Text

Amend 107 to License DPR-21,responding to 850826 Application to Refuel,Correcting Typos & Providing Clarification on Indicated Tech Spec Pages
	ML20138G082
Person / Time
Site:	Millstone
Issue date:	12/06/1985
From:	Charemagne Grimes; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20138G064	List: ML20138G060; ML20138G082; ML20138G100;
References
	NUDOCS 8512160231
	Download: ML20138G082 (41)
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/ o g UNITED STATES n- NUCLEAR REGULATORY COMMISSION g E WASHING TON, D. C. 20555

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THE CONNECTICUT LIGHT AND POWER COMPANY WESTERN MASSACHUSETTS ELECTRIC COMPANY AND NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION, UNIT N0. 1 DOCKET NO. 50-245 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 107 License No. DPR-21

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The application for amendment by the Connecticut Light and Power Company, Western Massachusetts Electric Ccmpany, and Northeast Nuclear Energy Company, (the licensees) dated August 26, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;

. B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public; and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E. The issuarce of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

A:

8512160231 851206 PDR .4 DOCK 05000245 P PDR

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Provisional Operating License No. DPR-21 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appei. dices A and B as revised through Amendment No.107, are hereby incorporated in the license. The Northeast Nuclear Energy Company shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION p -

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Christopher I. G lmes, Director UM33h Integrated Safeti Assessment Project Direct. rate Division of PWR Licensing-B

Attachment:

Changes to the Technical Specifications Date of Issuance: December 6,1985

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. 2-l A ATTACHMENT TO LICENSE AMENDMENT NO. 107

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PROVISIONAL OPERATING LICENSE NO. DPR-21 DOCKET NO. 50-245 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages (*) are provided to maintain document completeness.

REMOVE INSERT 3/4 1-1 3/4 1-1 3/4 2-5 3/4 2-5 3/4 3-2 3/4 3-2 3/4 6-5 3/4 6-5 3/4 6-11 3/4 6-11 3/4 9-1 3/4 9-1*

3/4 9-2 3/4 9-2 3/4 10-1 3/4 10-1*

3/4 10-2 3/4 10-2

' 3/4 11-1 3/4 11-1*

3/4 11-2 3/4 11-2 3/4 11-3 3/4 11-3 3/4 11-4 3/4 11-4 3/4 11-5 3/4 11-5 3/4 11-6 3/4 11-6 3/4 11-9 3/4 11-9*

3/4 11-10 3/4 11-10 B3/4 1-3 B3/4 1-3 B3/4 2-1 B3/4 2-1 B3/4 2-2 B3/4 2-2*

B3/4 2-2a B3/4 2-2a B3/4 2-3 B3/4 2-3 B3/4 2-4 83/4 2-4*

B3/4 3-1 B3/4 3-1*

B3/4 3-2 B3/4 3-2 B3/4 3-3 B3/4 3-3 B3/4 3-4 B3/4 3-4 B3/4 3-5 83/4 3-5*

B3/4 3-6 B3/4 3-6 83/4 6-3 B3/4 6-3 B3/4 6-4 B3/4 6-4*

B3/4 7-1 83/4 7-1 B3/4 7-2 B3/4 7-2*

83/4 7-5 83/4 7-5 e' B3/4 7-6 83/4 7-6 5-1 5-1 5-2 5-2 l

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LIMITING OWDITIm EOt OPERATIN SURVEIIIRCE IEUIIMOTT 3.1 REEIUt FwnUnCN SYSM1 4.1 IFElut IRmfrICN SYSIDi Applicability: 4plicability:

Applies to the imtnmentation arti associated devices Applies to de surveillan which initiate a rmetor scran arx3 provide autoratic of tJe instumentation aM ,

isolaticn of the Ibactor Protection Systen buses frun associated devices which initiate reactor scran and provide autarctic isolatim of reactor protecticn systen buses frtm Opirpowersumlies. deir power supplies.

Cbjective: (bjective:

To assure the operability of the Reactor Protection Syston.

To specify the type and frequency of surveillance to be

amlied to the ruactor protecticn imtrwentation.

Grcificaticn:

A. 'Ihe setpoints, mininun ruter of trip Specification:

systois, and mininun ruter of instnment channels that nust be cperable for each positim of the reactor node A. Instnmentation syston shall be functimalJy tested and switch shall be as given in Table 3.1.1.

calibrated as irdicated in Tables 4.1.1 ard 4.1.2, respectively.

B. Ibspmse Tine .

The time fran initiation of any channel trip to the B. . Ibily chrity reactor power cperaticn, the 'nuxinun

, de-energization of the scran solenoid relay sin 11 not fraction of limitirg pomr demity sMll be checkal and exceed 50 milliseccnis.

the APRM scram and rod block settirgs given by the C. Ibactcr Protecticn Syston Ibwer PtnitorirU equations in Specificaticns 2.1.2A ard 2.1.2B shall be determined to be valid, wo IPS electric power monitorirg channels for each inservim RPS PG set or altermte power stgly shall be C. The RPS electrical protection assorblies shall te cperable at all times except as follows: determined cperable as follows:

1. With cne RPS electric power rrrnitorirg channel for an 1.

imervice RPS PG set or altemate power sigly

At least once per 6 months by performance of a QWNEL MCTIGML 'IEr, and inoperable, restore the irnwrable channel to GYRAntE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or turrwe the associatcd RPS PG 2.

set or altermte power sigly frun servim. At least cnm per 18 nmths by dcamstratiro de OPERABILITY of over -voltage, urdermitage anl taler -frugtency protective imtnstentaticn by

2. With both RPS electric power mtnitoring channels for an irserviw RPS PG set or altermte power stp ly performnm of a OWNFL CALIBRATIm incitdity inoperable, restore at least one to OPERABIE status sinulated autonatic actuaticn of the protective within 30 mirutes or rurzwe de associated RPS PG set or relays, trimirg logic ard output cirmit breakers, altermte power stpply frun servi ard verifying tie follcwirg setpoints:

a. Over-voltage < (132)VAC,

b. thier -voltage > (108) VAC,

c. thder -frequency > (57)liz, and Amendment No. 78, 98,107 3/4 1-1

_ TABLE 3.2.3 Ill51RUHENTAil0N TilAT INITIAi[5 R0D DLOCK Hinimum Number of Operable instrument Channels pe ~

Trip System ( ) Instrument Trlp l.evel Setting I III APRM Upscale (Flow Blased) See Spec ifica t ion 2.1.2ll l 1 APRH Dnwnscale

. 1 3/125 full Sc' ale

,I (6) '

Rod Block Honitor Upscale (flow Blased) i .65U + 42 (2)

) (6) Rod Block Honitor Downscate 1 3/125 full Scale 3 IRH Do'wnscale (3) 13/125 full Scale 3 IRH Upscale i 100/125 rull Scale 2 SRH Detector not in Startup Position (4) 2 (5) SRM Upscale 1 105 counts /s ec .

I Scram Discharge Volume - Water level liigh I4 Is " $

I b e Imeer cap to 1

Scram Dischirge Volume - Scram Trip Bypassed N/A (1) For the Startup/llot Standby and Run posillons 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 APRH downscale need not be operable in th'e Startup/Ilot 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 thst

time the operable system is functionally tested immediately and daily thereaf ter; if this condition lasts longer than seven days, the system shall be tripped. If the first column cannot be met for both trip systems, the, systems shall be tripped.

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, (2) W ls the recirculation flow required to achieve rated core flow expressed in percent.

(3) IRN downscale may be bypassed when it is on Ils lowest range.

(4) This function may be bypassed when the count rate is ,100 cps or when all IRH range switches are above Posillon 2.

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

, upscale rod block is operable.

U , 107 3/4 2-5 AmendmentNo./ .

