ML20024G613

Proposed Tech Specs,Implementing Mods Described in Permanent Plant Changes to Accommodate Equilibrium Core Scram Reactivity Insertion Characteristics.

ML20024G613
Person / Time
Site:	Monticello
Issue date:	03/01/1974
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G609	List: ML20024G608 ML20024G612 ML20024G613
References
NUDOCS 9102130486
Download: ML20024G613 (28)

Text

~ .- .

Bases Continued: .

The operator will set the low low water ECCS initiation trip setting 6'6" 6'10" above 2.3 the top of the active fuel. I!cwever, the actual setpoint can be as much as 3 inches lower than the 6'6" setpoint and 3 inches greater than the 6'10" setpoint due to the deviatione discussed on page 18.

E. Turbine Control Valve Fast Closure Scram - The turbinesud control neurronvalve flux fast closurefrom resulting scram fastand PRT j

l is provided to anticipate the rapid increase in pressui2 failure of the bypass.

' closure of the turbine control valves due to a load rejection and subsequent This transient is less severe than the turbine stop valve closure with bypass failure and therefore adequate margin exists. Reference sections 14.5.1.1 and 14.5.1.2 FSAR and supplemental information entitled " Permanent Plant Changes to Accommodate Equilibrium Core Scram Reactivity Insertion Characteristics", dated January 23, 1974.

F. Turbine Stop Valve Scram - The turbine stop valve scran with PRT, like the load rejection scram with PRT, anticipates the pressure, neutron flux and heat flux increase caused by the rapid closure of the turbine stop valves and failure of the bypass. With a setting at 102 of valve closure for scram, and PRT, the increase in heat flux is limited such that adequate pressure and thermal marginc are maintained. The PRT opens safety / relief valves to limit the pressure and heat flux increases, and allows safety / relief valve reclosure as pressure decreases. For this event, the peak surface heat flux and MCilFR remain within limits. Reference FSAR Sectic, 14.5.1.2.2 and suppicmental information submitted January 23, 19','4 C. Main Steam Line Isolation valve Closure Scram - The main steam line isolation valve closure scram antic

• pates the pressure and flux transients which occur during normal or inadvertent icolation valve closure. With the scram set at 10% valve closure there is no increase in neutron flux.

H. Reactor Coolant Low Pressure Initiates Main Steam Isolatio, Valve Closure - The low pressure isolation of the main steam lines at 350 psig was provided to give protection against rapid reactor depressurization asd the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature which occurs when the main steam line isolation valves cre closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safetr limit. Operation of the reactor at pressures lower than 850 psig requires that the reactor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is ;,rovided by the IRM high neutron flux scram. Thus, the combinatien of main steam line low pressure isolation and isolation valve closure scram assures the availability of the neutron scram protection over the entire range of applicability of the fuel cladding integrity safety limit.

The operator will set this pressure trip at greater than or equal to 850 psig. However, the actual trip setting can be as much as 10 psi lower due to the deviations discussed on page 18.

22 7.3 BASES gDR102J30486 740301 p ADOCK 05000263 DR

4 2.0 SAFETY I,IMITS LIMITING SAFETY SYSTEM SETTINGS 2.2 REACIOR CDOLANT SYS' EM 2.4 REAC"t0R CDOIANT SYSTEM Applicability: Applicability:

Applies to limits on reactor coolant system Applies to trip settings of the instruments pressure.

and devices which are provided to prevent the reactor system safety limits from being ex-4 ceeded.

Objective: Objective: .

To define the level of the process variables To establish a limit below which the integrity of the reactor coolant system is not threatened at which automatic protective action is .

due to an overpressure condition. initiated to prevent the safety limits from being exceeded.

Specification:

Specification:

'the reactor vessel pressure shall not exceed A. Reactor Coolant High Pressure Scram shall 1335 psig at any time when irradiated fuel is be 6 1075 psig.

present in the reactor vessel Reactor Coolant System Safety /Relier Valves a

- B.

shall be set as follows:

1 6 valves at i 1080 psig.

4 23 2.2/2.h REv

1

~

l l

Bases Continued:

We 2.2 The normal operating pressure of the reactor coolant system is approximately 1025 psig.

nevere primary system pres- l turbine trip with failure of the bypass system represents the mostThe peak pressure in this sure increase resulting from an abnormal operational transient.

transient is limited to 1178 psig at the bottom of the vessel. The primary system overpressure (high flux) scram. Peak protection analysis assumes the closure of all ?!SIV's with indirect pressure at the vessel bottom is 1285 psig. .

25 2.2 BASES REV

Bases:

reactor coolant I 2.4 The settings on the reactor high pressure scraa, prompt relief trip system,and turbine stop valve

( system safety / relief valves, turbine control valve fast closure scram, closure scram have been established to assure never reaching the reactor coolant ystem pressure safety limit as well as assuring the system pressure does not exceed the range of the fuel cladding integrity safety limit.

The APRP neutron flux scram and the turbine bypass system also provide In addition to preventing power operation above 1075 psig, the l protection for these safety limits.

pressure scram backs up the APRM neutron flux scram for steam line isolation type transients.

The reactor coolant system safety / relief vaIves offer yet another protective feature for the reactor l In compliance with Section III of the ASME Boiler and Pressure coolant Code, Vessel system1965 pressure safety limit.the safety / relief valves nust be set to open at a pressure no higher than edition, i limit the reactor pressure to no more than 110 percent

' 105 percent of design pressure, and they mustThe safety / relief valves are sized according to the code for a conditi of design pressure.

I closure while operating at 1670 NWt. followed by no PSIV closure scram but scram from an indirect (high flux) means. With the safety / relief valves set as1285 specified psig. herein,'the maximum See FSAR Section 4.4.3 vessel and pressure

' (at the bottom of the presscre vessel) would be1974.

about Evaluations presented indicate that a total 23, supplemental information submitttd January of six dual purpose safety / relief valves set at 1080 psig maintain the peak pressure during the transient within the limits v11 owed by the ASME Code.

l The operator will set the retttor coolant high pressure scram trip setting at 1075 psig or lower. l However, the actual setpoint can be as nuch as 10 psi above the 1075 psig indicated In a like set manner, point the due to the deviations discussed in the basis of Specification 2.3 on Page 18.

operator will set the reactor coolant system safety / relief valve initiation trip settirg at 1080 psig or lower. However, the actual set point can be as much as 11 psi above the 1080 psig indicated Analyses were set point due to the deviations discussed in the basis of Specification 2.3 on Page 18. l l performed assuming a safety / relief valve setpoint of 1080 psig + 17.. l A violation of this specification is asrumed to occur only when a device is knowingly set outside of the limiting trip setting, or when a sufficient number of devi es have been affected by any means 26 2.4 BASES REV

1 Bases Continued:

3.1 cor. denser vacuum initiates a clo nre of the turbine stop valves and turbine bypass valves which t- the condenser.

