Text

Tech Specs Section 3 Re Limiting Conditions for Operation
	ML20033C605
Person / Time
Site:	Maine Yankee
Issue date:	11/30/1981
From:	; Maine Yankee
To:	;
Shared Package
ML20033C604	List: ML20033C603; ML20033C605;
References
	NUDOCS 8112030484
	Download: ML20033C605 (93)
	v • d • e

{{#Wiki_filter:.. _ 4 1; j ATTAf M NT 1 / m. r c J

i Maine Yankee, Atomic Power Company 1

i /a . /-, ~ Technica1TSpecifications. e ^ i V,.. Section 3 // ~ j s-j >c ,/ 4 o-Limiting Conditions for Operation i f<' . e., 4 .,,.,e /. .i k . i i r / ~ 911'2030484'8 % NOV 3 01981 KDRADOCK05 R

) n. 3.0 LIMITING CONDITIONS FOR OPERATICMS Aoplicability: . Applies to section 3 of these Technical Specifications. Objective: / To specify general regulatory recuirements for compliance with these specifica.tions and appropriate remedial actions when compliance cannot be attained.: 7j I Soecification: Violation of a LI;niting Canbition for Operation: A. ~ If b Limiting Condition for Operation (LCO) in Section 3 of the Technical Specifications,is not met, the following sequential re-medial actions must be taken until compliance with the specification is achieved.' }/ b perforft any remedial action permitted by the applied specification *

See Note 2.

commence a reactor shutdown and place the plant in a Hot Shutdown * - Condition within 6 hours after the discovery of the non-conforming condition or after any time period specified by the ' remedial action. 3. commence a reactor cooldown and place the plant in a Cold Shut-down Cordition within 30 hours after the discovery of the non-conforming condition or after any time period specified by the remedial action. B. Operability of safety related components with emergency power sources: If a system, subsystem, train, component or device is determined to be incperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: 1. Its corresponding normal or emergency power source is OPERABLE; and 2. all of its redundant system (s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this specification. Exceotion: This specification is not applicable in the Cold Shutdown or Refueling Shutdown Condition. Remedial Action: Unless both conditions B (1) and (2) are satisfied follow the actions specified in A above. J 3 U-l NnV 3 01981

a Basis: Specification A assures compliance with 10 CFR 50.36 which states: "When a limiting condition for operation of a nuclear reactor is not met,

the licensee shall shut down the reactor or follow any remedial action permitted by the technical specification until the condition can be met."
Specification 8 delineates additional conditions that must be satisfied to permit operation to continue by systems, subsystem, trains, components or devices recuired by these LCO's.

It specifically prohibits operation when one ECCS train is inoperable because its normal or emergency power source is incperable and a system, subsystem, compunant or device in the other train is inoperable for another reason.

Note:10 CFR 50.72 (a)(5) requires a licensee to notify the NRC Operations Center within one hour by telephone of the occurrence of any event recuiring initiation of shutdown of a nuclear power plant in accordance with Technical Specification limiting conditions for operation.

l l i i l l i t 3 0-2 l NOV 3 01981

3.1 IN-CORE INSTRUMENTATION-Applicability: Applies to the operability of the in-core instrumentation system for-the purpose of_ recalibration 'of the ex-core symmetric offset protection system. Cbjective: T63ecify the functional requirements which must be satisfied for the in-core instrumentation system to be considered operable for the purpose of calibrating the ex-core symmetric offset protection system. Specification: The ex-core symmetric offset protection system shall be re-calibrated monthly, utilizing the in-core instrumentation system, whenever reactor reactor power level is greater than 90% of the maximum power for 2 or 3 loop operation. For the in-core instrumentation system to be considered operable for this purpose: L 1. At least 75% of all in-core detector octant positions shall be represented, and 2. A minimum of 2 in-core detector locations per core Quadrant shall be operable. ~ NOTE: An operable in-core detector octant position shall consist of a position with a minimum of three operable fixed di.ectors or where a moveable detector trace can be taken. Remedial Action: Power shall be limited to 90% of maximum power for 2 or 3 loop operation (whichever applies) if re-calibration of the ex-core I symmetric offset protection system has not been accomplished witnin the previous 30 days. Basis: The in-core detector system uses 45 radial locations throughout the core. A number of these locations have additional provision for moveable detectors, while the remainder have strings of fixed self powered detectors. This instrumentation can be used to determine the power balance between the. top and bottom halves of the core in each of these locations. Moreover, a fixed detector string would still provide adequate capability with only three of its four rhodium detectors functioning. Thus the full system has more capability than would be needed for the calibration of the ex-core detectors. After the ex-core system is calibrated initially, recalibration is needed only infrecuently to compensate for changes in the core, due to fuel depletion and for changes in the detectors. "~ NOV 3 01981

