B11663, Proposed Tech Specs Changing Max Critical Power Ratio Operating Limits & MAPLHGR Curves for Cycle 10 & Correcting Typos

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Proposed Tech Specs Changing Max Critical Power Ratio Operating Limits & MAPLHGR Curves for Cycle 10 & Correcting Typos
ML20134K408
Person / Time
Site: Millstone Dominion icon.png
Issue date: 08/26/1985
From:
NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20134K392 List:
References
B11663, NUDOCS 8508300182
Download: ML20134K408 (32)
v • d • e


Text

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1 Docket No. 50-245 Bil663 n

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Attachment No. 1 Proposed Technical Specification Changes for Millstone Unit No. 1, Reload 10 i

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Figure 3.11.la - MAXIMUM AVERAGE PLANAR LINEAR llEAT GFNERATION RATE (MAPLIIGR)

VERSUS PLANAR AVERACE EXPOSURE. FUEL TYPE P8DRB282 Amendment No. 73, 87, 98 3/4 11-2 ,

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VERSUS PLANAR AVERAGE EXPOSURE. FUEL TYPE P8DRB28311 Amendment No. 98 3/4 11-3

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VERSUS PLANAR AVERAGE EXPOSURE. FUEL TYPE BP8DRB300 Amendment No. 98 3/4 11-4 4

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TABLE 3.11.1 s

OPERATING LIMIT MCPR'S FOR CYCLE 10 .

(OPTION B)

BOC 10 TO EOC 10 EOC 10 TO 70% COASTDOWN FUEL TYPE 1.42 1.42 , P8 x 8R ' '

1.42 1.42

  • BP8 x 8R OPERATING LIMIT MCPR'S FOR CYCLE 10 (OPTION A)

BOC 10 TO EOC 10 EOC 10 TO 70% COASTDOWN FUEL TYPE l.48 1.48 P8 x 8R 1.48 1.48 BP8 x 8R

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Amendment No. 98 -

3/4 11-10 i

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LIMITING OCNDITIN Em OPEPATIN SURVEIIIRCE RECUIREMENT 3.1 REACIm HU1wnm SEIH1 4.1 IDCIT MU1wrIN SEID1

  • Arplicability: ppplicability:
  • Applies to the instnmentation ard associated devices Applies to the arveillance of the instrumntation and which initiate a reactor scran ard provide autanatic associated devi s which initiate reactor scran and provide isolatim of the Ibactor Protection Systen buses fran autmotic isolation of reactor protecticn systen buses frun their power supplies. their pmer supplies.

(bjective: Cbjective:

To assure the operability of the Reactor Protection Systan. To specify the type and frequency of surveillance to be '

applied to the reactor protection instnmentation.

Specification:

A. 'Ihe setpoints, minim.m rutber of trip Specification:

system, ard mininta runber of instnment channels that nust be cprable for each positim of the reactor node A. Instnrnentation systen shall be functionally tested'ard switch shall be as given in Table 3.1.1. calibrated as irdicated in Tables 4.1.1 and 4.1.2, respectively.

  • B. Respmse Tine The time frcm initiation of any channel trip to the B. Ibily durirg reactor power operation, the maxintm de-energization of the scran solenoid relay shall not fraction of limiting power density shall be diecked ard exceed 50 milliseoJnds, the APRM scram and rod block settings given by the evaluations in Specifications 2.1.2A and 2.1.2B shall be C. Reactor Pmitectim Systan Ibwer ManitorirrJ determimd to be valid.

wo Ris electric power nonitoring channels for each inservice RPS FG set or alternate power supply shall be C. 'Ihe RPS electrical protection assenblies shall be cperable at all times except as follows: determined operable as follw s:

1. With cre RPS electric pwer mJnitority channel for an 1. At least once per 6 months by performance of a 4

inservice RPS FG set or altermte power supply QWNEL FINCTICNAL 'IEST, ard inoperable, restore the incoerable channel to OPEPABT status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or rurtwe the associated RPS FG 2. At least cnoe per 18 mtnths by davonstratirg the set or alternate power sugply frun service. OPERABIIIIY of wer-voltage, under-voltage ard under-ftnquency Irotective imtnmentation by

2. With both RPS electric power nonitoring channels for an performance of a OWNEL CALIBRATIm incltding inservice RPS FG ' set or alternate power su[ ply sinulated autonatic actuation of the protective inoperable, restore at least one to OPERABLE status relays, trippirg logic and output circuit breakers, within 30 minutes or tutove the associated RPS BC set or' and verifyirg the followirg mtpoints:

alternate power sugply frun service. ,

a. Over-voltage 1 (132)VAC,
b. Under-voltage > (108) VAC,
c. Under-frequency > (57)Hz, ard
d. Time-delay 1 (4.0) seconds.

7 :9 TABLE 3.2.3

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Ill51RUMCHTATION TilAT IH111A105 R00 BLOCK HInfmum Number of *

. Operable instrument Channett pe Trip System (t) l Instrument Trip l.svel Setting I III ApnH Upscale (Flow Blased) See Specification 2.1.2B 1

l ApRH Downscale > 3/125 full Sc' ale

,I (6) Rod Block Honitor Upscale (flow Biased) 1 . 65W + 42(2) j (6)

Rod Block Honitor Downscale > ,3/125 full Scale 3 1RH Do'wnscale (3) >

,3/125 Full Scale 3 IRH Upscale 1100/125. full Scale

  • 2 SRH Detector not in Startup position (4) 2 (5) SRH Upscale i 10 5counts /sec.

Is I4 Scram Discharge Volume - Water level High 1g]"jesn , , ,

e lower cap to 1

Scram Discharge Volume - Scram Trip Bypassed N/A (1) for the Startup/liot Standby and Run positions of the Reactor Mode Selector Switch, there shail be two operable or tripped trip systems for each function except the SRH rod blocks; IRH downscale are not operable in the RUN position and APRM downscale need not be operable in the Startup/ Hot Standby mode, if the first. colunn cannot be met for one of the two trip systens, this condition may exist for up to seven days provided that during that

  • time the operable system is functionally tested immediately and daily thereafter; if this condition lasts 8 -

Ionger than seven days, the system shall be tripped. If the first column cannot be met for both trip systems, the systems shall be trlpped.