Closure of the turbine stop and bypass valves causes eliminates the beat f aut a pressure transient, neutron flux rise, and an increase in surface heat flux. The turbine stop valve closure scram with PRT adequately preserves the margins to pressure and MCHER limits should Reference FSAR Section 14.5.1.2.2 and supplemental a *urbine trip vith bypass failure occur.The condenser low vacuum scram is a back-up to stop valve ,

in ormation subr itted January 23, 1974.

c1csure scram and causes a scram befc_e the stop valves are closed and tl.us the resulting transient is less severe.

Scram occurs at 23" Hg vacuum, stop valve closure occurs at 20" Hg vacuum, and bypass closure at 7" Hg vacuum.

l High radiation levels in the main steamline tunnel Aabove neramthat due towhenevet is initiated the normalsuchnitrogen and I

oxygen radioactivity e an tenindication of leaking times normal fuel.

full power background. The purpose of this scramf is to radiation Icvel excec reduce the source of s.scn radiation to the extent necessary to prevent e'xcessive re I

radioactive materials. to the main condenrer prevented by the air. ejector off-gas monitors which cause an isolation d d for a of f-gas line provided the instantaneous limit specified in Specification 3.8 is excee e -

l i

15-minute period.

The main steamline isolation valve closure scram is set to scram when the isolation valves 410% closed from full open. By scramming at this setting the resultant transient is insignificant.

cccur when the valves cloce.

Reference Section 14.5 .1.3.1 FSAR and supplemental information submittad February 13, 1973.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status. Reference Section 7.7.1 FSAR. .

Ths manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system provides protection against excessive power levels and short reactor periods in the 39 3.1 BASES REV

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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEIUANCE REQUIRE ~riENTS E. Reactor Building Ventiletion Isolation and Standby Gas Treatment System Initiation

1. - a . Except as specified in 3.2.E.1.b below, four radiation m nitors shall be operable at all times.

b. One of the two monitors in the venti-lation plenum and one of the two radia-tion monitors on the refueling floor may be inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the inoperable monitors are not restored to service in this time, the reactor build- -

ing ventilation syste-? Aall be iso-lated and the stae ** 7. m treatment system operat: untit opairs are complete.

2. The radiation monitors shall be set to trip as follows:

(a) ventilation plenum i 3 tr/hr (b) refueling floor 6100 mr/hr

3. When irradiated fuel is in the reactor vessel and the reacter water temperature is above 212"F, the limiting conditions for operation for the instrumentation listed in Table 3.2.4 shall be met.

F. Prompt Relief Trip System The limiting conditions of Operation for the instrumentation that initiates prompt. relief trip are given in Table 3.2.4.A.

3.2./4.2 49 REV

TABLE 3.2.4A Instrumentation that Initiates Prompt Relief Trip -

Min. No. of Operable Total No. of Instru- or Operating Instrument ment Channe.ls Per Trip Channels Per Trip Required Function Trip Settings System System (1,2,3) Conditions *

1. Turbine Stop Valve Note ' 2 2 A ,

Closure

2. Turbine Control Note 4 2 2 A Valve Fast Closure

3. Reactor Low Pressure 7900 4 950 psig 2 2 .A

- PRT Disable _

4. PRT Timer 3-8 sec 2 2 A

5. Turbine First Stage C07. 1 1 A (Note 5)

Pressure (7. of Rated) 6857. 1 1 B (Note 5)

Notes

1. There shall be two operable or operating trip systems for each function.

2. One' instrument channel may be bypassed to pernit testing of-the channel.

1

3. Upon discovery that minimum requirements for the number of operable or operating trip systeres or instrument channels are not satisfied action shall be initiated to:

(a) Satisfy the requirements by placing appropriate channels or systems in the tripped conditicn, or (b) place the plant under the specified required conditions using normal operating procedures.

60A FIV 3.2/4.2 -- . _ - - _ . _ _ _ __ _ _ -_ ._ _ _ .- _

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

e a

1 TABLE 3.2.4A (Continued)

Notes (Continuedl

4. PRT initiating signals operate directly from scram initiating instruments. Trip settings are previously established in Table 3.1.1.

5. The PRT system consists of two, two-channel three-valva subsystems. System functional requirements include one three-valve subsystem operable for power levels greater than seventy percent and the other three-valve subsystem operable for power levels greater than eighty-five percent. Should a component of oac of the subsystems become inoperable, reactor power will be reduced to the corre-

, sponding power level within 24 houra until repairs are implemented. All installed aafety/ relief valves are available for PRT service, itowever, only six are actively engaged at any one time. Should one of tLe six becone inoper.'ble. a spare valve can be activated in its place without shutting down.

• Required conditions when minimum condi.tions for operation are udt satisfied.

A. Reactor power less than 707. of rated B. Reactor power less than 857. of rated .

4

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1 i

4 4

4 1

4 3.2/4.2 60B 6-REV

Table 4.2.1 - Continued _

Minimum Test 2nd Calibration Frequency For Core Cooling.

