical form (aqueous, bulk liquid, adsorbed liquid, uncompacted/ compacted solid) and the major radionuclides (e.g. 3H, "C, 32P, 5*Ni, 63Ni, "Sr, 37Cs, "Co, etc.). Also indicate the EPA hazardous waste code [some typical codes are toluene-F005, lead-D008, mercury-D009, chromate-D007, cadmium-D006, and halogenated solvent cleaning / degreasing wastes-F001, more complete listings of these EPA HW Numbers are contained in Attachment 2, Tables 1-5]. Examples of Treatment Types include: Onsite incineration under Alternate Fuels, storage for decay, compaction / super compaction, offsite shipment to Quadrex.

encapsulation, stabilization in an inorganic cement, wastewater treatment for aqueous-based process solutions and facility wash wastes (neutralization, volume reduction and contaminant removal), and other (please describe).

WASTE PHYSICAL MAJ. RADIO CIASS EPA HAZ. TRFATED TREATMENT STREAM FORM NUCLIDES (A,B,C) CODES (ON/OFF- TYPE SITE) 5

2c. For each mixed waste listed in 2b indicate the effect of your current treatment and storage practices. For the data on " Ultimate Disposal" indicate the volume of waste which is not subject to any currently known treatment, yet must be treated and/or disposed of.

WAS'l . S'IOR AGli Cil ANGli IN VOI. REO. Ulli'.

STREAt1 (ON/OFF- VOI UMI! ACI1VITY II A7.COMil DISPOSAi, SITE) (ft') (mci) (mg/Kg) (ft')

3a. If you had mixed waste in storage at the end of 1990, please answer the following:

CUMULATIVE AMT.

WASTE t\CTIVITY IIA 7 COMPONENT S'ID Ril R3 1?f.]f3]

STREAM (mci) NAME [mygg) (ft)

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3h. Do you have any plans to reduce and/or climinate any of your indicated mixed waste streams?

Check YES: NO:

If your answer is YES, please indicate your planned method for reducing and/or eliminating the mixed waste streams, e.g., substitution, reduction of the indicated l practice, restrictions on the entry of hazardous materials into radioactive aicas, other (please describe):

WAS'ITi STREAM REDUCMON/EI,IMINATION METIIOD l

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A1TACllMENT 1 Indierte your waste streams in B. and the Tables of Question 2 by entering its code number from the following list.

Waste Stream Noa Waste Stream Name 201 Animal Carcasses / Biological Waste (Non-infectious) 202 Trash and or Solid Waste (not lead) - non-compacted 203 Trash and or Solid Waste (not lead) compacted 204 Filter Media - Dewatered 205 Filter Media - Solidified 2tM Gaseous Sources 207 Incinerator Ash or Residuals 20S lon Exchange Resins - Dewatered 209 lon Exchange Resins - Solidified 210 Irradiated Reactor or Pool Components 211 Liquids Aqueous - Absorbed 212 Liquids Aqueous - Solidified 213 Liquids Organic - (Solvents, Chlorinated Solvents, etc.)

214 Liquids Scit.tillation, containing "C and/or tritium - (fluids or vials) 215 Liquids Scintillation, containing radioisotopes other than "C and tritium - (fluids or vials) 216 Mineral Extraction Waste 217 Uranium Studges 218 Radioactive Scaled Sources, Devices, or Gauges 219 Solidified Evaporator Bottoms / Concentrates / Sump Sludge 220 Vitrified Ash or Resins 221 Waste Oils (Seal Oils from pumps for example) 222 Lead 223 Mercury 224 Other - (Specify) l l

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A1TACllMENT 2 Definitions:

WASTE - For purposes of this study, waste is defined as a material not able to be recycled which must be treated, stored, or shipped offsite for disposal! storage. This deGnition is meant to include waste oils or other materials which may be designated as " alternate fuels" and subsequently burned onsite or offsite.

LOW-LEVEL-RADIOACTIVE WASTE - Low-leve' radioactive waste (LLRW) is radioactive material that (a) is not high-level radioactive waste, spent nuclear fuel, or byproduct material as defined in section lle.