(2) W is the reciEculation flow required to achieve rated core flow expressed in percent.

(3) IRH downstale may be bypassed when it is on its lowest range.

(4) This _ functio ~n may be bypassed when the count ra te is > 100 cps or when all IRH range switches are above position 2.

(5) One of these lrlps may he bypassed. The SRH function may be bypassed in the higher IRH ranges when the IRH

, upscale rod block is . operable.

Amendment No. 87 . 3/4 2-5 (correction)

l 4

SItRvLILLANCE REQUIRtHtNT LlHITING CONDIT10ff f0R OPERATION

  • untrol rods or in the event powei opera-p wer operation. 'If a parti. illy or fully '

tion is continuing.with one fully or withdrawn control rod drive cannot be moved with drive or scrain piessiere the partially withdrawn rod which cannot be , ,

reactor shall be brought to a shutdown moved and for which control rod drive ,

condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> unless investiga- mechanism damage has not been ruled out.

tion demonstrates that the cause of the lhe surveillance need not be completed fallu.'e is not due to a failed control rod within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of inoperable drive mechanism collect housing. rods has been reduced to less than three

, and if it has been demonstrated that Control Rod Withdrawal control rod drive mechanism collet housing 8.

failure is not the cause of an ininovable

1. Each control rod shall be coupled to its control rod.

drive or completely inserted and the control rod directional control valves D. Control Rod Withdrawal disarmed electrically. Ilowever, for purposes of removal of a control rod drive, 1. The coupling integrity shall be verified as many as one drive in each quadrant may for each withdrawn control rod as follows:

be uncoupled from its. control rod so long as the reactor is in the shutdown or a. when the rod is fully withdrawn the

- refuel condition and Specification 3.'3.A.1 first time subsequent to each' is met. refueling outage or af ter maintenance, observe that the drive does not go

2. Jhecontrolroddrivehousingsupport to the overtravel position; and system shall be in place during power operation and when the reactor coolant . b. when the rod is withdrawn the first system is pressurized above atmospheric time subsequent to each refueling pressure with fuel in the reactor vessel, outage or after maintenance, observe unless all control rods are fully inserted discernible response of the nuclear and Specification 3.3.A.1 is met. instrumentation 1 however, for initial rods when response is not discernible, ,

subsequent exercising of these rods after the reactor is critical shall

  • be performed to verify instrumenta-Lion resporise.
2. ihe control rod drive housing support system shall be inspected af ter reasumbly and the results of the inspectinn shall be recorded.

Amndmen t flo. J.',' 22, 76 Y

_ _ _ _ - - - _ . - - _ - _ . - _ D. --..Q

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'(.-' 'LIMITI,NG' CONDITION FOR '0PERATION. ' ('$3 " ' SURVEILLANCE REQUIREMENTS

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D. Coolant Leakage D. Coolant Leakage Any time irradiated. fuel is in the reactor ves- Reactor coolant system leakage into the sel.. reactor coolant leakage into the primary primary containment shall be checked and containment from unidentified sources shall not recorded at least once per day. ,

exceed 2.5 spm. In' addition, the total reactor coolant system leakage into the primary con - E. Safety and Relief Valves

- tainment shall not exceed 25 gpm. If these .

- conditions cannot be met, initiate an orderly 1. Three of the relief / safety valves top shutdown and havd the reactor in the cold works shall be' bench checked or replaced ,

with a bench checked top works each re- '

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

- fueling outage. 'All six valves top works E. -Safety and Relief Valves 'shall be checked or replaced every two refueling outages. The. set pressure

- shall be adjusted to correspond with a steam set pressure of;

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1.. During power operation or whenever the reactor coolant pressure is greater.than 90 psig with .

irradiated fuel in the reactor vessel, the No. of Valves .' Set Point (psigl safety valve function of the six relief / safety s valves shall be operable, except as .specified in 1 1095 i 1%

3.6.E.5 below. (The solenoid activated relief 1 1110 i 1%

function of the relief / safety vaives shall .

4 1125 i 1%

be operable as required by Specification 3.Q.D.).

2. If Specification 3.6.E.1 is not. met, initiate 2. At least one of the relief / safety valves

, an orderly shutdown and have the reactor shall be disassembled and inspected each l coolant presaure below 90 psig within 24 refueling outage.

hours.

3, During each operating cycle with the reactor

3. When the safety / relief valves are required at low pressure, each safety valve shall be to be operable per Specification 3.6.E.1, manually ppened until operability has been the Valve Position Indication shall be verified by torus water level instrumenta- ,

t operable. Two of the six, channels may be Lion, or by the e Valve Position out of service provided backup indication Indication System, or by an'aud161e discharge for the affected valves is provided by the detected by'an individual located outside Valve Discharge' Temperature Monitor.' the torus in the' vicinity of each discharge.

3/4 6-5 Amendment No. JHf 99

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s 'e SilRVEILLANCE REQlllREMENT.

LIMITING CONDITION FOR OPERATION .

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2. If Specification 3.6.11.1 cannot be met, ~

one recirculation pump shall be tripped.

Operation with a single' recirculation pump is permitted for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the recir-culation pump is sooner made operable. If the pump cannot be made operable, the ,

reactor shall be in cold shutdown within

?4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />., .

3. The reactor shall not be operated unless '

the equalizer ljne'is isolated.

4. With the mode switch in the startupfhot standby or run mode, operation witi.out forced circulation shall not be permitted.

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e 3/4 6-11 Amendment No. If. 34 -

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'UnvfittAHCE REQulRlHItil tIHlitHG CONDITION FOR OPERATION

c. During the monthly generatur te.t. ,

the diesel fuel. oil transfer pump.

5. ' All station and switchyard 24 and 125 volt shall be-operated.

batteries and associated battery y chargers .-

are operable. ~ 2. Gas Turbine Generator

  • When the mode switch is in Run, the availability '
a. The gas turbine generator shall be H.

of power shall be as specified in 3.9.A. except fast started and the output breasess closed withlp 48 seconds once a si.onth as specified below:

  • to demonstrate opera tional readine.s.