Rod ", lock and Isoletion Instrumentation Calibration (3) Sensor Check (3)

Test (3)

Insttument Channel Ibne Fate 1 Once/3 nontns

3. Steam Line Iow Pressure Note 6 Once/ shift Stom Line liigh Radiation once/ week (5) 4 4

HPCI IS0!ATION Once/3 months None Steam Line High Flev Note 1 None

1. Note 1 Once/3 months

2. Steam Line High Temperature RCIC IS01ATION Once/3 months None Steam Line liigh Flow Note 1 None

1. Note 1 Once 3/ months

2. Steam Line liigh Temperature REACTOR BUILDING VENTIiATION Once/3 months :hice/ shift Note 1 (4) l 1. Radiation Monitors (Plenum) Note 1 Once/3 months

2. Radiation Monitors (Refueling Floor) l 1

OFF-CAS ISOIATION Notes (1,5) Note 6 Once/ shift 1., Radiation Monitors (Air Ejectors)

PROMPT RELIEF TRIP (PRT) SYSTEM Once/3 months None Note 1

1. PRT Disable (Reactor low Pressure) Note 1 Once/3 months None

2. PP.T Power Range Permissive (Turbine First Stage Pressure) Once/3 months None Note 1

3. PRT Timer Note 7 None None

4. Turbine Stop Valve Closure Note 7 None Note 7

5. Turbine Control Valve Closure NOTES:

5 (1) Initially once per month until exposure hours (M as defined on Figures 4.1.1) is 2.0 x 10 , thereafter according to Figure 4.1.1. with in interval not greater than three months. 62 Rei .

3 2/h 2

Tnble t.2.1 Continued NOTES:

ence per week (2) Calibrate prior to nomal shutdown and start-up an'i thereafter check once per shift and tee i until no longer required.

(3) Functional tests, calibrations If and sensor checks are not required when the systems are not required to be tests are missed, they shall be perfomed prior to returning the systems operable or are tripped.

to an operable status.

l (4) Whenever fuel handling is in prccess, a sensor check ahall be perfomed once per shift.

(5) A Functional test of this instrument means the injection of a simulated signs 1 into the instrument (not primpry sensor) to verify the proper instrument channel response alarm and/or initiatire action.

(6) 'ihis instrument will be calibrated every three months by means of a built in current source, and each refueling outage with e known radioactive cource.

(7)

Instrument functional test and calibration shall include verification of instrument channel response in the PRT system. Frequencies are established in TS Tables 4.1.1 and 4.1.2.

63 ggg 32/4.2 .

Bases:

32 In addition to reactor protection instrumentation which initiates a reactor ceram, protective instrumentation hns been provided which initiates action to mitigste the consequences of accidents which ere beyond the operators ability to control, or teminate a single operator error before it results in serious consequences. his set of Specifications provides the limiting conditions of operation for the primary system inolation function, initiation of the emergency core cooling system, and standby gas treatmeni systems. The objectives of the Specifications are (i) to assure the effectiveness of the protactive instrumentation when lequired by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, testing, or calibration, and (ii) to prescrite the trip settings required to assure adequate performance. his set of Specifications also provides the limiting cenditiens of operation for the control rod block syutem and the Prompt Relief System.

l Isolation valves are installed in those lines that penetrate the priusry containment and must be isolated during a loss of coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initieted by protective instrumantation shown in Table 3 2.1 which senses the conditions for which isolation *= ret 11 red. Such inscrumentation must be available whenever primary containnent integrity is required. We objective is to isolate i the primary containment so that the guidelines of 10 CFR 100 are not exceeded during on accident.

The instrumentation which initistes primary system isolation is connected. in a duel bus arrangement. l

%us, the discussion given in the bases for Specification 31 is applicable here.

%e low reactor water level instrumentation is set to trip vben reactor water level is 10'6" (7" on the instrument at 100% rated thermal pover) above the top of the active fuel. - his trip initiates closure of Group 2, and 3 primary containment isolation valves. Feference Section T.7.2.2 FSAR.

l For a trip setting of 10'6" above the top of the active itel, the valvu vill te closed before perforation of the clad occurs even for the maximu.a break in that lina and therefore the setting is adequate. I

%e low lcw re7ctor voter level instrumentation is set to trip when reactor vnter level is 6'6" above the top of the active fuel. his trip initiates closure of the Group 1 Primery containment isolation valves, Reference Section 7.7.2.2 FSAR, and also activates the ECC systems and starts the emergency diesel generator.

6h REV O 9/8/70 3 2 BASE

f Bases Continue 21 3.2 For effective emergency core cooling for the small pipe break the HPCI or Automatic Pressure Relief system must function since for these breaks, reactor pressure does not decrease rapidly enough to allow either cure spray or LPCI to operate in time. The arrangement of the trippingThe contacts is such trip settings given as to provide this function when necessary and minimize spurious operation. Reference sectfen 6.2.4 and is met.

in the specification are adequate to assure the above criteria 6.2.6 FSAR. The specification preserves the < f fectiveness of the system during perf ris of main-i.e., only tenance, testing, or calibration, and also minimizes the risk of inadvertent operaticn; one instrument channel out of service.

is reached, cause an isol tion Two air ejector of f gas monitors are provided and when their trip point rip l

l of the air ejector of f-gas line. Isolation is initiated when both instrunents reach their hir 4 There 17 4 ..nute point or one has an upscale trip and the other a downscale trip or two downscale.

delay before recombiner train inlet valve closure when the recombiners are in use and a 15 n > ue delay before off-gas isolation valve closure when the recombiners are bypassed in which the Both instruments are required for trip. The er i settings reactor operator may take corrective action.

of the instruments are set so that the maximum stack releas.e rate limit allowed _by Specificat_en 3.6.A.1 is l

I not exceeded.

Four radiation monitors are provided which initiate isolation of the reactor building and operation of the standby gas treatment system.

The monitors are located in the reactor building venti'ation .

plenum and on the refueling floor. Any one upscale trip will cause the desired action. Trip settings of 3 mR/hr for the monitors in the ventilation duct are based upon initiating normal ventilation iso-lation and Standby Gas Treatment System operation so as not to exceed the maximum release rate limit allowed by Specification 3,8. A.1 for the reactor building vent. Trip settings of 100 mR/hr for the monitors on the refueling floor are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling accident leaves the reactor building via the normal ventilation stack but that all the activity is processed ,

by the standby gas treatment system. .

The prompt relief trip (PRT) system initiates the opening of three or six safety / relief valves at reactor power levels dL70% and 21851 respectively, with the occurrence of a turbine stop valve o.

control valve closure. The PRT initiating action originates in two independent channels, each capable of satisfying system requirements through redundant instruments. Intervn1 timers with a low pressure back-up disable the PRT; self actuated pressure operation of the safety / relief valves remains unaffected.