(2) of the Atomic Energy Act (i.e. uranium or thorium mill tailings) and (b) the NRC classifies as LLRW consistent with existing law and in accordance with (a).

SOLID WASTE - The Resource Conservation and Recovery Act (RCRA) deGnes solid waste as "any gartage, refuse, sludge from a waste treatment plant, water supply treatment plant, or air pollution control facility and other discarded material, including solid, liquid, semisolid, or contained gaseous material resulting from industrial, commercial, mining, and agricultural operations, and from community activities," but does p_ot include

" source, special nuclear, or byproduct material as deGned by the Atomic Energy Act of 1954.. " [ RCRA Section 1004(27)]. EPA. NRC, and DOE interpret the exception for source, special nuclear, or byproduct material as referring only to the radionuclide component, and not to the entire waste mixture. [ low-Irvel Mixed Waste a RCRA Perspective for NRC Licensees, EPA /530-SW-90-057}.

IIAZARDOUS WASTE - A hazardous waste is defined in RCRA as " ..a solid waste, or combination of solid wastes, which because of its quantity, concentration, or physical, chemical, or infectious characteristics may..." pose a " substantial present or potential hazard to human health or the environment when improperly... managed." [ RCRA Section 1004(5)). A solid waste is a hazardous waste if it is a listed waste (see Tables 2-5), an ignitable, corrosive, reactive, toxicity characteristic (see Table 1), acute hazardous (see Tab!c 4), or a toxic (see Table 5) waste.

MIXED WASTE - For purposes of this project, mixed waste (MW) is defined as " waste that satisfies the definition of LLRW in the LLRW Policy Amendments Act of 1985 (LLRWPA.A) and contains hazardous waste that (1) is listed as hazardous waste in Subpart D of 40 CFR Part 261 or (2) causes the LLRW to exhibit any of the hazardous wasste characteristics identified in Subpart C of 40 CFR Part 261". In addition, the following are included in the definition of hazardous materials for the purpose of this study:

Oils and sludges, and other materials classified as hazardous by a RCRA-authorized state.

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IGNIT-ABILITY - A solid waste exhibits the characteristic ofignitability if a representative sample of the waste has any of the following properties: A liquid, other than an aqueous solution containing les than 24% alcohol by volume, that has a flash point less than 140' F.

or it is not a liquid and is capabic, under standard temperature and pressure, of causing fire through friction, absorption of moisture or spontaneous chemical changes and, when ignited, burns so vigorously and persistently that it creates a hazard.

or It is an ignitable compressed gas as defined in 49 CFR 173.300 and as determined by the test methods described in that regulation or equivalent test methods approved by the administrator under 40 CFR 260.20 21.

or It is an oxidizer as denned in 49 CFR 173.151.

CORROS-IVITY - A solid waste exhibits the characteristic of corrosivity if a representative sample of the waste has either of the following properties: It is aqueous and has a pli less than or equal to 2 or greater than or equal to 12.5 as determined by a pil meter using either an EPA test method or an equivalent test method approved by the administratoc under the procedures set forth in 49 CFR 260.20.21. The EPA method for pilis specified as Method 5.2 in " Test Methods for the Evaluation of Solid Waste, Physical! Chemical Methods".

of It is a liquid and corrodes steel (SAE 1020) at a rate greater than 6.35 mm (0.250 in.)

per year at a test temperature of 130' F as determined by the test method specified in NACE (National Association of Corrosion Engineers) Standard TM-01-69 as standardized in " Test Methods for the Evaluation of Solid Waste, Physical / Chemical Methods" or an equivalent test method approved by the Administrator under the procedure set forth in 49 CFR 260 20.21.