1.

From and af ter the date that incoming power The test shall continue until ti.e.

is available frod only one 345 kv line, gas turbine and generator are atfull load reactor operation is permissible only equilibrium temperature atUse of this unit to su durinif the succeeding seven days unless an output.

additional 345 kv line is sooner placed in power to. the system electrical net- ,

  • work shall constitute an acceptable l service. demonstration of operability.

er '

2. .from and af ter the date that incoming pow .

is not available from any 145 kv lir.e. b.

During each refueling outage, the reactor operation shall be permitted pro- conditions under which the gas turbine-vided both emergency puuer sources are generator is require'd wil) be spuulated operating andThe theHRC isolation condenier and a test conducted to verify that shall be notified, system it will start and be able to accept is operable.

i within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the precaution'; to be einergency loads wi11:19 48 seconds.

taken during this situation and the plaats The

' for restoretion af incoming power. ,

U. 11a t teries minimum fuel supply for the gas turbine during this situation shall be maintained 1 Station Batterles above 20,000 gallons.

4.

Every week the specific gravity and .

voltage of the pilot cell and tem-

3. f rom and af ter the date that either emer- perature of adjacent cells and overall gency power source or its associated I.us is made or f ound in be inoperable for any battery voltage shall he measured. ,'

reason, reactor operation is _ permissil.leaccording s

b. Every threeto Specific'ation months the measurements 3.5.F/4.'

such emergency power source and its hus are Sooner shade epul'alile, provided th.ht shall be made of voltage of each o

during such time two of f site lines (345

  • 9favity of each cell and toi.periture h '

or 27.6 kv) . ire operat.le. of every (lith cell.

t 3/4 9-2 Amendment'llo. //, //, 76 *

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SURVEILLANCE REQUIREffENT LitflTING CONDITION FOR OPERATION . _ _ . _

B. Core !!onitoring B. Core !!cnitoring I. Prior to making any alterations to the lore, During core alterations two SRil's shall be oper- the SRH's shall be functionally tested and able, one in the core quadrant where fuel or checked for neutron response. Thereafter, ,

control

  • rods are being moved and one in an the SRH's will be checked daily for response adjacent quadrant, except as specified in Para- when core alterations are being ma,de,

-graphs 3.and 4 below. For an SRif to be considered e op'erable, the following conditions shall be 2. Prior to spiral unloading or reloading, the-satisfied: ~

SRfl's shall be functionally tested. Prior

1. The SRif shall be inserted to the normal to spital unloading, the SRH's should also

, operating level. (Use of special movable be checked for neutron response.

dunking type detectors during fuel loading or major core alterations in e place of agormal detectors are permissible as long as the detector is connected into ,

the normal SRH circuit.)

2. The SRif shall have a minimum neutron induced count rate of three per second with all rods fully inserted in the core. .
3. Prior to uitloading, the SRH's shall be proven operable as stated above, however, i during. spiral unloading', the count rate may drop below 3 cps.

4e If required, special movable dunking type '

detectors can be inserted into the core, prior to reloading fuel assemblies into e the central core region (with all' control rods inserted). Before the ninth fuel assembly is loaded into the core in the close proximity of the movable dunking chambers or the SRM;s Paragraph 3.10.B.1 I and 2 apply. ,

3/4 10-2 Amendment No. 98 T

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liigh is anradiationindication levels in the main of leaking fuel. steamline tunnel above that due to the no .

A scram is i

background. The purpose of the scram is initiated whenever such radiation level exceeds seven times n excessive release of radioactive materials.to reduce the source of such radiation to the extent necessary to prev .

ent is prevented provided the by limittheforaira ejector oII gas monitors which cause an iDischarge of excessive amounts of radioactivity to th 15 minute period specified in specification 3.8 is not exceeded ** solation of the main co

, The full open main in steamline three outisolation valve closure. scram of four lines.

when the valves close. is set to scram when the isolation valves are 10% closed from Section 11.3.7 FSAR. By scramming at this setting the resultant transientThis sciam anticipates the pre is insignificant. Ref.

2 A reactor mode switch is provided which actuates or bypasses I

particular plant operating status. the various scram functions appropriate to the Ref. Section 7.2 FSAR.

The manual scram function is active in all modes, .

9 rods duting all modes of' reactor operation. thus providing for a manual means 'of rapidly inserting control i

The IRH and Standby and Startup/llot APRif systemmodes. provide protection against excessive power levels a d n short reactor periods in the refuel level information during startup but has no scram functions.A source range monitor (SRN) system is also prov 4

and Startup/Ilot Standby modes. eutron is not required in the Run mode. In the power range the APRM provides the required protections; thusThu

, the IRN system I

The high reactor pressure, high drywell pressure, be operational for these modes of reactor operation. scrams . s're They are, required therefore, fortoStartup/li required The requirement to have all scram functions except those listed in Note and Shutdown mode is to assure that shifting to a 8 of T bl e :3.1.1 operable' in the Refuel the need for the reactor protection system. the . Refuel mode during reactor power' operation does not di sinish As indicated in Note 11 of Table 3.1.1, no trip functions are required to be operable if all contlol rods are fully inserted, valved out and el assures maximum negative reactivity insertion. ectrically disarmed, since this condition I i

    • Per errata sheet dated 10-7-70

! Arendment No. 98

B3/4 1-3 1

_, _. . . . ._ _ - _ - . _ . _ .. ~ .. > -.

1.2 Bases ,

, In addition to reactor protection instrumentation which initiates a reactor scram, protective -instrmentation' has been provided, which initiates action to mitagate the consequences of accidents which are beyond the operqtor's-ability to control or terminates operator errors before they result in serious consequences. - This set of specifications provides the limiting conditions of operation for the primary systen isolation function, initiation of the emergency core cooling system, control rod block and standby gas treatment systenn The objectives of the specifications are to assure the effectiveness of the protective instrmentation when reifuired by its capability to tolerate a signal failure of any canponent of such systems even during periods when portions of such systens are out of service for maintenance. When necessary, one channel may be made inoperable for brief intervals to

conduct required functional tests and calibrations and to prescribe the trip settings required to assure adequate performance.