SibivklllAllCER[QUIRIH(Hi LlHlilHG COH0lil0N FOR OPERAtt0N

tantrol rods or in the event puwee upena-p wer operation. If a pas ti.sily or fully

tion is continuing.with onc fully or withdrawn control rod drive cannot be partially withdrawn rod which cannot be moved with drive or scram pressure the reactor shall be brought to a shutdown moved and for which control rod drive .

mechanism damage has not been ruled nut.

condition within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months unless investiga- lhe surveillance need not be completed tion demonstrates that the cause of the within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the number of inoperable f allu,e is not due to a failed control rod drive mechanism collet housing. rods has been reduced to les's than th!ee and if it has been demonstrated that

' control rod drive mechanism collet housing

8. Control Rod liithdrawal f ailure is not the cause of an insnovable

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

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

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

first time subsequent to each refuel condition and Specification 3.3. A.I refueling outage or af ter maintenance.

is met. observe that the drive does not go The control rod drive housing support to the overtravel position; and 2.

' system shall be in place during power when the rod is withdrawn the first operation and when the reactor coolant . b.

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time subsequent to each refuelino system is pressurfred above atmospheric outage or af ter maintenance, observe pressure with fuel in the reactor vessel, discernible response of the nuclear e unless all control rods are fully inserted Instrumentation

however, for initial

' and Specification 3.3.A.) is met. -

rods when response is not discernible, subsequent exercising of these rods after the reactor is critical shall

be performed to verify instrumenta-Lion response.

2. the control rod drive housing support system shall be inspected af ter s easunbly and the results of the inspectinn shall be.rerosded.

' 3/4 3-2 N ndmeni No..A,' 22'. /p,107 6

[.,i LIMITING CONDITION FOR OPERATION .- SURVEILLANCE REQUIREHENTS D. Coolant Leakane II . Coolant I.eakage Any time irradiated fuel is in the reactor Reactor coolant system leakage into the vessel, reactor coolant leakage into the primary containment from unidentified sources prisaary containment shall be checked and shall not exceed 2.5 gpm. In addition, the recorded at least once per day.

total reactor coolant system leakage into th E. Safety and Relief Valves primary containment shall not exceed 25 gpm.

If these conditions cannot be met, or if leak

1. Three of the relief / safety valves top rate cannot be determined, initiate an orderly shutdown and have the reactor in the cold Ms sid h I d ddd n WM shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , wi u a idddW mb M m gg ,g gg g ,

E. Safety and Relief Valves shall be checked or replaced every two refueling outages. The set pressure shall be adjusted to correspond with a

1. During power operation or whenever the reactor steam set pressure of; 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 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 valves shall 4 1125 i 1%

be operable as required by Specification 3.5.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 coolant pressure 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 instrumente-operable. Two of the six channels may be tion, or by the Valve Position f

,out of service provided backup indication Indication System, or by an audilile 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 distbarge.

AmendmentNo.[,[,g,107 3/4 6-5

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i LIMITING CONDITION FDR OPERATION SilHVElllANCE REQUIRENENT l 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the recir-culation pump is sooner made operable. If the pump cannot be made operable, the reactor shall be in cold shutdown withlp 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .,

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! 3. The reactor shall not be operated unless

4 4 the equalizer line is isolated.

1 4. With the mode switch in the startup/ hot ,

standby or run mode, operation without forced circulation shall not be ,

, permit ted.

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l AmenenInt No. II. ff,107 3/4611 1

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 AUXILIARY ELECTRICAL SYSTEM Apolicability: Appl ica bil i tyg, Applies to the auxiliary electrical power system. Applies to the periodic testing requirements of the auxiliary electrical system.

Objective:

Objec tive:

To cssure an adequate supply of electrical power during plant operation. Verify the operability of the auxiliary electrical system.'

Specification:

A. The reactor shall not be made critical unless all of the following conditions are satisfied: A. Emergency Power Sources

1. One 345 kv line, associated switchgear, and 1. Diesel Generator auxiliary startup transformer capable of automatically supplying auxiliary power. a. The diesel :,cnerator shall be started and loaded once a month to demonstrate

2. Both emergency power sources are operable. operational readiness. The test shall continue until the diesel engine and

3. An additional source of power consisting of the generator are at equilibrium one of the following: temperature at full load output. Dur-ing this test, the diesel starting

a. The 27.6 kv line, associated switch- air compressor will be checked for gear, shutdown transformer to supply operation and its ability to recharge power to the emergency 4160 volt buses. air receivers.

b. One 345 kv line fully operational and b. During each refueling outage, the capable of carrying auxiliary power conditions under which the diesel to the emergency buses, generator is required will be simulated and test conducted to denonstrate that

4. a. 4160 voit buses five and six are it will start and be ready to accept energized and the associated 480 load within 13 seconds.

volt buses are energized.

3/49-1

LIMITING CONDITION FOR OPtRATION SUHVI ill ANCE REQtilk!HI NT

5. All station and switchyard 24 and 125 volt c. During the monthly generatos sesi. .

batteries and associated battery chargers the diesel fuel oil transfer pumps i are operable. ; hall be operated, i *

8. When the mode switch is in Run, the availability 2. Gas Turbine Generator of power shall be as specified in 3.9.A, except as specified below: 4. The gas turbine generator shall be fast started and the output brealers

1. From and after the date that incoming power closed within 48 seconds once a inonth l is available frodi only one 345 kv line, to demonstra te opera tional readine .s.

reactor pperation is persuissible only J The test shall continue until the during the succeeding seven days unless an gas turbine and generator are at 3

additional 345 kv line is sooner placed in equilibrium temperature at full load service. output. Use of this unit to supply i power to the system electrical net-i 2. From: and af ter the date that incowing power work shall constitute an acceptable is not available from any 345 kw line, demonstration of operability.

reactor operation shall be pennitted pro-

. vided both emergency power sources are b. During each refueling outage, the

! operating and the isolation condenser system conditions under which the gas turbine-

} is operable. The NRC shall be notified, generator is required will be simulated 3

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the precautioev. to be and a test conducted to verify that taken during this situation and the plans it will start and be able to accept j for restoration of incoming power. The emergency loads within 48 seconds.

, mi nimur.: fuel supply for the gas turbine

during this situation shall be maintained 8. Itatteries 4

above 20,000 gallons, i 1. Station Batteries

3. From and af ter the date that either emer-gency gnawer source or its associated bus is a. Every week the specific gravity and

made or found to be inoperahic for any voltage of the pilot cell and tem-

reason, reactor operation is permissible

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perature of adjacent cells and overall

! act.ording to Specification 3.5.F/4.SI' unless battery voltage sina11 be measured.

i such emergency power source and its hus ..

j are .ooner asade operalil. , provided that b. Every three months the measurements 4 during such time two of f site lines (345 shall be made of voltage of earli j or 27.6 Lv) arc operat.le. cell to nearest 0.01 volt, spet ilic j gravity of each cell and tmperature j of every fif th ceII.

j Amendment No. 3I, 76,107 3f4 9 2

I LIMITING CONDITION FOR OPERAil04 '

5HRVtittANCE REQUIREMENT 3.10 REFUELING AND $ PENT FUtt MAM0 LING Applicability:

i' Applies to feel handling. core reactivity limitattens. Appiles to the perledic testing of these later-and spent feel haneling. 1schs and lastruments used during refueling and spent feel handling.

j Objective:

1 O6jective:

i To assere core reactivity is within capability of the control reds, to prevent criticality during refueling. Te verify the operability of lastrumentatten and and to assere safe hand 11ag et spent feel casts. Interlocks used in refueling and spent feel handling.