68 3.2 BASES REV

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

Ba_ses Continued: _

3.2 The settings of the instruments provided in Table 3.2.4A ensure that pressure and thermal margins are maintained during the worst-case single-tailure-caused abnonnat operational transient, i.e., turbine trip with failure of the bypass valves. In addition, the PRT utilizes a 1 out of 2 logic system. In accordance i ^uith IEEE-279 an exception is taken to the minimum operating requirements to allow a short period of time, during which an instrument channel may be bypassed to allow for testing. For this logic the single failure protection is temporarily defeated.

Although the operator will set the sec points within the trip settings specified in Tables 3.2.1, . s, 3.2.2, 3.2.3, and 3.2.4, .the actual values of the various set points can dif fer cppreciably from the value the operator is attempting to set. 'Ihe deviations could be caused by inherent instrument error, operator setting error, dr. ft of the set point, etc. 'Ihe re fore, these deviations have been accounted for in the various transient analyses and the actual trip settings may vary *.)y cne following amounts.

l I

3.2 BASES (g FIV

Table 3.2.5 - Continued Trip Function and Deviations Trip Function Deviation ,

Instrumentation That Initiates Emergency Low-Low Reactor Witer Level

-3 Inches

. Core Cooling Systems Table 3.2.2 Reactor Low Pressure (Pump -10 psi Start) Permissive

High Drywell Pressure +1 psi Low Reactor Pressure (Valve -10 psi Permissive w l

Instrerentation That Initiates IRM Downscale -2/125 of Scale Rod Block IRM Upscale +2/125 of Scale Table 3.2.3 APRM Downscale -2/125 c f Scale APRT! Upscale See Basis 2.3 - Page 24 RBM Downscale -2/125 of Fcale RBM Upscale Same as APRM Upscale Instrumentation That Controls the Prompt PRT Disable (Reactor Low 17.

Relief Trip (PRT) Systen Pressure)

PRT Timer 1 0.5 sec.

PRT Power Range Permissive i 37. of rated pressure (Turbine First Stage Pressure)

A violation of this specification is assumed to occur only ,when a device is knowingly set outside of j the limiting trip settings, or, when a sufficient number of devices have been affected by any means i such that the automatic function is incapable of operating within the allowable deviation while in a reactor mode in which the specified fu'nction must be operable or when actions specified are not initiated as specified.

3.2 BASES 70 REV

t t

1 .

- 4 i

! 30 LIMITING CONDITIONS FOR OPERATIO!! h.0 SURVC,IM CE PEQUIRD4EIITS i

i i

< C. Scram Insertion Times C. Scram Insertion Times ,

4 1. The average scram insertion time, During each operation cycle, each i based on the de-energization of the operable control rod shall be sub-i scram pilot valve solenoids as time jected to scram time tests from the zero, of all. operable control rods fully withdrawn position. If testing j in the reactor- power operation con- is not accomplished during reactor 3

dition shall be no greater than: power operation, the measured scram  ;

insertion times shall be extrapolated  ;

i $ Inserted From Avg. Scram Insertion to the reactor power operation condi-l Fully Withdrawn Times (sec) tion utilizing previously determined 5 0.375 correlations.

I' 20 0.900

' 50 2.00 l 90 3.50 l

~

2. The avera6e of the scram insertion

times for the three fastest control

c rode of all groups of four control .

rods in a two by two array shall be no greater than

l Percent of j Rod Length Inserted . Seconds

5 0.398 l 20 0.954 50 2.120 l

! 90 3.80 l

1 i

79 j )*)[k.) .

REV 3'

4 I

-w. _ _ *

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

Bases Continued 3.3 and 4.3:

consequences of reactivity accidents are functions of the initial neutron flux. The require-ment of at least 3 couats per second assures that any transient, should it occur, begins at or above the initial value of 10% of rated power used in the analyses of transients f rom cold conditions. One opera,ble SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's are provided as an added conservatism.

5. The consequences of a rod block monitor failure have been <-valuated and reported in the Dresden II SAR Amendments 17 and 19. These evaluations, equally applicable to Monticello, show that during reactor operrtion with certain limiting control rod patterns, the withdrawal of a designaced single control rod could result in one or more fuel rods with MCHFR's ess than 1.0.

During use of such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods to assure its operability will assure that i:rprcper withdrawal does not occur. It is the responsibility of the Engineer, Nuclear, to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable rods in other than limiting patterns.

C. Scram Insertion Times

'ae

control rod system is desigq?d to bring the reactor suberitical at a rate fast enough to ensure the maintenance of adequate fuel thermal and reactor pressure margins ror non-accident events. This requires the negative reactivity insertion in any local region of the core and in the overall core to be at least as great as the (End-of-Cycle equilibrium core) scram reactivity insertion curve used in the analyses submitted on January 23, 1974 The required average scram times for three control rods in all two by two arrays and the required average scram times for all control rods are based on inserting this amount of negative reactivity locally and in the overall core, respectively. Under

these conditions, the thermal limits are never reached during_thq..transienta.xequiring control rod j' scram as presented in the FSAR and the supplemental information ~ submitted January 23, 1974. The limitiag operational transient is that resulting from a turbine stop valve closure with failure of the turbine bypass system. Analysis of this transient shows that the 7egative reactivity rates resulting from the scram with the average response of all the drives as given in the abovt Specification, provide I the required protection, and MCHFR remains greater than 1.35. in the analytical treatment of the transients, 290 milliseconds are allowed between a neutron sensor reachint _he scram point and the start of motion of the control rods.

. 3.3/4.3 BASES 85 REV .--____- ______-_

Bases Continued:

margin, the BCIC system (a non-safeguard systen) has been required to be operable during this time, (hCO 6pm).

since the BCIC system is capable of supplying significant water mkeup to the reactor E. Automatic Pressure Belief M cc

'llaf valves of the automatic precrure relief subsyst(,n are a backup to the IIPCI subsystem.

pipe break in Th ?y enable tbc cc.e spray system or IECI to provide protection against the tra).:.

the event of HPCI failure, by depressurizing the reactor vencel rapidly enough to actuate the core sprays or LPCI. Either of the two core spray systems or IICI provide sufficient flow of coolant to limit fuel clad temperatures to will below clad n: cit and to assure that core gecmetry Of remains intact.

these three, Three s'a r etv/ relief valveu are included in the automatic pressure relief system. See l only two are required to provide sufficient capacity for the automatic pressure reliefsysten.

section 4.4 and 6.2.5.3 FSAR.