I Isolation valves are installed in those lines that penetrate the primary contalment 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 initiated by protective instrmentation shown in table 3.2.1 which senses the conditions for which isolation is required. Such instrmentalon must be available whenever primary containment integrity is required.The objective is to isolate the primary containment so that the guideline values of 10 CFR I 100 are not exceeded during an accident. e The instrumentation which initiates primary system isolation is connected in a dual bus arrangement. %us, the discussion given in the bases for Specification 3.1 is applicable here.

< %e low reactor water level instrinnentation is set to trip when reactor water level is 127 inches above the top of the active fuel. %is trip initiates closure of Group 2 and 3 primary containment isolation valves but does not trip the recirculation pumps. Ref. Section VII-4.4 FSAR. For a trip setting of 127 inches above the top of the active fuel and a 60-second valve closure time the valve will be closed before core uncovery occurs even for the maximm breuk in the line; and therefore, the setting is adequate. ,

The low low reactor water level instrtsnentation is set to trip when reactor water level is 79 inches above the top of the active fuel. This trip initiates closure of Group 1 primary containment isolation valves, Ref. Section VII-4.4 FSAR and also activates the ECC subsystems and starts the emergency diesel generator and the gas turbine generator and trips the recirculation pumps. his trip setting level was chosen to be high enough to prevent ;

spurious operation but low enough to initiate ECCS operation and primary system isolation so that post accident cooling can be effectively accanplished and the guideline values of 10 CFR 100 will not be violated. For the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, ECCS.

initiation and primary system isolation are initiated in time to meet the above criteria.The instrumentation also covers the full range or spectrtan of breaks and meets the above criteria. Ref. Section VI-2.7 FSAR. The Isolation Condenser system has been added to the ECC system to insure that cladding integrity is maintained for postulated small break WCA conditions in the recirc. discharge piping with a gas turbine failure and LPCI injection into the q damaged loop.

(1) NEDO-24085-1, Ioss-of-Coolant Accident Analysis Report for Millstone Unit 1 Nuclear Power Station.

B 3/4 2-1

F i.. j -

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Illgh pressure actuation of the Isolation Condenser (IC) will be a backup to direct activation on Low-Low Invel; similar to other ECCS systems. Activation is based on the high pressure signal (1085 PSIG for 15 seconds) which.

, occurs after MSlV closure on Low-Low water level, SRV_ actuation, and subsequent repressurization. The activa-tion of the IC requires only the opening of normally closed valve 10-3 in the condensate return. line.

This valve is powered by the safety-grade DC battery.

grade AC or DC power and are_ al so used 'for containmentAll valves inAll isolatioh. theare system are in nonnally powered byposil the open safety-loi (other than 10-3). The IC systen is safety Class 2 and is scismically' qualified. The shell side water volume is surficient for approximately 30' minutes of operation at rated conditions without makeup. Two sources of makeup are available. For small break mitigation, less than 10 minutes of operation is required, and generally at less than rated conditions.

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. D 3/4 2-2a Amendment if0.67

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Two sensors on the isolation condenser supply and return lines are provided to detect line failure and actuate isolation action. The sensors on the supply and ret. urn sides are arranged in a 1 out of 2 logic and to meet the single failbre criteria, all sensors and instrumentation are required to be operable. The isolation settings and valve closure times are such as to prevent core uncovery or exceeding site limits.

The instrumentation which initiates ECCS action is arranged in a dual bus system. As for other vital instrumenta- '

tion arranged in this fashion, the Specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.

The control ro'd block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to < l.07. The trip logic for this function is I out of n; e.g. , any trip on one of the six APRH's, eight IRH's, or four SRH's will result in a rod block. %e minimum instrument channel requirements assure suf ficient instrumentation to assure the single f.iilure criteria is met. The minimum instrument channel requirements for the IRH and RBH may be reduced by bne for a short period of time to allow for maintenance testing and calibration.

I

' The ArRH rod block trip is flow biased and prevents significant approach to McPR-1.07 especially during operation at reduced flow. The APRM provides gross core protection, i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that fuel damage limits are not exceeded.

The RBH provides local protection of the core, i.e., the prevention of fuel damage in a local region of the core, for a single rod withdrawal error. ne trip point is flow hissed. The worst case single control rod withdrawal error has been analyzed for the initial core and also prior to each reload; the results show that with sp'ecified trip settings, rod withdrawal is blocked within an adequate margin to fuel damage limits. His margin varies slightly from reload to reload and, thus, each reload submittal contains an update of the analysis. Below % 70%

power, the withdrawal of single control rod results in HCPR > l.07 without rod block action, thus requiring the RBH systemt 'o be operable above 30% of rated power is conservative. Requiring at least half of the normal LPRH inputs from each level to be operable assures that the RBH response will be adequate to prevent rod withdrawal j errors.

The IRH rod block functions assure proper upranging of the IRH system, and reduce the probability of spurious scrams during startup operations.

l A downscale indication on an APRH or,IRH is an indication the instrument has failed or the instrument is not sensitive enough or the neutron flux is below the instrument response threshold. In these cases the instrument will not respond to changes in control rod motion and thus control rod motion is prevented. ne downscale trips are set at 3/125 of full scale.

AmendmentNo.)ff,61' B 3/4 2-3 i

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O'

4.2 Bases

1 The instrumentation listed in Table 4.2.1 will be functionally tested and calibrated at regularly scheduled intervals. Although this instrumentation is not gAnerally considered to be as .important to plant safety as the Reactor Protection System, the same design reliability goal of 0.99999 is generally applied for all applications of (1 out of 2) X (2) logic. Therefore, on-of f sensors are' tested once/3 months, and bl-stable tript associated with analog sensors and amplifiers are tested once/ week.