Specificatlen:

Spec t ficatlent i A. Refueline Interfecks_

1 A. Refueline Inter 1ects During core alterations. the reactor mode switch shall be locked in the REFUEt, Prlw te any fue) Mfg. with tk had, , ,

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position and the refueling interlocks off t Nacter essel, ik mfullag later-shall be operable, or all control rods locks shall be functlenally tested. They i

shall be fully inserted, valved out shall also be tested at weekly intervals and electrically disarmed.

I81IMIM any Mpair M aneclatd.mith the laterlocks.

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3 j If the refueling interlocks are not operable.

( all control rods shall be verified to be

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fully inserted. valved out and electrically disarmed in accordance with specification 4.1.C.

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StfRVEll.i.ANCE REQUIREffENT LlHITING CORDITION FOR OPERATION' D. Core Honitoring B. Core tionitoring ,

Prior to making any alterations to the fore, During core alterations two SRit's shall be oper- l.

the SHil'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 made.

graphs 3 and 4 below. For an SRH to be considered operable, the following conditions shall be 2. Prior to spiral unloading or reloading, the satisfied: SHH's shall be functionally tested. Prior

1. The SRH shall be inserted to the normal to spiral unloading, the SRH's should also operating level. (Use of special movable be checked for neutron response.

i dunking type detectors during fuel

. loading or major core alterations in place of normal detectors are permissible as long as the detector is connected into the normal SRM circuit.)

2. The SRM shall have a minimum neutron induced count rate of three per second with all rods fully inserted in the core.

j 3. Prior to unloading, the SRH's shall be proven operable as stated above, however, during spiral unloading', the count rate

. may drop be. low 3 cys, i

i 44 If required, special movable dunking type detectors can be inserted into the core, prior to reloading fuel assemblies into the central core region (with all' control '

rods inserted). Before the ninth fuel i assembly is loaded into the core in the close proximity of the movable dunking chambers or the SRM s Paragraph 3.10.B.1 and 2 apply.

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3/4 10-2 Amendment No. 95, 107 .

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. SMITING CONDITION FOR OPET,i. TION . ..

! SURVEILLANCE REQUIREMENT , 3.11 REACTOR FUEL ASSDBLY 4.11 REACTOR FUEL ASSEMBLY ,

Applicability Applicability The Limiting Conditions for Operation associated with the Surveillance Requirements apply t,e the para-the fuel rods apply to those parameters which monitor meters which monitor the fuel rod operating the fuel rod operating conditfor, conditions.

Objective objective The Objective of the Limiting Conditions for Opera- The Objective of Surveillance Requirements is tion is to assure the perfonnance of the fuel rods, to specify the type and frequency of survelliance to be applied to the fuel rods.

Specifications Specifications A. Average Planar Linear Heat Generation Rate (APLHGR) A. Average Planar Linear Heat Generation Rate (APLHGR)

1. During power operation, the APLHGR for each type of fuel as a function of average The APLHGR for each type of fuel as a planar exposure shall not exceed the limit-

function of average planar exposure shall ing value shown in Figure 3.11.1. be detennined daily during reactor operation at > 251 rated thermal power.

2. If at any time during operation it is detennined by nonnal surveillance that the limiting value for APLHGR specified in Section 3.ll.A.1 is being exceeded, action shall be initiated within 15 mimates to restore operation to within the prescribed .

limits. If the APLilGR is not returned to l within the prescribed limits within two (2) i hours, the reactor shall be brought to the

Cold Shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

! Surveillance and corresponding action shall continue untti reactor operation is within the prescribed limits.

Amendment No. 73, 76 3/4 II-I

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1 0 5000 10.000 15,000 '26,00,0 25,000 30,000 35,000 40,000 PLANAR AVERACl! EXPOSURE mud /t Figure 3.11.la - MAXIMUM AVERACE PLANAR LINEAR llEAT CENERATION RATE (MAPLi!GR)

VERSUS PLANAR AVERACE EXPOSURE. FUEL TYPE P8DRB282 .

Amendment No. 73, 87, 98, 107 3/4 11-2

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VERSUS PIANAR AVERAGE EXPOSURE. FUEL TYPE BP8DRB300 Amendment No. 98,107 3/4 11-4

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LIMlilllG CONDITI0ll FOR OPERAftoll SilRVElllANCE R[l)UIREM(NT C. Minless TrItIcel Power Astle (MCHtJ, C. to"slede Crit leal Power Rat le (P' Crit)

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I nuring power operalten.*MCPR shall be as shown in 1. NCrit shall be determined daily during Table 3.11.1 If at any time during operation reactor power operation at > 251 rated

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It is determined by poe1 mal surveillance that the flieswal poieer and following any change in lletting valise far MCPR is being exceede.l. action power level or distrileotten that would cause 4 .

shall s.e enlilated within 15 salmstes to restare operation with a limiting control rod pattern operation to within the prescribed limle s. If as descrlhed in the bases for spet.lfication the steady state MCPO41 eiet returned to within '3.3.8.5.

the prescrlhed limits within two (?) hours. the reacter shall be brenght to the l'enld Shutdown 2. Utilization of Option B Operating limit

~

i MCPR values requires the scram time testing conditlen within 36 hosers. Surveillane.e and corresponding actlen shall continese natil reactor, of 15 or more control rods on a rotating l . operation is within the prescribed lletts, basis every 120 operating days.

For core flows other then rated the MCPIt's in l

Table 3.11.1 shall be multiplied by E g. where j Kg is as shown la flyere 3.11.2.

D. If any of the lletting values identified in Specifications 3.II.A. O, or C. are enceeded, ,

! even if corrective actlen is taken, as pre.

scribed,a Reportable Octorrence reiport shall l be solmitted.

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k ndoent No. J. JS, /7,0 7

J/4 11 9 I.

1

TABLE 3.11.1 OPERATING LIMIT MCPR'S FOR CYCLE ll (OPTION B)

BOC 11 TO EOC EOC 11 TO 70% COASTDOWN FUEL TYPE l

1.42 1.42 P8 x 8R 1.42 1.42 BP8 x 8h I

OPERATING LIMIT MCPR'S FOR CYCLE ll (OPTION A)

BOC 11 TO EOC EOC 11 TO 70% COASTDOWN FUEL TYPE 1,48 1.48 P8 x 8R 1.48 1.48 BP8 x 8R Amendment No. 98. 107 3/4 11-10

.n

Nigh radiation is an indication of leaking fuel.

background.

levelsis in the main steamline tunnel above that due to the normal nitrog A scram The purpose of the scram is to redureinit iat ed whenever sus h radiation level exceeds seven tien and oxygen mes normal excessive release of radicartive materials.

is prevented Discharge which caus of excessive amounts of radioactivity tothe sou provided the by limittheforairaejector off gas monitors the site envilous .

15-minute period specified in specification 3.8 is note an isolat ion of the main condenser of f gas exceeded.**

i ' The main steamline isolation valve closure, scram is set to full open in three out of four lines. This scram anti. D c when the valves close. By scramming at a= when the and

-ressure isolation flux valves are 101 closed from l Section II.3.7 FSAR. this setting the transient, which would occur 4 ecsuis-at transient is insignificant. Ref.