3. BCIC tinuous mkeup water to the reactor core when the reactor The purcping The RCIC systejis girovided to supply cis isolated from the turbine and when the feedvater system is not a Ler level above the core without any capacity of the RCIr system is sufficient to mintain the vIf the water level in the reactor vessel decreases to the RCIC other vater system in operatpn. The system may also be mnually initiated at initiation level, the system autcrmtically starts.

any time.

The HPCI syst.0m provides an alternate method. of supplying makeup water to the reactor should the normal feedvater become unavailable.

Therefore, the specification calls for an opern-I bility check'of the HPCI system shoald the BCIC system be found to be inoperable.

I

=

112 3.5 PASES ,

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4.0 SURVEILIANCF RFRUUTCETI'S t 3 0 LUIITIIN COIIDITIO!iS FOR OPERATIC!r l

h.6 PRIMARY SYSTD1 BOUITDARY 36 PRI11ARY SYSTD4 BOUTIDARY Applicability:

_ Applicability:

Applies to the periodic examination and testing Applies to the operati.7 status of the reactor requirements for the reactor coolant system.

coolant system.

Objective:

Objective:

To assure the integrity and safe operation of the To determine the ccndition of the reactor coolant l

- system and the operation of the safety devices reactor coolant system. related to it. I l

Specification:

Specification: l A. Thermal Limitations A. Thermal Limitations

1. During heatups and cooldownn recirculation

1. The average rate of reactor coolant loops A and B temperatures shall be per-l temperature change during nomal 0heatup manently recorded at 15 minutes intervale or cooldown shall not exceed 100 F/hr.

when averaged over a one-hour period.

2. The temperatures lir,ted in 4.6.A.1 shall

2. The pump in an idle recirculation loop be pemanently recorded subsequent to a shall not be started unless the temper- heatup or ecoldown at 15 minute intervals ature of the coolant within the idle re- until three consecutive readings are withir cireilation loop is within 50 F of the 5 degrees of each other.

reactor coolant temperature.

115 Ra 3 6/h.6

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=

V i

h.0 SURVEILIANCE hFRUIRrTEI.'TG 30 LDfITING CCNDITION G TOR OPEFATION (b) When the ecntinuous conductivity mor.1-tor is inoperable, e. reactor coolant sample should be takert at least once per shift an3 analyzed for conductiv-ity and chloride ion content.

h. If Specification 3.6.c.1, 3.6.c.2, and 3.6.

C.3 are not met, norntl orderly shutdown shall be initiated. .

D. Coolant Leakage D. Coolant Leakage Beactor coolant t,ystem leakage into the dry-Any time irradiated fuel is in the reacter vessel, well shall te checked and recorded at least and reactor coolant temperature is above 2120F, once per day.

reactor coolant leakage into the primary contain-ment from unidentified sources shall not exceed 5 gpm. In addition, the total rea: tor coolant syste'n leakage into the primary ec,nainment shall not exceed 25 g m. If these condit'ons cannot be l met, initiate an orderly shutdown ans.' have the re-actor placed in the cold shutdown cent.*' ion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

118 3 6/h.6 REV

(-

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SCRVEILIANCE REQUIREMENTS i

E. Safety / Relief Valves and Prompt Relief Trip (PRT) t E. Safety / Relief Valves and Pronpt Relief Trip System (PRT) System

1. During power operating conditions and whenever 1. a. A minimum of six safety / relief valves reactor coolant pressure is greater than 110 psig shall be bench checked or replaced and temperature is greater than 345 F. with a bench checked valve each re-fueling outage. The nominal setpoint

a. She safety valve function (self-actuation) of all operational safety / relief valves of six safety / relief valves shall be operable. shall be i1080 psig.

~

b. The solenoid activated re1ief function b. At least two of the safety relief (Antomatic Pressure Relief) shall be valves shall be disassembled and operable as required by Specification 3.5.E. inspected each refueling outage.

2. During reactor power operation, the prompt c. The integrity of the safety / relief

.elief trip (PRT) system function of six valve bellows shall be continuously safety / relief valves shall be operable monitored.

in accordance with Specification 3.2.F.

d. The operability of the bellows monitorin:

system shall be demonstrated at least once every three months.

2. Surveillance of the PRT System shall be as follows:

Item Frequency Valve operability Each Operating Cycle Simulated Automatic Actuation Test Each Operating Cycle 3.6/4.6 119 REV

h.O SURVEILT/UICE REQUIRE 7'EITT3 30 LDiITI;;G COITDITIOI!S FOR OPETATIOi!

F. Structural Integrity F. Structural Integrity

-imary syste:a The nondestructive inspections listed in Table The st uctural integrity of k.6.1 shall be performed as specified. The boundary shall be maintained a., e level re- results obtained from compliance with this l

t quired l'y the original acceptance standards specification will be evaluated after 5 years throughout the life of the plant. and the conclusions of this evaluation vill be reviewed with the AEC.

l G. Jet Pumps f G. Jet Pumps

• Whenever the reactor is in the Startup Whenever there is recirculation flow with the reactor in the Startup or Run modes, jet pump or Run modes, all jet pumps shall be oper- operability shall be checked daily by verify-able. If it is determined that a Jet pump is ing that all the following conditions do not inoperable, the plant shall be placed in a occur simultaneously:

cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1. The two recirculat. ion loop flows are unbalanced by 15% or more when the recirculatlor. pur pa are operating at the same speed.

2. The indicated value of core flow rate is 10% or more less than the value de-rived from loop flow measurements.

120 3 6/i;.6 RW 0 9/8/7c

Bases Continued s.6 and 4.6:

D. Coolant Leakage The former 15 gpm limit for leaks from unidentified sources was established assuming such leakage was coming '

f rom the primary system. Tests have been conducted which demonstrate that a relationship exists between the size of a crack and the probability that the crack will propagate. From the crack size a leakage rate can be determined.

For a crack size which gives a Icakage of 5 gpm, the probability of rapid prc,pagation is less than 10-5 Thus, an unidentified leak of 5 gpm when assumed to be from the primary system had less than one chance in 103,000 of propa-gating, which provides adequate margin. A leakage of 5 gp, is detectabic and measurable. The 2'. hour period allowed for determination' of Icakage is also based on the lor probability of the crack propagating.