Those instruments which,- when tripped, result in a rod block have their contacts arranged in a 1 out of n logic, and all are capable of being bypassed. For such a tripping arrangement with bypass capability provided, the is an optimum test interval that should be maintained in order to maximize the reliability of a given channel. This takes given by:

account of .the fact that testing degrades ~ reliability and the optimum interval between tests is approximately i=I /2t fr

, where i = the optinum interval between tests I , ,

t = the time the trip contacts are disabled from performing their function while the test is in progress

r. = the expected failure rate of the relays.

To test the trip relays requires that the channel be bypassed, the test made, and the system returned to its initial state. It is assumed this task requires an estimated 30 minutes to complete in a thorough and workmanlike manner and that the relays have a failure rate of 10-6 failures per hour. Using this data and the above operation, the j optimum test interval is: '

i l

i= J2(0.5) = 1 x 103 hous '

j 10-6 = 40 days The sensors and electronic apparatus have not been included here as these are analog devices with readouts in the control room and the sensors and electronic apparatus can be checked by comparison with other like instruments. The checks which are made on a daily basis are adequate to assure operability of the sensors and electronic apparatus, and the test interval given above provides for optimum testingmf'the relay circuits.

i The above calculated test interval optimizes each individual ch6nnel, considering it to be independent of all others.

As,an example assume that there are two channels with an individual technician assigned to each. Each technician

( 2) UCRL-50451 Improving Availability and Readiness of Field Equi lmient Through periodic Inspection, Benjamin Epstein, Albert Shif f, Julv 16,1968, Pg.10, Equation (24), Lawrence Radiation Labora tory. ,

B 3/4 2-5 ,

9 3.

The peak fuel enthalpy content of 280 cal /gm is below the energy content at which rapid fuel dispersal and primary ence 1. system damage have been found to occur based on experimental data as is discussed in ref' Since Millstone Unit No. I has referenced the report, General Electric Standard Applications for-Reload Fuel (Reference 4), the assumptions regarding the control Rod Drop Accident are applica-

' ble to Hillstone Unit No.1. By using the analytical models described in this report coupled with .

conservative or worst-case input parameters, it has been detennined that for power levels less than 20%

of rated power, the specified Ilmit on in-sequence control rod or control rod segment worths will limit the peak fuel enthalpy content to less than 280 cal /gm. Above 20% power even single operator errors cannot result in out-of-sequence control' rod worths which are sufficient to reach a peak fuel enthalpy ,

content of 280 cal /gm should a postulated ' control rod drop accident occur.

Each core reload will be analyzed to show conformance to the follwoing bounding conditions:

a. Accident
b. reactivity curves equal to or less than those assumed in Reference (4).
c. Doppler Up reactivity coefficients equal to or more negative than those assumed in Reference (4) to 0.02AK, .

Reference (4). actual scram reactivity feedback function equal to or greater than data presented in If the above conditions are all net, these conditions is not met, then the reload is within the generic RDA analysis. If any one of' to demonstrate compliance with the design limit of 280 cal /gn.then a more detailed, plant-speci (3) C., paone, C. J., and llaun, J. H.,

Reactor Addendum No. 2 Exposed " RodCores " Supplement Drop Accident 2 Bolling Analysis of Large January 1975.

- NED0-10527Stir Water (4) NEDE-240ll-PaA, General Electric Standard Application for Reactor Fuel.

Anendment No. 22, AB, 61 g 374 3_3 i Q _ _ _ --

If-It is recognized that these bounds -are conservative with respect to expected operating conditions.

any one of the above conditions is not satisfied, a more detailed calculation will be done to.show com-pliance with the 280 cal /gm design limit. , ,

Should a control rod drop accident result in a peak fuel energy content of 280 cal /gr.;, less than 660 ~

(7 x 7) fuel rods are conservatively estimated to perforate. This would result in of fsite dosesFor twice j

J that previously reported in the FSAR, but still well below the guideline values of 10 CFR 100.

i 8 x 8 fuel, less than 850 rods are conservatively estimated to perforate, which has nearly the same consequences as for the 7 x 1 fuel case because of the operating rod power differences.

1 j

The BWM provides automatic supervision to assure titat out-of-squence control rods will not be References withdrawn Section or in-

'VII.10 of

serted
1.e , it limits operator deviations from planned withdrawal sequences.

the FSAR. It serves as an independent backup of the'Rormal withdrawal procedure followed by the opera-j In the event that the RWM is out of service when required, a second independent operator or In engineer tor. this can manually fulfill the operator follower control rod pattern conformance function of the RWM.

case, procedural control is exercised by verifying all. control rod positions af ter the withdrawal of each j group, prior to proceeding to the next' group. Allowing substitution of a second adequately monitorinoependent proper rod sequenc-operator or i

engineer in case of RWM inoperability recognizes the capability to Above 201 power, there is no ing in an alternate manner without unduly restricting plant operations.

requirement that the RWM be operable since the control rod drop accident with out-of-sequence ro 4

result in a peak fuel energy content of less than 280 cal /gm.

is required to be operating during a startup for the withdrawal of a significant number of control rods j for any startup.

1 i . ,

l j . .

i i

l l

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\

0 3/4 3-4 r

) Amendment Ho'. ZZ, A9, 61 I

i f

J Th3 ceram times' for all control roda will be determined at the. time of cach refueling outage. %e weekly control . .

rod exercisa test serves as a periodic check against deterioration of the~ control rod system and also verifies the ability of the control rod drive to scram since if a rod can be moved with drive pressure, it will scram .

because of higher pressure applied - during scram . The frequency of exercising the . control rods under the conditions of three or more control rods out of service provides even further assurance of ~ the reliability of the '

4 remaining control rods. .

The occurrence of scram times within the limits, but significantly longer than average, will be ~ viewed as a possible warning of systmatic problem with control rod drives especially if the ntsnber of drives exhibiting such i scram times exceeds eight, the allcwable nmber of inoperable rods. -

+

D. Control Rod Acctsnulators

%e specification for the ntsnber of accumulators which may be valved out of ' service is based on a series 'of two dimensional XY diffusion theory calculations at 2&C. These analyses prove that the reactor will be suberitical even when the central control rod of each 3 x 3 nine rod array is fully withdrawn.