I A reactor plant particular modeoperating switch is status.

provided which actuates or bypasses the vari Ref. Section 7.2 FSAR. ous scram functions appropriate to the i The manual scram function is active in all modes, rods during all modes of reactor operation. thus providing for a manual means of rapidly inserting control ,

}

j The IRN and APRH system provide protection against and Startup/Not Standby modes. excessive power levels and short

reactor periods in the refuel level information during startup but has no scram functionsA source range monitor is also provided to supply(SRH) system additional neutron and Startup/Not Standby modes, .

is not required in the Run mode. in the power range the APRM provides the required protThus the IRN an ections; thus, the IRH system i The hiah reactor pressure, high drywell pressure,

} be operational for e th'se modes of reactor operation. scrams n. Theyareare, required therefore,for Startup/

required to Hot Sta The requirement to have all scram functions except those listed in Note 8

and Shutdown mode is to assure that shifting to the Refuel mode durin of Table 3.1.1 operable' in the Refuel i j the need for the reactor protection system. g reactor power operation does not dioinish As indicated l'n Note j ' assures maximum negative reactivity insertionto be operable .

if all electrically controlsince disarmed, rodsthis arecondition fully inserted, v I

9 1 ** Per errata sheet dated 10-7-70 I

1 j Amendment No. pp, 107 i 53/4 l-3

'3.2 Basest In addition to reactor protection instrunentation which initiates a reactor scram, protective instrunentation has been provided which initiates action to mitagate the consequences of accidents which are beyond the operator's cbility to control or terminates operator errors before they result in serious consequences. %is set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the emergency core cooling system, control rod block and standby gas treatment systems. %e objectives of the specifications are to assure the effectiveness of the protective instrumentation when required by its capability to tolerate a signal failure of any caponent of such systems even during periods when portions of such systems cro 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.

Isolation valves are installed in those lines that penetrate the' primary contairinent 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 instrumentation shown in table 3.2.1 which senses the conditions for which isolation is required. Such instrunentaion must be available whenever primary contairunent integrity is required.%e objective is to isolate the primary containment so that the guideline values of 10 CFR l

100 are not exceeded during an accident.

Tha instranentation which initiates primary systen isolation is connected in a dual bus arrangement. Mus, the discussion given in the bases for Specification 3.1 is applicable here.

We low reactor water level instrunentation is set to trip when reactor water level is 127 inches above the top of i'

the active fuel. %is trip initiates closure of Group 2 and 3 primary contairunent isolation valves but does not trip the recirculation punps. Ref. Section VII-4.4 FSAR. For a trip setting of 127 inches above the top of the

cctive fuel and a 60-second valve closure time the valve will be closed before core uncovery occurs even for the maximum break in the line; and therefore, the setting is adequate.

%e low low reactor water level instrunentation is set to trip when reactor water level is 79 inches above the top i of the active fuel. Wis trip initiates closure of Group 1 primary contairunent isolation valves, Ref. Section VII-4.4 FSAR and also activates the ECC subsystens and starts the emergency diesel generator and the gas turbine I generator and trips the recirculation purps. Wis trip setting level was chosen to be high enough to prevent cpurious operation but low enough to initiate ECCS operation and primary system isolation so that post accident cooling can be effectively accmplished and the guideline values of 10 CFR 100 will not be violated. For the complete circunferential break of a 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary systen isolation are initiated in time to meet the above criteria. We instrunentation also l covsrs the full range or spectrun of breaks and meets the above criteria. Ref.(1). We Isolation Condenser system I hrs been added to the ECC system to insure that cladding integrity is maintained for postulated small break IOCA conditions in the recire, discharge piping with a gas turbine failure and LPCI injection into the damaged loop.

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

l Amendment No. 67,107 B 3/4 2-1 4

}

l l

j i lhe high drywell pressure instsumentation is a back-up to the wal o len l in.trumenta tion and in addi tems to j initiating LCCS it causes isolatlun of Group 2 isolation valwe'.. Our the breaks discussed aluve, this instrumenta-1 tion will initiate ECCS operation at about the same time as the low low water level instrumentation; thus t he resu l t s j given above are applicable here also. Group 2 isolation value:s include the drywell vent, purge and sump isol.iiii.n valves, and reactor building ventilation isolation valves , t.roup 2 as.tuation also initiates the SBGIS, lityh de fwell

pressure activates only these valves because high drywel pressure could occur as the result of non-safety relas.d

{

causes such as not purging the drywell air during startu . Total system isolation is not desirable for these cor..le-i tions and only the valves in Group 2 are required to close. The low low water level instrumentation initiates j protection for the full spectrum of loss of coolant accidents and causes a trip of all primary system-isolation volves.

! Venturfs are provided in the main steamlines as a means of measuring steam flow and also limiting the loss of mats i .

Inventory from the vessel during a steamline break accident. In addition to monitoring the steam flow. Instrumenta-tion is provided which causes a trip of Group 1 1 solation valves. The primary function of the instrumentation I:. to detect a break in the main steamilne, thus Group 1 valves are closed. For the worst case accident, main steamline break outside the drywell, this trip setting of 1201 of rated steam flow in conjunction with the flow limiters and 1

main steamilne closure. Ilmf t the mass inventory loss such that fuel is not uncovered, fuel temperatures remain less i

than 1500*F and release of radioactivity to the environs is well below to CfR 100 guideline values. The main steam-j line high flow break detection is a one out of two twice logic for each individual steamline, four detectors per i

line for a total of 16 detectors. lAien a steamline is isolated by closing both main steam isolation valves the 4

operable f astrument channels per trip system requirements are not required to be met because the protection afforded by the renalning operable logic in the in-service steamitnes provides complete recognition of the steam flow

measurements required for correct protective action.

?;

i Temperature monitoring Instrumentation is provided in the main steamline tunnel to detect leaks in this area. Trips are provided on this lastrumentation and when exceeded cause closure of Group 1 1 solation valves. Its setting of l .

200*F is low enough to detect leaks of the order of 5 to 10 gpm; thus, it is capable of covering the entire spectrum j

l of breaks. For large breaks. It is back-up to high steam flow instrumentation discussed above, and for small breaks i

with the resultant small release of radioactivity, gives isolation before the guidelines of 10 CFR 100 are exceeded.

! lifgh radiation monitors in<the main steam 11ae tunnel have been provided to detect gross fuel failure. This instru-

{ mentation causes closure of Group 1 valves, the only valves required to close to prevent further release to the '

environment. Iff th the established setting of seven times novinal background, and main steamline isolation valve closure, fission product release is Ilmited so that 10 CFR 100 guideline values are not exceeded for the most rapid failure mechanism postulated (control rod drop accident).

Pressure instrumentation ~ provided which trips when main steamline pressure at the turbine drops below '825 Psf9. A trip of this Instrumentation results in closure of Group 1 isolation valves. In the " Refuel." " Shutdown." and "Startup/Ilot Standby" mode this trip function is bypassed. This function is provided primarily to provide protection against a pressure regulator malfunction which would cause the control and/or bypass valves to open. liith the trip set at 825 psig inventory loss is Ilmited so that fuel is not uncovered and peak clad temperatures are much less than 1504*r; thus, there is no release of fission products other than those in the reactor water.

Amendment 7,'76, 87 s 3/4 2-2 j

i .

l

High pressure actuation of the Isolation Condenser (IC) util be a backup to direct activation on Low-Low level; stattar to other ECCS systems. Activation is based on the bly.h pressure signal (1085 PSic for 15 seconds) which occurs a,fter MSiv closure on Low-Low water level, SRV actuation, and subseqtgent repressurization.

tion of the IC requires only the opening of normally closed valve IC-3 in the condensate return line. The activa- l This valve is powered by the safety-grade DC battery.

grade AC or DC power and are also used for containmentAll valves inAlltlie isolatinh. system are areInpowered normally the openbyposil safetylon" (other than 10-3). The IC syston is safety Class 2 and is scismically quallfled. The shell side water volume is suf ficient 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.

i 1

2 1 .

Amendment No. 67, 107 8 3/4 2-2a I

l

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 return sides are arranged in a 1 out of 2 logic and to meet the single f ailure cr_ teria, 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 instrumentatfre.which initiates ECCS action is arranged in a dual bus system. As for other vital inst rumen.ta-tion arranged in V.es+ feehton, the Speelfication preserves the effectiveness of the system even during periods when maintenance or testing is being performed.