The capacity of the drywell sump pumps is 100 gpm and the capacity of the drywell equipeent drain tank pumps is also 100 gpm. Removal of 25 gpm from either of these sumps can be accomplished with considerable , margin.

An annual report will be prepared and submitted to the AEC summarizing the primary coolant tc dry. ell leakage measurements. Other techniques fer detecting leaks and the applicability of these techniques tc the Monticelio Plant will be the subject of continued study. .

< E. Safety / Relief Valves and Prompt Relief Trip Testing of all safety / relief valves each refueling outage ensures that any valve deterioration is detected. A tolerance value of 17. for safety / relief valve setpoints is spccified in Section III of the ASME Boiler and Pressure Vessel Code. Analyses have been performed with all valves assumed set 17. higher (1080 psig + 17.) than the nominal setpoint; the 1375 psig code limit is not exceeded in any ca=e.

The safety / relief valves are used to limit reactor vessel overpressure and fuel thermal duty through prompt relief trip ar.d self actuation. -

The required safety / relief valve steam flow capacity is determined by analyzing the transient accompanying the mainsteam flow stoppage resulting from a postulated r31V Cd.osure from a power of 1670 tN . The analysis assumes a multiple-failure wherein direct scram (valve position) is neglected. Scram is assumed to be from indirect means (high flux). In this event, the safety / relief valve capacity is ass.;med to be 717. of the full power steam generation rate.

3.6/4.6 BASES 134 REV

Bases 3.6 and 4.6 (Continued)

The safety / relief valves have two functions; i.e. automatic vessel depressurization or over-pressure protection. We former is a solenoid actuated function (Automatic Pressure Relief) in which external instrumentation signals of coincident high drywell pressure and low-low water Icvel initiate ope,ing of the valves. %is function provides backup to the IIPCI system for small break protection and is discussed in Specification 3.5.E. In addition these valves can be operated manually.

%e over-pressure protection function utilizes six safety / relief valves, three of which are operated for the Automatic Pressure Relief function. All six valves are capable of direct, self-actuation or indirect, prompt-relief trip (PRT) actuation.

He primary overpressure protection (for ASME Code consideration) is provided via the pressure-actuated integral bellows and pilot valve that cause main valve operation for any plant event wherein valve setpoint pressure is attained. Article 9 of the ASME Pressure Vessel Code Section III, Nuclear Vessels, requires that the bellows be monitored for failure since this would defeat the function of the safety / relief

,  ! valve.

Provision also has been made to detect failure of the bellows monitoring system. Testing of this  !

system quarterly provides assurance of bellows integrity.

When the setroint is being bench checked, it is prudent to disassemble one of the safety / relief  ;

valves to examine for crud buildup, bendin6 of certain actuator members or other signs of possible l deterioration.

F. Structural Integrity 1

A pre-service inspection of the cc=ponents listed in Table h.6.1 has been conducted to assure that the system is free of gross defects and as a reference base for later inspections. In addition, the facility has been designed such that gross defects should not occur throughout life. he inspection program was based on the proposed AmE Code for In Service Inspection of Iluelear Fenctor Coolant Systems which was follosed except where accessibility for inspection was not provided. his inspection provides further assurance inae, tWess defects are not occurring after the system is in service. lhis inspection vill reveal problem areas st mld they occur before n leak develops. -

1 3.6/h.6 InSES 135 REV O 9/8/70

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

i Bases Continuad 3.6 and 4.6:

Design confimation and constnietion adequacy will be demonstreted ( ariNC the plant startup and power ascension test program. As part of this program, cold and hot alb ~at.on tests en certnin reactor vessel internals vill be perfomed. %e tests, described in a ' c c uer to Dr. P. A. Morris, dated March 5,1070, are designed to obtain confimatory data on the de sign features of Monticello as j compared to Dranden Unit 2 design. Thus, the basis for the Monticello vibration test prograr is predicated on obtainin6 satisfactory data which confins cemon desirn features from earlier IT,JR l

plants such as Dresden Unit 2. In the event that data from these earlier plants era not availatle bercre routine power cperation of Monticello, the matter vill be reviewed by the AEC.  !

Ee program outlined in Table h.6.1 is limited to inspections of the prbnory coolant system. It is anticipated that the dato collected during the first five years of operation vill provide a suitable basis to evaluate the need for inspecting other portions of the facility (such as the main steam lines

, downstream of the main steamline isolation valves). %ese data along with the overall operating experiences vill be reviewed to detemine the inspection program to be implemented for the lifetime of the '

facility. %e resultc of this study together with the proposed lifetine inspection program vill be nutmitted to the AEC in accordance with Cpecification 6.7.C.3.

%e special inspection of the min feed and steam lines is to provide added protection against pipe whip. he Group I velds are celacted on the basis of an analysis that shows these velds are the highest stress velds and that due to their physical location, a break vould result in the Icast interference and maximum energy upon impact with the dryvell. Rese velds are the only ones which offer ,

any significant risk and vill be included in itture inspections as detemined by the study described nbove.  :

Group II velds are selected because without regard for the operatin6 stress levels and inter-fering equipent, they have sufficient theoretical energy to penetrate and vould propel tha pipe

• toward the contaiment. Rey are therefore included in the first inspection. I&on consideration of Impact angle, interfering equipment and distance pipe travels, no substantial risk is involved and i no extra inspection is needed.

In addition, extensive visual increction for leaks vill be made periodically on critical systems.

Le inspection program specified eneempasses the major nrens of the vessel and piping systems within the dryvell. %e inspection period 13 based on the observed rate of growth of defects from fatigue studies sponsored try the AEC. These studies show that it requires thousands of stress cycles at -

l stresses beyoni any expected to occur in a reacter system to propagate a crack. The test frequency 4

established is at intervals such that in comparison to study results only a small number of stress cycles, at values belov limits vill occur. On this basis, it is considered that the test frequencies are adequate.

3.6/4,6 BASES 136 REV

Bases Continued 3.6 and 4.6:

~

'Ihe t:ye of inspec+1cn plma al for mch ecq cnent *lepenls on Icentir>r., acessibility, and type of expected dnfect. Direct vicual exaninstion ic proposed wherever ;ossible since it ic sensitive, fact ,nd reliable. !4agnetic porticle ani 31guld penetrant ircrecticnc are planned vuera pr,etical, ,nd where added sensitivity is re ;uired. Ultracenic testim and radiogr,phy rhs11 he ua-1 rhere derets can occur on cencealed surfaces.