{ E. Reactivity Anmalies During each fuel cycle, excess operating reactivity varies as fuel depletes and as any . burnable poison in supplementary control is burned. The magnitude of this excess reactivity may be inferred frm critical rod .

l configuration. As fuel bormp progresses, anmalous behavior in the excess reactivity may be detected by '

comparison of the critical rod pattern selected base states to the predicted rod inventory at that state. Power '

operating base cconditions provide the most sensitive and directly interpretable data relative to core reactivity. Furthernore, using power operating base conditionis permits frequent reactivity ccmparisons. Requiring a reactivity cmparison at the specified frequency assures that a cmparison will be made before the core i reactivity chapge exceeds 1% 4k are not expected and require thorough evaluation. One percent reactivity. limit is

] considered safe since an insertion of the-reactivity into the core would not lead to transients exceeding design ,

! conditions of the reactor system.

F. Power / Flow Operating Map Allowable cmbinations of thermal power and total core flow are restricted to Curve 1 of Figure 3.3.1. Analyses j show that reactor ascension to full power may proceed on a modified power / flow line bounded by the rod block line

] up to a point called the 100% intercept point (100% power /87% total core flow), frm which continued power

{ increases may proceed in a direct linear manner to the 100% power /100% flow point (5,6,7).

(5) " Millstone Point Nuclear Power Station - Unit 1 Ioad Analysis License Amendment Sutrnittal"NEDO-21285,

  • June,1976.

l (6) " Millstone Unit 1 - Ioad Line Limit Analysis" NEDO-21285-1, Novmber,1977.

(7) " Extended Ioad Line Limit Analysis -

Millstone Point Nuclear Power Station Unit 1" j ,NEDO-24366, September,1981.

9 B 3/4 3-6 -

_e

(j LIMITING CONDITION FOR OPERATION ' SURVEILLANCE REQUIREMENTS ) .

! However, there are various conditions under which the dissolved oxygen content of the reactor coolant water

! could.be higher than 0.2-0.3 ppe, such as refueling, hot standby'and reactor startup. During these periods with steaming rates less than 1 percent of full flow (80,000 pounds per hour), a more restrictive limit of 0.1 ppe han been established to assure the chloride-oxygen combinations are maintained at conservative levels., ,

At steaming rates of at least I percent of full flow (80,000 pounds per hour), boiling occurs causing demeration of the reactor water, thus, maintaining osygen concen'tration at low levels.

4 When conductivity la in its proper normal range, pH and J.aloride and other inspurities affetting cunductivity must also be within their normal range. Een conductivity becomes ibnormal then chloride measurements are made to determine whether or not they are also out of their normal operating values. This would not necessarily l

be the case. Conductivity could be high due to the presence of a neutral asit; e.g., Na 2SO4 , which would not i

have an effect on pH or chloride. In such a case, high conductivity alone is not a cause for shutdown. In l some, types of water-cooled reactors, conductivities are in fact high due to purposeful addition of additives.

In the case of BWRs', however, where no additives'are used and where neutral pH is maintained, conductivity j provides a very good measure of the quality of the reactor water. Significant changes therein provide the i operator with a warning mechanism so he can investigste and remedy the condition causing the change before -

limiting conditions, with respect to variables affecting the boundaries of the reactor coolant, are exceeded.

Hethods available to the operator for correcting the off-standard condition include operation of the reactor i cleanup system, reducing the input cf impurities and piscing the reactor in the cold shutdown condition.

} The major benefit of cold shutdown is to reduce the temperature dependent corrosion rates and provide time for i the cleanup system to reestablish the purity of the reactor coolant. During startup periods and hot standby, l

which are in the category of less than 11 of full flow (80,000 pounds per hour), conductivity may excee'l 2 sho/cm because of the initial evolution of gases and the initial addition of dissolved metals. During this Je period of time, when the conductivity exceeds 2 sho (other than short-term spikes), samples will be taken to i assure that the chloride concentration is less than 0.1 ppe.

The conductivity at the reactor coolant is continuously monitored. The samples of the coolant which are taken ,

l i every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> will serve as a reference for calibration of these monitors and is considered adequate to sasure j accurate readings of the monitors. If conductivity is within its normal range, chlorides and other impurities i will also be within their normal ranges. The reactor coolant samples will also be used to deterhine the l chlorides. Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content. Wile conductivity monitoring assures that pH is in the normal range, samples of reactor coolant are taken and tested for pH once a week as, a check. Isotopic analyses to determine major contributors to activity can be performed by a gamma scan.

f]

! D. Coolant Leaka n i

The 2.5 gpa limit for leaks from unidentified sources won established by assuming the leakage was from the primary

system. Tests demonstrate'that a relationship exists between the size of a crack and the probability that a crack j will propagate.

Amendment No. ,

. (

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) .

- A .- m ,,

- ( &. ,

3.7 Baies i A. 1. Primary Containment

limit the off-site doses to values less than those specified in 10 CFR 100 in the event of a break in the '

4 primary. system piping. Thus, containment jntegrity is specirled whenever the potential for violation of

f the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure. An exception is made to this requirement during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at this time which will greatly reduce the chances of a pipe break. The reactor may be taken critical during this period; however. restrictive operating proceduret i will be in effect again to minimize the probability of an accident occurring. Procedures and the Rod Worth l

Minimizer would limit control worth to 'less than 1.5% AK. A drop of a 1.5%AK rod does not result in any i fuel damage. In addition, in the unlikely event that an excursion did occur, the reactor building and '

i standby gas treatment system, which shall be operational during.this time, offer a sufficient barrier to 1 keep off-site doses well within 10 CFR 100 guideline values.

i

2. Suppression Chamber 1

! The pressure suppression pool water provides the heat sink for the reactor Primary system energy release i following a postulated rupture of the system or for releases through the safety relief valves. The Pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1035~psig.

Sin'te all of the gases in the drywell are considered purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting from isothennal compression plus the vapor pressure of the 11guld must not exceed 62 psig, the suppression chamber design pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the i

suppression chamber.