The control ro'd ulock functions are provided to prevent excessise control rod withdrawal so that HCPR does not decrease to < l.07 The trip logic for this function is I out of a; e.g., any trip on one of the six APRM's, eight IRH's, or four SRM's will result in a rod block. The minimum instrument channel requirements assure suf ficient instrumentation to aneure the single f ailure criteria is met. The minimum instrument channel requirements for the IRM end RBM may be reduced by 6ne for a short period of time to allow for maintenance testing and calibration.

l l The AFRM rod block trip to flow blamed and prevents significant approach to HCPR-l.07 especially during operation at reduced flow. The APRH provideo groes core protection, i.e., limits the groes 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 RBN 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. The trip point is flow biased. 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 specified trip settings, rod withdrawal is blocked within an adequate margin to fuel damage limits. This margin varies slightly from reload to reload and, thus, each reload submittal contains an update of the analysis. Below % 70%

power, the withdravel of single control rod results in HCPR > I.07 without rod block action, thus requiring the RBH system t"o be operable above 30; of rated power is conservative. Requiring at least half of the normal LPRM

' inputs f r'on each level to be operable assures that the RBH response will be adequate to prevent rod withdrawal errore.

The IRM rod block functions assure proper upranging of the IRN system, and reduce the probability of spurious scrans i

during startup operations.

A downeeste indication on en AFRH or IRM is an indication the instrument 1.se f ailed or the inst rument is not sensitive enough or the neutron flus is below the Instrument response thre. hold. In these cases the i ns t rumen t will not respond to changes in control rod motion and thus control rud motion is prevented. The downscale trips

are set at 3/125 of full scale.

l

! e Ameadment No. )6. II. J 07 8 3/4 2-3 -

i

1 1

5 To prevent excessive fuel clad temperature for the small pipe break, the FWCl or Isolation Condenser systems must function since for these breaks reactor pressure does not decrease rapidly enough to allow either core

, spray or LPCI to operate in time. The automatic pressure relief function and Isolation Condenser system are i

provided as back-ups to the FWCI in the event the FWCI does not operate. The arrangements of the tripping contacts are such as to provide these functions when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criterion is met. Ref. Section VI-2.0 FSAR. The specification preserves the effectiveness of the system during periods of maintenance, testin i calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel ou't g, of or

service.

Two air ejector off-gas monitors are provided and when their trip point is reached, cause an isolation of the

air ejector off-gas line. Isolation is initiated when both instruments reach their high trip point or one has j an upscale trip and the other a downscale trip. There is a 15-minute delay before the air ejector off-gas isolation valve is closed. This delay is accounted for by the 30-minute holdup time of the off-gas before it is released to the stack.

Both instruments are required for trip but the instruments are so designed that any instrument failure gives a l downscale trip. The trip' settings of the instruments are set so that any instrument failure gives a downscale

trip. The trip settings of the instruments are set so that the instantaneous stack release rate limit given in i Specification 3.8 is not exceeded.

i Three sets of two radiation monitors are provided which initiate isolation of the reactor building and steam

! I tunnel ventilation and operation of the standby gas treatment system. One set of monitors is located in the reactor building ventilation exhaust duct, one set is located in the vicinity of the fuel pool and the other i

set is located adjacent to the steasi tunnel ventilation exhaust duct. A high level trip on any one of the six

{ l monitors or two isolate the downscale ventilation trips on system. Tripany one set settings of of 100monitors mr/hr on willthe initiate the standby fuel pool monitor,11gas mr/hr treatment on the system and *

, ventilation duct monitor, and 12 mr/hr on the steam tunnel ventilation monitor are based on initiating normal ventilation isolation and standby gas t'reatment system operation.

Amendment No. 57,100 8 3/4 2-4 i

1 . .

3.3 Bases j A. Reactivity 1. imitations 4 '

The core reactivity limitation is a restriction to be applied principally to the design of new fuel which may be loaded in the core or into a particular refueling pattern. Satisfaction of the limitation can only he demonstrated at the time of loading and must !w> such that it will apply to the entire subsequent

' fuel cycle. The reactivity of the loaded core will be limited so the core can be made subcritical by .

~

at least R + 0.330 AK' 'in the most reactive condition during the operating cycle, with the strongest control rod fully withdrawn and all others fully inserted. The value of R in %AK is the amount by which the core reactivity, at any time in the operating cycle, is calculated to be nreater than at the time of the checi.,

i.e., the initial loading. R must be a positive quantity or zero. The value of .08% AK has been added*

i to the nonnal value of 0.25% AK to allow for potential maximum settling of the B4C powder in the inverted control rod tubes still remaining in the core. A core which contains temporary control or other burnable neutron absorbers may have a reactivity characteristic which increases with core lifetime, goes through a maximum and then decreases thereafter. See Figure 3 1.2* of the FSAR for such a curve.

The value of R is the difference between the calculated core reactivity at the beginning of the operating j cycle and the calculated value of core reactivity any time later in the cycle where it would be greater

than at the beginning, for the first fuel cycle R was calculated to be not greater than 0.10% AK. A new i value of R must be detemined for each fuel cycle.

I I '

The 0.33% aK in the expression R + 0.33% aK is provided as a finite, demonstrable, subcriticality margin.

This margin is demonstrated by full withdrawal of the strongest rod and partial withdrawal of an adjacent i

rod to a position calculated to insert at least R + 0.33% AK in reactivity. Observation of subcriticality j in this condition assures subcriticality with not only the strongest rod fully withdrawn but at least a i

R

0.331 AK margin beyond this. *

2. Reactivity Margin - Stuck Control Rods i

i Specification 3.3.A.2 requires that a rod be taken out of service if it cannot be moved with drive pressure.

! If the rod is fully inser tt ' ind then disamed electrically,* it is in a safe position of maximum contribu-tion to shutdown reactivity. If it is disarmed elctrically in a non-fully inserted position, that position shall 14 consistent with the shutdown reactivity limitation stated in Specification 3.3.A.I. This assures that the core can be shutdown at all times with the remaining control rods assuming the strongett, operable control rod does not insert. An allowable pattern for control rods out of service, which shall meet j this specification, will be available to the nperator.

I i*

i To disarm the drive electrically, four amphenol type plug connectors are removed from the drive insert and withdrawal

! solcnoids rendering the drive imunoveable. This procedure is equivalent to valvi'nq out the drive and is preferred because, i

in this condition, drive water cools and minimizes crud accumulation in the drive.

'Sec October ??,1974 Inverted Control Rod Inspection and Analysis Report _ and Technical Specifications Change Requ_est. _

! Amendment No. A, 76 8 3/4 3-1

a 1

j -

l 1 The number of rods positted to be valved out of service eww=1d be many more than eldt allomed by th

, myocification, 3mati 8-ly late in the opereirmi cycles housser, the ammwmann of name than ei@t could be i indiastive of a generic crutral red &ive pddam arut the ammtarr will be dutdem. Also if emage Wthin the i contml pre generic rod *d==

&ive auctimimi ed in portfastar, cruds in &ive internal hommisgs, ammet be ru]md cut, than a

{ are=+% a rumher of &iums cannot be ruled odt.

j assisted intauparadar corrosim home nrn=rud in the erdid houslag d &ives at smusmal1his amm.cire==s===

type of j cracking rundd ame in a samhar d rwwinrred, scrum <rudd be guemented in the afamr*=d rods. &iums azul if the eacks prgegeted until numeranos of the swdle housi

mesmand const housing and respired increased menet11anos arter det=r+ing me senza ro

==rere will not be ryar=*=d with a lange nusher of rods with Fallad mile housings. Em rumomal of up arme than aie red per gedrart is pausitted for maiseensnoe provided the men **==i margin =paciflad by 3.3.A.1 is met.