If the re-calibration is not performed, the mandated power reduction assures safe operation of the reactor since it will compensate for an error up.to 10% in the ex-core detector system. Experience at Connecticut Yankee has shown that drift due to changes in the core or instrument channels is very slight. Thus limiting the operating levels to 90% of the maximum two or three loop power levels is very conservative for both operational modas. 3.1-2 P

3.2 REACTOR C00LANT SYSTEM ACTIVITY Aoplicability: Applies to measured maximum activity in the reactor coolant system. Objective: To ensure that the reactor coolant activity does not exceed a level commensurate with the safety of the plant personnel and the public. Specification: A. The specific activity of the primary coolant shall be limited tr less than or ecual to 1.0 micro C1/ gram DOSE EQUIVALENT I-131. Exceotion: With the specific activity of the primary coolant greater than 1.0 micro C1/ gram DOSE EQUIVALENT I-131, but less than 60 micro Ci/ gram DOSE EQUIVALENT I-131 operation may continue for up to 48 hours provided that the cumulative operating time under these circumstances does not exceed 800 hours in any consecutive 12 month period. Remedial Action: 1. With the total cumulative operating time at a primary coolant specific activity greater than 1.0 micro Ci/ gram DOSE EQUIVALENT I-131 exceeding 500 hours in any consecutive six month period, prepare and submit a report to the Commission per the reporting recuirements of Section 5, within 30 days indicating the number of hours above this limit. 2. With the specific activity of the primary coolant greater than 1.0 micro Ci/ gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or greater than 60 micro Ci/ gram dose ecuivalent I-131 for more than one hour, be at least subcritical with Tavg less than 5000F within the next 6 hours. 8. The specific activity _pf the primary coplant shall be limited to less than or equal to 100/E micro Ci/ gram (E defined below). _ ith the specific activity of the primary coolant Remedial Action: W greater than 100/E micro C1/ gram, be at least subtritical with Tavg less than 5000F within 6 hours. i C. With the specific activity of the primary coolant greater than 1.0 micro Ci/ gram DOSE EQUIVALENT I-131 or greater than 100/E micro C1/ gram, perform the sampling and analysis requirements of Item 1 of Table 4.2-1 until the specific activity of the primary coolant is restored to within its limits. A report shall be prepared and l submitted to the Commission and shall contain the results of the specific activity analysis plus the following information: f 1. Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded, i l l 3.2-1 NOV 3 01981

i l 2. Fuel burnup by core region, 3. Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded, 4. History of de-gassing operation, if any, starting 48 hours prior to the first sample in which the limit was exceeded, and 5. The time duration when the specific activity of the primary coolant exceeded 1.0 micro C1/ gram DOSE EQUIVALENT I-131. NOTE: E is the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in Mev) for isotopes, other than lodines, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. I Dose Ecuivalent I-131 is determined as that concentration of I-131 (micro Ci/gm) which alone would produce the same thyroid dose as the cuantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation.of Distance Factors for Power and Tect Reactor Sites." BASIS: The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour doses at the site boundary will not exceed an appropriately small fraction of the Part 100 limit following a steam generator tube rupture. l l 3.2-2 ' NOV 3 01981

3.3 REACTOR COOLANT SYSTEM OPERATIONAL COMPONENTS Aoplicability: Applies to the operating status of the reactor coolant system equipment. Objective: To specify conditions of reactor coolant system components for reactor operation. Soecification: A. Reactor Coolant Pumps 1. At least one reactor coolant pump or one low pressure safety injection pump operating in the residual heat removal mode shall be in operation providing flow through the reactor when the reactor coolant system baron concentration is being reduced. 2. At least one reactor coolant pump shall be in operation providing flow through the core with its steam generator capable of performing its heat transfer function whenever the reactor is in a critical condition. 3. At least three reactor coolant pumps shall be in operation providing flow through the core 'with their steam generators performing their heat transfer function whenever the reactor is in a power operation condition. Exception: The reouirement of 2 and 3 may be modified during initial testing to permit power levels not~ to exceed 10% of rated power with three loops operating on natural circulation. 9. Pressurizer Safety and Relief Valves 1. At least one pressurizer code safety valve shall be operable whenever fuel is in the reactor and the reactor coolant system is isolated from the residual heat removal system and the head is on the vessel. 2. At least two pressurizer code safety valves shall be operable whenever the reactor is critical. 3. One power operated relief valve (PORV) and its associated block valve shall be operable whenever the reactor coolant system temperature is greater than 210oF. Remedial Action: In the event that a PORV or its associated block valve becomes inoperable, either restore the PORV or block valve I to operable status or close the associated block valve and remove I power from the block valve; otherwise, be in at least hot shutdown within the next 12 hours and in colo shutdown within the following 30 hours. i 3.3-1 lNOV 3 01981