} Using the minimum or maximum water volumes given'in the specification, containment pressure during the design j Maximum water v 1

basisaccigentisapproximdely42psigwhichisbelowthedesignof62psig.

100,400 f t results in a downcomer submergenceresults of 3.33 feet and, t '

i in a submergence of 3.0 feet. The majority of the Bodega tests were run with a submerged length of four l feet and with complete condensation. Additional condensation tests were run in the Mark i full Scale ',

Test Facility (FSTF) at downcomer submergence varying between 1.5 and 4.5 feet and complete condensation

' of steam resulted. Thus, with respect to downcomer submergence, this specification is adequate.

i j The maintenance of a drywell-suppression chamber differential pressure of 1.00 psid and a suppression ,

j chamber water level corresponding to a downcomer suleergence range of 3.0 to 3.33 feet will assure the l

Post-LOCA suppression pool swell hydrodynamic forces are minimized and consistent with loads assumed '

! for structural analysis of the suppression chamber.

N -

Amendment No. ff, 73 -

  • B 3/4 7-1

b .

5. Oxyaen Concentration g

. e The relatively small containment volume inherent in the CE-BWR pressure suppression containment and the. ,

large amount of zirconium in the core'are such that the occurrence of a very limited (a percent of so) ,

reaction of the zirconium and steam during a Inss of coolant accident would lead to the liberation of sufficient hydrogen to result in a flammable concentration in the containment. Subsequent ignition of the hydrogen if it is present in sufficient quantities to result in excessively rapid recombination, could result in a loss of containment intergrity.

i The 4% oxygen concentration minimizes the possibility of hydrogen combustion following a loss of coolant accidelit. Significant quantities of hydrogen could be generated if the core cooling-systems did not sufficiently cool the core.

9 The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is more probable than the occurrence of the loss of coolant accident upon which the specified oxygen concentra-tion limit is based. . Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety without significantly reducing the margin of safety. Thus to preclude the possibility of starting the reactor and operating for extended periods of time with significant.

leaks in the primary system, leak inspections are scheduled during startup periods, when the primary system is at or near rated operating temperature and pressure. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period to provide inerting is judged to be sufficient to perform the leak inspection and establish the required oxygen concentration. The primary containment is normally slightly pressurized during periods of reactor operation assuring no air e

, in-leakage through the primary containment, llowever, at 1 east once a week, the oxygen concentration will

~

be detgrained as added assurance.

B. Standby Gas Treatment Systes The standby gas treatment system is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment isolation conditions. Both standby gas treatment system fans are designed to automatically start upon containment isolation and to maintain the reactor building pressure to the design negative pressure so that all leakage should be in-leakage. Each of the two fans has 100 percent capacity.

~

liigh efficiency particulate absolute (IIEPA) filters are installed before and after the ci:arcoal absorbers to minimize potential release of particulates to the environment and to prevent clogging of the iodine absorbers. The charcoal absorbers are installed to reduce the potential release of radioiodine to the environment. The in place test results should indicate a system leak Lightness of less than 1 percent bypass leakage for'the charcoal absorbers and a llEPA efficiency of at least 99 percent removal of DOP particulates. The labor'atory carbon sample test results should indicate a Amendment 101

.. 'B 3/4 7-5 .

a s

radiocctiva methyl icdida traoval efficiency of at leent 95 parcent for expacted accid:nt ce:ditiens.

If the efficiencies of the HEPA filters and charcoal absorbers are as specified, the resulting doses '

will be less than the 10 CFR 100 guidelines for the accidents analyzed. Operation of the fans significantly different from the design flow will change the removal efficiency of the HDPA filters and charcoal absorbers.

Only one of the two standby gas treatment systems is needed to clean up the reactor building ptmo-sphere upon containment isolation. If one system is found to be Inoperable, there is no immediate

, threat to the containment system performance and, reactor operation or refueling operation may continue while repairs are being made. During refueling two of f-site power sources. (345KV or 27KV) cad one emergency power source would provide an adequate and reliable source of power and allow completion of annual diesel or gas turbine preventative maintenance.

C. Secondary Containment The secondary containment is designed to minimize any ground level release of radioactive materials which might -

result from a serious accident. The reactor building provides secondary containment during reactor operation,

, when the drywell is sealed and in service; the reactor building provides primary containment when the reactor is

shutdown and the drywell is open, as during refueling. Because the secondary containme;t is an integral part of the complete containment system, secondary containment is required at all times that primary containment is required.

7 D. Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression

! system. Automatic initiation is required to minimize the potential leakage paths from the contain-I cent in the event of a loss of coolant accident.

1 1 .

l t

Amendr?nt No. 8 3/4 7-6 ,,s ~

\ '

1 i

5.0 DESIGN FEA'IURES .

S.1 Site The Unit I reactor building is located on the site at Millstone Point in Waterford, Connecticut.%e nearest site boundary on land is 1620 feet northeast of the reactor building, which is the minimm distance to the boundary of the exclusion area as described in 10 CFR 100.3(a). No part of the site which is closer to'the reactor building than 1620 feet shall be sold or leased except to (i) %e Connecticut Light and Power Cmpany, Western Massachudetts Electric Cmpany or Northeast Nuclear Energy Cmpany or their corporate affiliates for use in conjunction with normal utility operations and (ii) to the two leasees under the leases referred to in the following paragraph.

A United States Navy research Laboratory and a' desalination pilot operation of the Maximum Eva'porator Division of the Cuno Engineering Corporation may be permitted to operate within the exclusion area under leases which make activities and persons on the leased premises subject to health and safety requirements of the owner of the site.

5.2 Reactor A. W e core shall consist of 580 fuel assemblies.

B. The reactor core shall contain 145 cruciform-shaped control rods. We control material shall be hafnim and/or boron carbide powder (B4C) cmpacted to approximately 70% of theoretical density.

5.3 Reactor Vessel The react'or vessel shall be as described in Table IV-1 of the FSAR. The applicable design codes shall be as described in Table IV-1 of the FSAR.

5.4 Conatainment A. The principal design parameters and applicable design codes for the primary containment shall be as given in Table V-1 of the FSAR. .