REAC1'-

IVITY - A solid waste exhibits the characteristic of reactivity if a representative sample of the waste has any of the following properties:

It is normally unstable and readily undergoes violent change without detonating.

or It reacts violently with water.

or It forms potentially explosive mixtures with water, or When mixed with water, it generates toxic gases, vapors, or fumes in a quantity sufficient to present a danger to human health or the environment.

or It is a cyanide or sulfate bearing waste which when exposed to pl1 conditions between 2 and 12.5, can generate toxic gases, vapors, or fumes in a quantity sufGcient to present a danger to human health or the envirocment.

of It is capable of detonation or explosive reaction if it is subjected to a strong initiating source or if it is heated under confinement.

or It is readily capable of detonation or explosive decomposition or reaction at standard temperature and pressure.

or 2- %

i It is a forbidden explosive as defined in 49 CFR 173.51 or a Class A explosive as delined in 49 CFR 173.53 or a Class B explosive as defined in 49 CFR 173.88.

TOXICITY CIIARACI'liR-ISTIC - A solid waste exhibits the charat teristic of toxicity if, using the test methods described in Appendix 11 of 40 CFR Part 261 or equivalent methods approved by the administrator under the procedures set forth in 40 CFR 260.20.21, the extract from a representative sample of the waste contains any of the contaminants listed in Table 1 at the concentration equal to er greater than the respective value given in that Table. Where the waste contains less than 0.5cc filterable solids, the waste itself, after filtering, is considered to be the extract for the purposes of this section.

A solid waste that exhibits the characteristic of toxicity has the EPA IlW Number specified in Table 1 which correspc.nds to the toxic contaminant causing it to be hazardous.

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[THESE TABLES WILL BE TYPED FOR PRETEST QUESTIONUAiRE]

TABt.E 1.-M AXIMUM CONCENTRATION OF CON.

T AMIN ANT S FOR THE TOXICITY CHAR ACTERIS-TIC

, , Fegia.

! " A W Cy.:am 4 .: ' # ~Y No' . CAS N: - -i tevet irn; L)

DMt t., sen.: 7440-35-2 50 D)?5 B rum .. 7440-??-3 *00 0 DDiS Sen:ene 'i-43-2 C5 DM5 . Ca:mism 7440-43-9 i0 D319 Caro , . wathar ce f5-2~-5 C5 D020 ' Ch r:ane 57 't-9 C 03 D 21  : CNroden:ene. 105-9?-7 ' C0 0 D 22 'CNrVrm 67-65 3 60 DX7 'C5omvm. 744C 47-3 5 O.

D;23 e.C'e! N . 95 '.F 7

• 2X 0 0:24 ' r ..C'e so! . 1 DE- 3 9-4

• 2 ?: ,

D025  : C'e s>.. 105-44-5 ' 2 bb 0 D026 C'e55

• 200 3 D:;6 2.4 D. 94-7 -7 10 0 D 27 1.4 Dc.'r:De .:e. e.. *05-46-7

. 75 D:28 1.2 Delreel*4ne . '07 ~>C-2 C5 D:29 1.1 D.:n o oe:nylene. 75-35-4 07 0030 2.4 D.n:r:t:Isene.. 12' *4-2 C 13 D:12 Entin 72-2:~5 C2 D 31 Ha; a:n.:. ian: ca 75-44-b 005 ep .:e)

D *2 wera:myrde :eme. itE4* ' .' ' 3 D033 He x a:Ny edt.a c.e ne.. 57-65-3 C5 D;34 . Heia:Nyee:nane.. 67721 30 D008 Lea:L 7439-92-1 50 DC'3 , L e. Ane. 52-25-9 C4 DM9 , Mercu y .. 7.39-97-6 02 D014 Mein:ry:Nor . 72-43-5 50 0 DL35 j Methyi etny'

• eione.. 7P 43-3 200 0 D336  : N.tropen;ene.. 96-95-3 20 DC37 Petta:Nor opher.o! . 87-86-5 100 0 CC3B Pr*5ce , 110-66-1 '50 D310 . Selenv". .

0 0011 Saver. . . '.7440-22-4 l

7722-49-2 50 DC39 , Tette:Nar pelhetene .. 127- E-4 07 D015 :T aphe,e.. , 6X1-25-2 0A D340 , Tr.:nkotoeinyle^c . 79-01-6 05 D041 2.4.5. Tre:hlorophenot ..! 55-95-4 400 0 D342 2.4.6.TecNor no~ enut.. 59-06 2 ' 20 D017 ' 2.4 5.TP (Saves!.. ..i 93 72-1 10 D043 l V.nyl CNCVde.. ..' 75-01-4 02 m gm.eum,e i Hazarsovs waste namner.