B. Control ax! Mthdramm1 4

1.

j threnal rod &optat accidsuts as disrus===d in the PSWt ari land to a sipifkast crue damsp.

intmytty is maintained, the pommitdlity of a red a====d arw Edm=t is ethdruend. 1he cuartrumm1 positiosa Emme I provides a positisur dents as m1y moogled &isme may rescit this positim. Mth the r==r*=w crittani, neutron j instrtmes**irme regerne to red samumast provides a verificeirut that the rod is ardirusing its drive. Jhmunos of such response to &ive =r=====t =rudd ismiir=** asi esinasaded cizidition.

qtam& art is posedtted for audntununce prorided the d=**=ct usagtn is met.1he neur of no mine that esse &ive per

==1 2.

The control rod hose'ing agpart restricts the r=*nend =====mrt of a cortnot rod to lams that timme int $ms in the estrummty mumate esertc of a housing failme.

The amenat of numreivity M cnald be adSed by tids amull mort of rod with&am1, primary ruwdat apetsma. Wdds is 3mus than a m1 s'ngle with&asm1 incemuset, will not omtritade to muy emage to the

! given in mar +irwi 6.6.3. The &mskyt hamia is giumn in Secticri 6.6.1 of the P9ut, and the damip ammEnd*rwi is I

sirre there would then be so &iving force to raphaty eject a Adtitionally, drive howdng.1histhe agpartapport is rrt is no rewired if all outral rods are fully inmorted sires tte rencear tanM rtsunin ahcritioni esen in the sumfe d rrepime= ejar+fm of the strcrgest coremt rod. '

t 2

J l

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i Jharuhmut au 21, 107 l B V4 3-2 1

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, .- i vv s= =m

A .a w --L L- s a 4 a s, i

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i, li;i'idlIl l l

I!lltk:I<!!!!

11i, 1: p Q

s tl 11! da p

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in i o il 3

' I 'lg ti 2' l jj j , a el I l l gjjf jj!'- l } lI

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l I! ]5 l liiliDlii!!!!'!il

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

o .

Should a control rod drop accident result in a peak fuel energy content of 280 cal /gm, less than 660 (7 x 7) fuel rods are conservatively estimated to perforate. This would result in of fsite doses twice that previously reported in the FSAR, but still well below the guideline values of 10 CfR 100. For 5 x 8 fuel, less than 850 rods are conservatively estimated to perforate, which has nearly the same consequences as for the 7 x 7 fuel case because of the operating rod power dif ferences.

The RWM provides automatic supervision to assure that out-of-squence control rods will no*t be withdrawn or in-serted: 1.e., it limits operator deviations from planned withdrawal sequences. Reference Section ~VII.10 of l the FSAR. It serves as an independent backup of the normal withdrawal procedure followed by the opera-tor. In the event that the RWH is out of service when required, a second independent operator or engineer

, can manually fulfill the operator-follower control rod pattern conformance function of the RWM. In this case, procedural contrn1 is exercised by verifying all control rod positions af ter the withdrawal of each group, prior to proceeding to the next group. Allowing substitution of a second independent operator or

~

engineer in case of RWM inoperability recognires the capability to adequately monitor proper rod sequenc-Ing in an alternate manner without unduly restricting plant operations. Above 20% power, there is no requirement that the RWM be operable since the control rod drop accident with out-of-sequence rods will ,

result in a peak fuel energy content of less than 280 cal /gm. To assure high RWH availability, the RWH is required to be operating during a startup for the withdrawal of a significant number of control rods for any startup. .

i .

.i .

s l.

Amendm' e nt No.12. A9, 9),107 B 3/4 3-4

^ 4.

The Source It has no scram function. It Range Monitor does provide the(SRM) system operator with performs no automatic a visual Indication of safety neutrensystem level. functfont This f.e..is needed for

! knowledgeable and efficient reactor startup at low neutron levels. The resystrement of at least 3 counts

) per second assures that aileguate monitoring capability Is available. One operable SAM channel would be adequate to monitor the approach to criticalltp using homogeneous patterns of scattered control rod

! withdrawal. A minimuu of two operable SRit's al'e provided as an added conservattse. .

i i

5. The Rod slock Monitor (ROM) Is designed to automatically prevent fuel damage in the event of erroneous

! , rod withdrawal from locations of high power density during high power operation. Two channels are

~

provided, and one of these may he hypassed from the console for maintenance and/or testing. Tripping of 2

nne of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the operator who withdraws control rods according to a written sequence. The specified restric-tions with one channel out of service conservatively assure that fuel damage will not occur due to rod i 1

I withdrawal errors when this condition esists. During reactor oseration with certain Ilmf ting control rod patterns, the withdrawal.of's designated single control rod could result in one or more fuel rods with

{

MCrit's less than 1.07 During use of such patterns, It is judged that testing of the ROM system prior

te ~ withdrawal of such rods to assure it,s operability will assure that taproper wf tt 'rawal does not occur.

I It is the responsibility of the Reactor l'ngineer to identify these Ilmiting patterns and the desfgnated 1

. rods either when the patterns are Initially established or as they develop due to the occurrence of in-operable control rods in other than ilmiting patterns. -

t i

i C. Scram Insertion Times . !

j ' T'he control red system f s desfgned to bring the reactor subtritical at a rate fast enough to prevent fuel :

!. damage; f.e.. to prevent the PICPR fras becenta, less than J.07 The Ilmiting power translent is that. I resulting from a generator lead rejection celacident with failure of the turbine bypass systas. Analysts of this transtent shows that the negative reactivity rates resulting free the scram with the average response of l all the drfves as given in the above specification, provide the required protectles, and MCpfl retelns greater than 1.06. Amenhent 26 shows the control rod scram reactivity fasertfon data used in analyzing the transients.

The Ilmit on the number and pattern of rods permitted to have long scram times is specf fled to assure that tlui reactivity f asertion rate effects of rods of long scram times are minfaired. Grouping of long scram time reds is prevented by not allowing sdere than one controi rod fn any group of four control rods to have long fnsertfon tfees. The ulnlaus amount ef react 1vIty to be fnserted dut' lng a scram fs controlled by peralttIng ne more it:en j 101 of the operable rods to have long scram times. In the analytical treatment of the transient. 290 milli-seconds are allowed between a neutron senso'r reaching the scram point and the start of motion of the control rods. This is adequate and conservative when compared to the typical time delay of about 210 all18 seconds estimated from scram test results. Approntmately the first 30 afillseconds of each of these tfee intervals .

result' from the sensor and cfrcutt delays; at this point. the pflot scram soleneld deenergf res. Appros taa tely 120 milliseconds later, the control rod motion is estimated to actually begin. However. 2tm al111 seconds f s conservatively assumed for this time Interval fn the transfent analyses and this is also included in the allowable scram insertfon times of Specification 1.J.C. The time to deenergire the pflot valve scram salenold Is measured durfng the calibration tests required by Spectfication 4.1 l

j Amendment No. 4 H, 87 83/43-5 l !

1he scrm times for all control tuk will be determined at the time of each refuelirg outage. 'Ihe weekly omtrol rod exercise test serves as a periodic check agalmt deterioraticn of the cmtrol rod systs and alm verifies the &ility of the control rod drive to serm since if a rod can be moved with drive pressure, it will scram because of higher pressure qplied durirg scrm. 'Ihe frequency of exercisirg the cxntrol tuk under the (nnditions of three or mrre mntrol ruk out of service provides even further assurance of the relisility of the runnining control ruk.

'ihe occurrence of scrm tines within the limits, but significantly longer than the average, will be viewed m a possible warniry of systanatic prtblon with control ru! drives especially if the nteber of drives exhibiting sud scrm tires exceeds eight, the allcuable ntster of immsle rub.

D. OJntrol Ibd Amtmulators

'Ihe specificaticn for the nteber of accimulators which may be valved out of servi is based on a series of two dimensional XY diffusion theory calculatiam at 20 C. "Ihese analyses prove that the reactor will be stheritical even when the central control red of each 3 x 3 nine ruf array is fully withdrawn.