B.  % e secondary containment shall be as described in Section V-3 of the FSAR and the applicable codes shall be as described in section XII of the FSAR.

d 5-1 l

1

r_ _ _ _ .

C. Penetrations to the primary containment and piping passing through such penetrations shall be designed 'in accordance with standards set fourth in Section V-2 of the FSAR.

5.5 Fuel Storage .

A. The new fuel starage facility shall be such that the K,gg dry is less than 0.90 and flooded is less than 0.95. g B. The K of the spent fu?l storage pool shall be less than or equal to 0.90.

eff 5.6 Seismic Design i The reactor building and all contained engineered safeguards are designed for maximun credible earthquake ground motion with an acceleration of 17% of gravity. Dynamic analysis was used to determine the earthquake acceleration applicable to the various elevations in the reactor building. ,

.~..,,.

5-2 O

- 4' , ,

Docket No. 50-245 B11663 4

Y Attachment No. 3 Millstone Unit No. 1 Technical Specification Typographical Error Corrections, Clarifications and Reference Updates 4

4 August,1985 i -

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1. T.S. 4.1.B-Reactor Protection System, Page 3/4 1-1 Change " peak heat flux" to " maximum fraction of limiting power density" to match present day terminology. Also, the words "to be valid" are added for clarification.
2. ' TABLE 3.2.3 - INSTRUMENTATION THAT INITIATES ROD BLOCK, Page 3/4 2-5 Rod Block Monitor (RBM) upscale equation was incorrectly changed in Table 3.2.3'in a prior submittal. This change corrects that error)
3. T.S. 3.3.B.1 - REACTIVITY CONTROL, CONTROL ROD WITHDRAWAL, Page 3/4 3-2 Removal of the word "not" from the phrase "as many as one drive in each quadrant may be uncoupled from its control rod so long'as the reactor is not in the shutdown or refuel condition" is required for clarification. Control rod drives are not removed'for maintenance purposes at power as required

-by the wording of the existing Specification. Dose, temperature, shutdown margin, and control rod drive housing support restraints dictate that rod drive removal only occur in the shutdown or refuel condition. Therefore,the removal of the word "not" is required.

4. T.S. 3.6.F - SAFETY AND RELIEF VALVES, Page 3/4 6-5
a. Change "F" to "E" in this section title to provide correct lettering sequence.

. b. Remove the word " Acoustic" from 4.6.E Surveillance Requirements. A previous Technical Specification Change removed the word from the Limiting Condition For Operation section on the same page.

5. T.S. 3.6.H.3 -- RECIRCULATION PUMP FLOW MISMATCH, Page 3/4 6-11.

The terms ' equalizer valves are closed" and " equalizer line is isolated" are equivalent. In either case the recirculation loops are isolated from each other.

6. T.S. 3.9.B.2, AUXtLIARY ELECTRICAL SYSTEM, Page 3/4 9-2.

Change the word "the" to "and". ~f'

7. T.S. 3.10.B.4 - REFUELING AND SPENT FUEL HANDLING, CORE MONITORING, Page 3/4 10-2. ,

This Technical Specification is being rewritten to allow flexibility in core monitoring during fuel loading. If 3 counts per second (cps) can be maintained on the in-core

e Source Range Monitors (SRMs) by loading irradiated fuel near them early in the fuel loading sequence, dunking chambers will not be required. The proposed wording change provides a this flexibility.

8. 3.1 Bases, Page B 3/4 1-3 Reference to Table 3.1.1 indicates the note should be NOTE 8 rather than NOTE 7 as stated.
9. 3.2 BASES r>
a. Page B 3/4 2-1. The existing FSAR reference is incorrect.

Also the Bases sre updated to reference the current LOCA analysis.

b. Page B 3/4 2-2a. This change provides a clarification of the sequence of events following MSIV closure for actuation of the Isolation Condenser on high reactor pressure.
c. Page B 3/4 2-3. The words "of 127 inches of water and 79 inches of water" are not trip settings of the Isolation Condenser and are being removed.
10. 4.2 BASES, Page B 3/4 2-5 The reference number is being changed from (1) to (2) to maintain consistency of numbering of references in the Bases.
11. 3.3.B BASES, CONTROL ROD WITHDRAWAL
a. Page B 3/4 3-3.'The Bases are updated to reflect current General Electric Rod Drop Accident methodology and References.
b. Page B 3/4 3-4. An incorrect section of the FSAR is referenced. The section should be Vll.10 rather than 7-9.
12. 3.3.F BASES, POWER / FLOW OPERATING MAP, Page B 3/4 3-6.

The Bases are updated to reflect current power / flow map and restrictions, including references to Load Line Limit -

Analysis and Extended Load Line Limit Analysis.

13. 3.6 and 4.6 BASES, COOLANT CHEMISTRY,_ Page B 3/4 6-3 Review of this paragraph indicates th'at the word " normal" should actually be " abnormal".

~

14. ~ 3.7 BASES ,
a. Page B 3/4 7-1. The section title is added.
b. Page B 3/4 7-5 and 7-6. The [itles and Sections are rearranged to provide consistency and order in this section of the Bases.

15.'5.0 DESIGN FEATURES

a. 5.2 REACTOR, Page 5-1. This section is updated to include the provision for control rods containing hafnium.
b. 5.5 FUEL STORAGE, Page 5-2. Section 5.5.B requires K e

of the Spent Fuel Pool to be less than or equal to 0.90.ff The criteria in ' Section 5.5.C that U235 loading be less than or equal to 15.2 - gm/cm was derived without taking credit for Gadolinia or reactivity depletion due to burnup. These overly conservative assun.ptions result in maximum allowable calculated bundle enrichments lower than that of General Electric production bundles currently being used in most BWR's. The removal of Section 5.5.C is justified as long as the condition specified in Section 5.5.B is satisfied at all times.

a e

er a

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b' l ' &

- Docket No. 50-245 l

B11663 t

f.

  • 4 9

Attachment No.2

'" Supplemental Licensing Submittal for Millstone' Unit No. 1 Reload 10" 23A4696, dated August 1985 1

P i ,

August,1985 l<

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