The prompt relief trip (PRT) function of the safety / relief valves provides an anticipatory actuation of the safety / relief valves for transients involving turbine stop valve closure er turbine control valve fast closure. Although the PRT system is intended to provide anticipatory pressure relief for turbine trip transients with failure of the bypass valves, the PRT system is entirely independent of w the bypass system. The PRT system, by providing an anticipatory open signal to the safety / relief valves aids in maintaining fuel, thermal and pressure margins and has therefore been designed on the basis of Engineered Safeguards add meets tl e requirements of IEEE 279.

The PRT system is divided into two independent redundant channels which are programmed on a power level schedule into three modes of operation relative to the safety / relief valves coupled to it. The modes of operation are listed below:

Reactor Power PRT/S/RV's 8 57. 6

.70% 3 707. O Mode selection is automatic through a biasing signal based on turbine first stage pressure. The power level schedule has been established to ensure the full powar transients remain the most limiting and adecuate pressure and fuel thermal margins are provided.

The PRT system will not preclude safety / relief valve self actuation or Automatic Pressure Relief Operation.

Deactivation of the PRT system signal is effected through redundant interval timers and a low reactor pressure switch or the attainment of the self-actuation reset pressure, should the pressure remain above the se l f-actuation setpoint for a period exceeding the interval timer setting.

Spare safety / relief valves nay be installed that are adaptable to PRT service with minor interconnection changes to permit the substitutien for an inoperable PRT valve while operating in a reduced power mode.

The PRT System is incorporated to of fset changes in the scram reactivity insertion rate which occur with increasing exposure out to the equilibrium exposure. All transients have been analyzed to account for both the slower insertion rates and PRT installation in " Plant Changes to Accommodate Enullibrium " ore Scram Reactivity Insertion Chscacteristics". This report was submitted to the Commissionon January 23, 1974..

~

136A 3.6/4.6 BASES ver

e l L

Bases Continued 3.6 and 4.6:

The PRT system demonstrates substantial improvement in fuel thermal and pressure response to the PRT-coupled - i turbine and generator trip transients. This effect is manifested in the results of a turbine trip with bypass failure transient where prt reduces peak vessel pressure by 74 psi cnd heat flux by 137, compared with the same event without PRT. PRI provides ef fective compensation for effects on plant performance caused by the changing scram reactivity. The PRT 3-mode design provides flexibility in maintaining pressure and fuel thermal  ;

margins and minimizing the duty cycle on the safety / relief valves at low power. Positive disabling of the PRT action is insured by a timer with a nominal setting of 5 seconds and a lov vessel pressure signal set at 900-950 psig.

G. Jet pumps Failure of a jet pump no le accembly hold down mechanism, no :le asse-bly and/or riser, would lacrease the cross-sectional flow area for blowdown following the. design basis dcuble-ended lina treek.

'Iherefore, if a failure occurred, repairs nust le mvle.

The detection technique is as follows. With the two recirculation pie:ps balanced in speed to vithin

+ 5%, the flou rates in both recirculation loops vill be verified by Control Eocn monitoring instrunents.

Tr the two flow rate values do not differ by more than 10%, riser and no le assembly integrity has been verified. If they do differ by 10% or more, the core flow rate measured by the jet puy diffuser differential pressure system must be checked against the core flo.i rate derived from the cassured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10%

or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nor:le (or riser) and the plant shut down for repairs. If the potential blevdown flov area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will "run out" to a substantially hicher flow rate (approximstely 115% to 120% for a single nor le failure). If the two loops are talanced in flow at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates veuld be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leaksge path past i the core thus reducing the core flow rate. IM reverse flow through the inactive jet pump would still be indicated by 9 positive differential pressure but the net effect vould be a slight decrease (37 to 6%)

in the total core flow measured. This decraese, tocether with the loop flow increase, would rastit in a lack of correlation between meeural and derived coro flow rate. Finally, the affected jet pump diffuser i differential pressure signal would be roduced because the backflow would be less than the nor nal forward flow.

i 137 3.6/h.6 N rc

1

& 3 k.

w .

N LI141 tin 3 CONT 1ITIOU3 FOR OPERATIO!; b.O SURVEILIA' CE FILUIFE~E'TT3 3.0

6. If the specifications of 3.7. A cannot bc  ;

met, the reactor shall be placed in a cold shutdown andition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Standby Gas Treatment System B. $*andby Gas Treatment System

1. Except. as specified in 3.7.B.3 below, 1. Standby gas treatment system surveillance both eircuits of the standby can treat- l shall be performed as indicated below:

ment system shall be operable at all times when secondary containment a. At least once per operating cycle it integrity is required. shall be demonstrated that:

(1) Pressure drop across the combined high-efficiency and charcoal filters is less than 7.0 inches of water, and (2) Inlet heater output is at least 15 kw.

b. During each refueling outage prior to refueling, whenever a filter is changed, whenever work is performed that could affect filter systems efficiency, and at intervals not to exceed six months between refueling outages, it shall be demonstrated that:

(1) ne removal efficiency of the installed particulate filters for particles having (

a mean diameter of 0.7 microns shall be 148 3 7/h.7

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

, e

% \

6

~

,I l

. 30 LIMITIn3 CONDITIONS FOR OPERATIOH L.O fiUR?EILIAIICE FFeUIRma.'TS equal to or greater than 97/ based on an in-place dioctyl phthal.'te (DOP) test.

(2) The renoval efficiency of the charcaal filters is not less than 99% far freon based on a freon test.

c. At least once each five years remvable charcoal cartridges shall be removed and l

adsorptien shall be demonstrated.

d. At 1mst oney per operating cycle autcentic initiation of each branch of the standby 6as treatment system shall be demonstrated.

2. From and after the date that one circuit 2. When one circuit of the standby gas treatment of the standby gas treatment system is system becomes inoperable, the opemble made or found to be inoperable for any 71rcuit including its emergency power source reason, reactor operution is pemissible' s'nll be demonstrated to be opemble imadi- ,

only during the succeeding seven days un- a t ?ly. The operable circuit of the Standby less such circuit is sooner mde -operable, Gas Treatment System shall be demonstrate 1 provided that during such seven days all to be operable daily thereafter.

active components of the other standby ,7as  ;

treatment circuit includin6 its emergenty  ;

[

i power source shall be operable.