E. Ibactivity Ananalies During each fuel cycle, excess cperating reactivity varies as fuel ckpletes and as any burnable poison in sigplatentary control is burned. 'Ihe magnitude of this excess reactivity may be inferred frun the critical rui configuraticn. As fuel burnup rugesses, arunalous behavior in the excess reactivity may be detected by crrparison of the critical rod pattern selected base state to the predicted red inventory at that state. Ibwer operatirg base conditicrs provide the most sensitive and directly interpretable data relative to cxxe reactivity. Ebrthernore, using pmer creratiro base conditions permits frecuent reactivity cxmparisms. Ibquirirg a reactivity carparison at the specified frequency assures that a ocipariscn will be made before the core reactivity charpe exceech 1% A k. Deviations in cxxe reactivity greater than 1% A k are not expected and require thoruch evaluation. One permnt reactivity limit is corsidered safe since an insertion of the reactivity into the core would not lead to tramients exceedirp design aJnditions of the reactor systen.

, F. Power / Flow Operatiro Map Allowable carbinaticm of thermal power and total core flow are restricted to Ourve of Figure 3.3.1. Analyses shm that rnactor ascemicn to full power may pro ed cn a nodified pmer/ flow line bomded by tre rod block line tp to a point onlled the 100% intercept point (100% power /87% total core f1cw), frun which omtinuedflow increases may pruJeed in a direct linear manner to the 100% pmer/100% ficw point.(5,6,7)

(5) "MillstEne Ibint tbclear Ptwer Station - Unit 1 toad Analysis Liceme Amduit submittal" MID - 21285, June 1976.

(6) "Millstcm Unit 1 - Inad Line Limit Analysis" FEID-21285-1, Novetter, 1977.

(7) "E>tendal toad Line Limit Analysis - Millstcne Ibint ?bclear Ibwer Station Unit 1" MID-24366, Septerber,1981.

Amendment No. 52,107 8 3/4 3-6

e

?

Ilowever, there are vario os comlitions under which the dissolved oxygen content of the reactor coolant water could be higher 4

than 0.2-0.3 ppm, such as refueling, hot standby and reactor startup.1)oring these periods with steaming rates less than 1% of ,

full flow (80,000 pounds per hour), a more restrictive limit of 0.1 pgun has been established to aswre the chimide-oxygen combinations are maintained at conservative levels. At steaming rates of at least 1% of full flow (80,000 poumis per hour), [

, hoilmg occurs causing deacration of the reactor water, thus maintaining oxygen concentration at low levels. -

When their normal comhsctivity is in itsbecomesabnormalthen range. When conductivity proper normal range, pil and chloride chloride measurements are madeand other whether to determine impuritics or not af f they are also out of their normal operating values. This would not necessarily be the case. Conductivity could be high due to the presence of a neutral salt; e.g., Na2SOg, which would not have an cifect on pit or chloride. In such a case, high conductivity alone is not a cause for shutdown. In some types of water-cooled reactors, conductivitics are in fact high due to

] purposeful acklition of additives. In the case of BWRs, Imwever, where no additives are used and where neutral pil is maintained, comiuctivity provides a very good measure of the quality of the reactor water, Significant changes therein provide j

the operator with a warning mechanism so he can investigate and remedy the condition causing the change before limiting cmulitions, with respect to variables affecting the boundaries of the reactor coolant, are e.xceeded. Methods available to the operator for correcting the ofI-standard condition include operation of the reactor cleanup system, reducing the input of ,

i impuritics and placing the reactor in the cold shutdown condition. The major beneGt of cold shutdown is to reduce the j temperature dependent corrosion rates and provide time for the cleanup sysIcm to re-establish Ihe purity of the reactor coolant. During startup periods and hot standby, which are in the category of less than 1% of full flow (80,000 pounds per hour),

comhsctivity may exceed 2 mho/cm because of the initial evolution of gases ami the initial addition of dissolved metals.

l During this period of time, when the conductivity exceeds 2 mho (other than short-term. spikes), samples will be taken to

assure that the chloride concentration is less than 0.1 ppm.

i The conductivity at the reactor coolant is continuously monitored. The samples of the coolant which are taken every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months i

{ will serve as a reference for calibration of these monitors and is considered adequate to assure accurate readings of the '

l 4

monitors. ll conductivity is within its normal range, chlorides and other impurities will also be within their normal ranges. The reactor coolant samples will also be used to determine the chlorides. Therefore, the sampling frequency is considered adequate ,

to detect long-term changes in the chloride ion content. While conductivity monitoring assures that pil is in abe normal range, '

l umples of reactor coolant are taken and tested for pli once a week as a check. Isotopic analyses to determine major j contributors to activity can be performed by a gamma scan.

1'

11. Coolant Leakar.e The 2.5 gpm limit for leaks from unidentified sources was 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 will propagate.

i

, Amendment. No. 99,107 3

B 3/4 6-3 i

{ '

I i

i l____ _______________________ _ _ _ _

i t

i I

for a crack stro sAlch gives a leakage rate of 2.5,ps, the probability of rapid propagation'is ; less thea 10-3 A leakage rate of 2.5 gpa is detectable aml measurable.

I the 25 spe lluit on total leakage to,the containment was established by considering the removal,capabilitt.ts j' #f tha peaps. . The capacity of eithed of tie drywell floor drain sump gwanps is 50 gpa and the capacity of eltin.r -

of the dryuell' equipment drain sump pumps is also 50 gpe. Removal of 25 gpse from olther of these sumps can be '

accomplished with considerable margin. l l The performance of the reas: tor coola'nt leak detection system will be evaluated during the first year of '

cosmercial operation and tne conclusions of this evaluation wl11 be reported to the AEC.

The main steam Ilne tunnel leakage detection system is capable of detecting small leaks. The system per-4 formance ation willreported will be be evaluated to theduring Atc. the first five years of plant operation and the conclusions of the evalv-E. Safety and Relief Valves Present emperience with the new safety / relief valves indicates that testing of at least 501 of the safety valves per refueling outage is adequate to detect failures or deterioration. The tolerance value is spect-fled in Section III of the ASME Boller and Pressure Vessel Code as ilt of design pressure. An analysis has been Ilmit ofperformed 1375 psigwhich is not.shows that with all safety valves set 11 higher time reactor coolant pressure safety exceeded.

The relief / safety valves have two functions: 1.e..: power relief or self-actuated by high pressure. The- ;

solenoid actuated function (automatic pressure rellefJ in which dent high drywell pressure and low-low water level initiate the,enternal toinstrumentation signals of coincl-la Specification 3.5.D. In addition, the valves can be operated valves manually.

open. This function is discussed The safety function is performed by the same relief / safety valve wl'th a pliot valve causing main valve operation. -

It is understood that portions of the -

l during operation, therefore, the Valve plantPosition sustIndication be shutdoun cannot be repaired toThe or replaced such repairs accomplish 30-day period to do this allows the operator the flexibility to choose'his. time for sleutdoung meanwhile, because of the redundancy provided by the valve discharge tesperature monitor and the continued monitoring of the remaining be compromised.valves by both methods, the ability to de,tpct the opening of a safety / relief valve would tot

The valve operability is not affected by failure of the Valve Position Indi-cation System.

Amendment No. 61,73, 76, 98 8 3/4 6-4

-,sm-w-e-v.- ___ _ _ _ _ _ _ _ __

3.7 Bases A. Primary Containment

, The integrity of the primary containment and operation of the emergency core cooling system in combination, limit the off-site doses to values less than those specified in 10 CFR 100 in the event of a break in the primary system piping. Thus, containment integrity is specified whenever the potential for violation of 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 i pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring. Procedures and the Rod Worth 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

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

i i

! 2. Suppression Chamber -

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system or for releases through the safety relief valves. The pressure '

suppress'on chamber water volume must absorb'the associated decay and structural sensible heat released

during primary system blowdown from 1035 psig.