Chem. cal abstracts servce nurnbe'

• Quant, tat.on hv! rs Q'ealer (* an ;"4 Cal:Utaled reggiatory level The quantaaton Imit tnerelo'e be:omes the regu'atyy le ici.

• it o . m and p-Ciesol con.:entratons cannot be citteren-i.ated. tne total : resol (D026) concent'ation is used Tv regutaiory level of totat cresol is 200 rng/I 2-2

e The followir g telid w u'es are lhted hmirdous u:ules from notarecific sources un! cts they are peludest imder il 060 00 atid 260 22 amt Ihted in Atmen-Table 2 _ -- . . .

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t., ..e ,.,,w t . e s.e t

e. .tm.4 .o., e ~ t-

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f.9 *e*C .

g g, . g **g 1 dis eq n;4 tt hetsgesWeJ troena's .. Lid en on f tenN le b ec ed@ 8.e thvene, , t ?l i

tw mspoen . eae me sheir*e (m.se ie 1 1 1 s+ec eust.e P.e e t e tes t u, ~ , .e ,. . , -. .,0 ,cw s. t.p cie, *

, ,e ,. e . t ;

.~,.~.,e.e,t ye id ., a m .e,..e....-en,.~,. .e, e d te oeh*t eut a es.we, a: n e'.,, Me o,'c.e t., i<,.. heyenes et t e.. e.e ... ..,'airo e t.. to'nne.

. east t e e.u

. .e-t .Q u.e..,

e f ;*i 17 f (ns f sp..y m .vn JN;

• t e 37.5.g g;,8W 8 4ttsgeneig d t'48 e nog I,4e.Mw@g t'+g\ Ae #ee p.,as"4 4 * =p e ha n

. v g i .. s. s4 < a. . . .eae e < *, u .n, ~ t i ; eew.e t i; .mx, weeg emo sw.mi.se s ee.

• s **m e's.am awaenee', e.ma * *

• i.e. e ee.e , .

e* u-o t **** m .v. t '* e'e's r *'v='s t e c e i.te a ev ie' v w ee.5 .. w : a m:. 3, .t-m ei s.% w m-.s o* ev e < .e has.yneiec sp.,ws w v..s, -

4 ste $ eA f ;% I 8X4 e 8 WS s'eJ ts# lets#an hem Fe '.t ees'y O' t'd.e g;,eeme st*'e e **% 4 M t.49 8t'49N eN ***Ve t 9 sy3 * *e ne 3. =; et maa a use a.iv.s tw e~t ' r.e *, 4( e t.v== eP.careee,eie iv

.c ce e e ev mettei w ,v tesone

• hv s< o o rittJ%. .aspw e%J i

• e ** a J Sf %3t* %F' Vow e'hs ty e t Me e Nt (. ort (*m%g 'g 4 f 6te. DUq t*g 4W,e gM e' %3* 9,qhge* e e4 b.betWt 4'* l se %)eT b.pbee4 Dul re l b e'%'t (WUadt*ng '

te+pe ,se tee of **ere 98 P e d'4=ee 'ip* heWeRo.ed 60'er%*t 894 4 BJ e' e,f 8 te

• 14" ', %' W f* > , IDt *196. **e l 98 LI"e nW 8* e 08 8%ete b#*,9'% tt'ef #5$ 4C 4 8 Y. ? I */ 4 4 14I LC% e*J ties gg,.gy.e g p gene Peg set gee *y p' g'g ge tpgal 69 een'% {

,x, se

.* 3:

..u.., u e ,I m.t.s...

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m ...p-.:r m.:,,~,y,w m Table 3

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2-4

Table 3 (continued) 7 ig!

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y , Table 4 (continued) 1 3 4 9 h

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4 .. Table 4 (cont inued) E

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