3 If this condition cannot be met, procedures [

shall be initiated immediately to establish l

, the conditions listed in 3 7.c.1. (a) throu,.h l (d), and compliance shall be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. ,

l 3 7/h.7 lh9

, re/

I  :

.-e < - - _ _ - - - _ _ _ _ . _ _ _ _ _ _

e i

e

/

Bases Continued: "$

4.7 High efficiency particulate filters are installed before and after the charcoal filters to minimize potential release of particulates to the environment and to prevent clogging of the iodine filters.

An efficiency of 997. is adequate to retain particulates that may be released to the reactor building following an accident. This will be demonstrated by in-place testing with DCP as testing medium.

l Individual filter units will be tested and certified to have a removat efficiency of equal to or I greater than 99% for particles having a mean diameter of 0.3 microns at the tLee of purchase.

The test interval for filter efficiency was selected to minimize pluggin6 of tha filters. In addition, ,,

retention capacity in terms of microcuries of iodine per gram of charcoal will be demonstrated. This will be done by removing small test cartridges of the same charcoal filter material. These cartridges complement the charcoal filter system and will be availabic for withdrawal and testing. These tests will normally be performed every five years unless filter efficiency ceriously deteriorates. Since shelf lives greater than five years have been demonstrated, the test interval is reasonable.

D. Primary Containment Isolation Valves Those large pipes comprisin6 a portion of the reactor coolant system whose . failure could result in uncovering the reactor core are supplied with automatic isolation valves (except those lines needed for emergency core cooling . ystem operation or containment cooling). The closure times specified herein are adequate to prevent loss of more coolant from the circumferential rupture of any of these lines outside the containment than from a steam line rupture. Therefore, this isolation valvo closure time is sufficient to prevent uncovering the core.

In order to assure that the doses that may result from a steam line break do not exceed the 10 CFR 100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses suggest that fuel rod cladding perforations -

would be avoided for main steam valve closure times, including instrument delay, as long as 10 5 seconds.

However, for added margin the Technical Specifications require a valve closure time of not greater than 5 seconds.

The primary containment isolation valves are highly reliable, have low service requirement, and most are normally closed. The initiating sensors and associated trip channels are also checked to denon-strate the capability fer automatic isolation. Peference Section 5 2.2.4.3 and Table 5-2-3 FSAR.

The test interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10~7 that a line will not isolate. More frequent testing for valve operability results in a more reliable system.

h.7 BASES 166 fa

AEC DTSJ 1UTION FOR PART 50 DOCKET MATEF  :. 1726 s

(TDiPORARY FORM) CONTROL NO:

FILE:

IROM: DATE OF DOC DATE REC'D LTR MDiO RFT OTl!TR ort hern Stat es: Pove t'onpany inneapolis, ?tinn. 55401

v. L.O. ?tr'e r 3-1-74 3-4-74 X TO: ORIG CC OTIER SENT AEC PDR, XXX SENT LOCAL PDR XXX F. O'Lentv 3 minned CLASS 'UNCLASS PROP IhTO INPUT NO CIS REC'D DOCTIT NO:

xxx 40 50-263 DESCRIPTION: ENCLOSURES:

r trann the followinr...... Proposed channen to tech specs, notarized 3-1-74, consist of rev & add'1 pgs, tables &

fins to the tech speen..

[C N'[IU'. J (40 cvs encl rec'd)

PLANT NAME:  !!ont icello -[],'T ]}{lf73 FOR ACTION /IhTORMATION 3-4-74 .i n BUTIIR(L) SCIMENCER(L) /ZIDiANN(L) REGAN(E)

W/ Copies W/ Copies W/ 7 Copies W/ Copies CLAPZ(L) STOLZ(L) DICRER(E)

W/ Copics W/ Copies W/ Copies W/ Copies GOLLER(L) VASSALLO(L) }J11GHION(E)

W/ Copies W/ Copics W/ Copies W/ Copies KNIEL(L) SCRDIEL(L) YOUNGBLOOD(E)

W/ Copies U/ Copies W/ Copics W/ Copies I?'TER"AL DISTRT LUTION

• PIG FkLL TECH REVIE'J DZ'; TON A / T I"D

1. ,"AEC'PDR IENDRIE GRIMES LIC ASST BRAITMA'i tGC, ROOM P-506A SCHROEDER GAMMILL / DIGGS (L) SALTZM/d 4!UNTZINC/ STAFF MACCARY KASTNER GEARIN (L) B. HURT CASE 10;ICHT BALLARD GOULLOURNE (L) PLAS GIAMBUSSO PAWLICRI SPANGLER LEE (L) 11CDONALD BOYD SHA0 MAIGRET (L) /'DUEE v/Inpu t MOORE (L)(P'R) STELLO ENVIRO SERVICE (L)

DEYOUNG(L)(ETR) HOUSTON MULLER 51EPPARD (E) INI'0 er$ROVHOLT (L) NOVAR DICKER SMITH (L) C. MILES P. COLLINS ROSS 13;IGHTON TEETS (L) B. FING DENISE IPPOLITO YOUNGBLOOD WADE (E) ./A Cabell o,dtEG OPR TEDESCO REGAN WILLIAMS (E)

FILE & REGION (3) LONG PROJECT LDR WILSON (L)

MORRIS LAINAS S. REED (L)

STEELE BENAROYA HARLESS VOLIEER 2 EXTERNAL DISTRIBUTION 81fl

1. 1 - LOCAL PDR *11nnenpolis. " inn,

/1 - DTI E( AI ' *" AT HY )

f (1)(2710)-NATION \L 1AB'S 1-FDR-SAN /LA/NY d - USIC(nUCiUuAN) 1-ASLEP(E/W Eldr,,Pe. 529) 1-GERALD LELLDUCHE 1 - ASLE(YC.~.:/3 AYRE / l-U. PENNINGTC:,, L . E-201 GT BR007. HAVEN N'T IAL WOODARD/"H" ST. 1-CONSULTANT' S 1- AGMED(Rutn Gussman)

' I6 - CYS ACRS 1;OLDING NE"O!ARR/ DLC'.E /AGB ABI AN Di- B- 12 7, C .

1-GERALD ULRIKSON. . 0PNL l-RD. .rULLER. . F-309 GT

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