! Since 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 isothemal compression plus the vapor pressure i of the liquid must not exceed 62 psig, the suppression chamber design pressure. The design volume of the i 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 suppression chamber.

2 Using the minimum or maximum water volumes given'in the specification, containment. p m re during the design j

basis 100,400 accipresults ft in a downcomert is approximately submergence 42 psig of 3.33 feet which and, the is minimum below thevolume design98,000 of 62 psig. Maximum water volu ft3 results l

in a submergence of 3.0 feet. The majority of the Bodega tests were run with a submerged length of four 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.

l The maintenance of a drywell-suppression chamber differential pressure of 1.00 psid and a suppression chamber water level corresponding to a downcomer submergence range of 3.0 to 3.33 feet will assure the post-LOCA suppression pool swell hydrodynamic forces are minimized and consistent with loads assumed for structural analysis of the suppression chamber.

Amendment No. 41, #3, 107 8 3/4 7-1

The maximum temperature at the end of blowdown tested during the Humboldt Bay OI and Bodega Bay (2) tests was j 170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant. I Tests done in the FSTF showed complete condensation with bulk temperature as high as 185'F with a corresponding surface temperature of 230'F. Regarding condensation of steam released through the SRVs and quenchers, test data has shown complete enndensation beyond the 200'F limit of the NRC Acceptance Criteria.

Based on the minimum water volume of 98,0003 ft that the Millstone suppression pool contains, the expected bulk

~

pool temperature rise during the reactor blowdown is less than 70*F. With an initial pool temperature of 90*F, considerable margin exists between this postulated pool temperature and the temperature for which complete condensation was demonstrated.

f For an initial maximum suppression chamber water temperature of 90'F and assuming the normal complement of pumps (2 LPCI pumps and 2 emergency service water pumps) in each loop, containment pressure is required during a small i period of the total accident to maintain adequate net positive suction head (NPSH) for the core spray and LPCI pumps. The availability of the containment pressure required to maintain adequate NPSH during this interval is assured by the contairunent spray interlocks as described in Amendment 18.

If a loss of coolant accident were to occur when the reactor water temperature is below 330*F, the containment pressure will not exceed the 62 psig design pressure, even if no condensation were to occur. The maximum allowable pool temperature, whenever the reactor is above 212*F shall be governed by this s meification. Thus, specifying water volume temperature requirements applicable for reactor water temperatures a mye 212*F provides additional margin abnve that available at 330*F.

! In addition to the limits on temperature of the suppression chan6er pool water, operating procedures define the action to be taken in the event'a relief valve inadvertently opens or sticks open. This action would l include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat

exchangers,and(3)initiatereactorshutdown.

Robbins, C. H., Tests of a Full Scale 1/48 Segment of the Humboldt Bay Pressure Suppression Containment, GEAP-35%,

November 16, 1960.

Bodega Bay Preliminary Hazards Summary Report, Appendix 1 Docket 50-205 December 28, 1962.

NEDE 24539P Mark I Containment Program Full Scale Test Program Final Test Report, April 1979.

NEDO 24575, Mark I Containment Program Plant Unique Load Defnition, Millstone Nuclear Power Station -

Unit 1, March 1979.

1 NRC Acceptance Criteria for Mark I Containment Long Term Program, Rev. 1. February 1980.

NUREG-0661 Mark I Containment Long Term Program, July 1980.

Amendment No. I3, 5I, 73 B 3/4 7-2 4

,,-**'t"1*g'**--"*pW'=*'F"w*'w9-*w etw. w www e-ew es e- w g i--we-* e v --- ww www*w+vnm=mi--~e-*iew-ww--e- - - + - - = =e -- ww=---- ii-----=---+------------------------v---

6. Oxyaen concentration 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 loss 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 to present in sufficient quantities to result in excessively rapid recombination, could result in a loss of cont 8c at intergrity.

The 4% oxygen concentration minimizes the possibility of hydrogen combustion following a loss of coolant accident. Significant quantities of hydrogen could be generated if the core cooling systems did not ,

i sufficiently cool the core.

j 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 leek inspections during a startup is judged i i

prudent in teons 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 l leaks in the primary system, leak inspections are scheduled during startup periods, when the primary system i

is at or near rated operating temperature and~ pressure. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period to provide inerting is judged i

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 in-leakage through the primary containment. However, at least once a week, the oxygen concentration will be determined as added assurance.

3. 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 4

pressure to the design negative pressure so that all leakage should be in-leakage. Each of the two ,

l fans has 100 percent capacity.

I High efficiency particulate absolute (NEPA) filters are installed before and af ter the charcoal absorbers to minimize potential release of particulates to the environment and to prevent clogging j of the iodine absorbers. The charcoal absorbers are installed to reduce the potential release of radiciodine to the environment. The in-place test fesults should Indicate a system leak tightness '

of less than 1 percent bypass leakage for' the charcoal absorbers and a IIEPA efficiency of at least I

99 percent removal of DDP particulates. The laboratory carbon sample test results should fadicate a i

l Ame,ndment JDA, J02,107 ~B 3/4 7-5 e ,

radicsctiva methyl icdida removal efficiency of at least 95 parcent for expsetad cccidznt ccaditicas.

,,- i If the efficiencies of the llEPA filters and charcoal absorbers are as specified, the resulting doses will 'Ine less than the 10 CFR 100 guidelines for the accidents analyzed. Operation of the fous -

nignificantly different from the design flow will change the removal ef ficiency of the HEPA 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 off-site power sources (345KV or 27KV) ,

and one emergency power source would provide an adequate and reliable source of power and allow completiun of annual diesel or gas turbine preventative maintenance. -

1 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 containment is an integral part of

' the complete containment system, secondary containment is required at all times that primary containment is required.

1 i D. Prjaa ry 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-

,, neut in the event of a loss of conlant accident.

Amendment No. 8, JOI,107 8 3/4 7-6 w---+ e

-- e- m -,m, e.-,-a w-w --e--w,,,-- e&we -e.w,- y-w - - - -- m -, - - - - - . - - - --- -. ---

5.0 DESIOi FENITJRES 5.leSite 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 minimun 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) We Connecticut Light and Power Cmpany, Western Massachusetts Electric Cmpany or Northeast Nuclear Energy Cmpany or their corporate affiliates for use in conjunction with normal utility operations and (ii) to the two leasee's under the leases referred to in the follw ing paragraph.

A United States Navy research Laboratory and a desalination pilot operation of the Maximtsn Evaporator 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 requirementn of 4

the owner of the site.

5.2 Reactor A. %e core shall consist of 580 fuel assablies.

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

5.3 Reactor Vessel

%e reactor vessel shall be as described in Table IV-1 of the FSAR. W e applicable design codes shall be as described in Table IV-1 of the FSAR.

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

B.

We secondary contairunent shall be as described in Section V-3 of the FSAR and the applicable codes shall 1

be as described in section XII of the FSAR.

Amendment No.16, 76, 9J ,107 5-1 -

4

C.

designed in accordance with standards set forth in Section .

e V-2 of the F 5.5 Fuel Storage A.

The new storage facility shall be such that the K is less than 0.95. eff dry is less than 0.90 and flooded B. The K value*f[ of the spent fuel storage pool shall be less than or equal to 0.90,This K bundles issatisfied less thanif the maximum exposure - dependent K of the individual fuel eff 1.35.

5.6 Seismic Design earthquake ground motion with an acceleration of 17% of gravity.The e reacto reactor building.to determine the earthquake acceleration applicable to the various elevations in theDyna Amendrent No. 37,107 5-2

.

ML20138G082

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