C. Pressurizer l. The pressurizer shall be operable with at least one bank of proportional heaters and a water level during normal system operation between -28 and 60 percent whenever the reactor coolant-system Tavg is greater than 5000F. 2. The pressurizer spray system must be lined up to provide continuous pressurizer spray flow whenever the reactor is critical. Basis: Reactor coolant pump flow and steam generator heat transfer capabilities are specified to -assure adecuate core heat transfer capability under all operating conditions from criticality to full power. Three loop operation is specified to assure plant operation is restricted to conditions considered in the LOCA analyses. The exception permits testing to determine decay heat removal capabilities of the primary system prior to higher power operation while on natural circulation. Following a loss of offsite power, stored and decay heat from the reactor would normally be removed by natural circulation using the steam generators as the heat nink. Water supply to the steam generators is maintained by the auxiliary feedwater system. Natural circulation cooling of the primary system requires the use of the pressurizer heaters or high pressure safety injection pumps to maintain a suitable overpressure on the reactor coolant system. Alternatively, in the event that natural circulation in the reactor coolant system is interrupted, the feed and bleed mode of reactor coolant system operation can be used to remove decay heat from the reactor. This method of decay heat removal requires the use of the emergency core cooling system (ECCS) and the power-operated relief valves (PORV's) in the pressurizer. The PORVs can be operated either manually or automatically in the Maine Yankee design. Block valves are provided upstream of the relief valves to isolate the valve in the event that a PORV valve fails. When reactor coolant baron concentration is being reduced, the process must be uniform throughout the reactor coolant system volume to prevent stratification of reactor coolant at a lower boron concentration which could result in a reactivity insertion. Sufficient mixing of the reactor coolant is assured by one low pressure safety injection (LPSI) pump operating in the RHR mode. When operated in this mode it will circulate the reactor coolant system volume in less than 12 minutes. The pressurizer volume is relatively inactive; therefore, it will tend to have a baron concentration higher than the rest of the reactor coolant system during a dilution operation. A continuous pressurizer spray flow will maintain a nominal spread between the baron concentration in the pressurizer and the reactor coolant system during the addition of boron. Without residual heat removal, the amount of steam which could be generated at safety valve lift pressure with the reactor subcritical would 3.3-2 NOV 3 01981

be less than half of one valve's capacity. One valve, therefore, provides adequate. defense against overpresserization when the reactor is suberitical. Overpressure protection is provided for s 1 critical conditions. The safety valves are sized to relieve steam at a rate equivalent to the peak volumetric pressure surge rate. For this purpose one safety valve is sufficient; however, a minimum of two safety valves is required by Secticn III of the ASME Code. 3.3-3 l iNOV 3 01981

3.4 COMBINED HEATUP, COOLDOWN AND PRESSURE-TEMPERATURE LIMITATIONS Acolicability: Applies to temperature and pressur.e conditions during heatup and cooldown of the reactor coolant system. Objective: To maintain operational limits within design boundaries of the reactor coolant system. Specification: A. Reactor Coolant System 1. The rcactor coolant system shall be operated within the limits set forth in Table 3.4-1 and the pressure-temperature limits derived from (2) below. Remedial Action: If the reactor coolant system is subject to conditions outside of the above limits the reactor shall be brought subcritical and an ergineering analysis of the con-sequences shall be made prior to restoration of power cperation. 2. The pressure-temperature limits for reactor coolant system operation shall be revised at each refueling using the following procedure. a. The pressure-temperature limits for reactor coolant system operation shall be as developed by superimposing fluence-dependent heatuo and cooldown limits into the basic ASME Section 3 limits of operation (Figure 3.4-1). At each refueling the heatup and cooldown limits will be modified to account for material property changes in the rec: tor vessel projected dirough the next core cycle in accordance with the following procedure: 1. Project the cumulative MWH(t) on the vessel through the [ next core cycle. 2. Dete_mine the associated fluence to the vessel from Figure 3.4-2. ( 3. Determine the shift in RTNDT at the 1/4t and 3/4t from Figure 3.4-3. 4. The beginning of life heatup and cooldown limit lines in Figures 3.4-4 through 3.4-7 shall be shifted parallel to the temperature axis (horizontal) in the direction of increasing temperature, a distance equivalent to the shift in RTNOT at the 1/4t and 3/4t as applicable. 3.4-1 NOV 3 01981 1

The following table provides the shift parameter to be applied: CURVE SHIFT PARAMETER Heatup, upper limit 1/4t Hestup, all other rate limits 3/4t Cooldown, all limits 1/4t 5. Superimpose the shifted Figures 3.4-4 through 3.4-7 onto Figure 3.4-1 to provide the appropriate operational limits for heattp and cooldown during normal and hydrostatic test operations. B. Reactor Core 1. The reactor shall not be critical if the reactor coolant pressure is less than 400 psig or greater than 2400 psig. 2. The reactor shall not be critical (other than for the purposes of low power physics tests) if the temperature of the reactor coolant is: a. less than liloF plus the shift in RTNDT at the 1/4t (as determined in A.2.a.3), or b. within 400F or less of the applicable heatup curve (as determined in A.2.a.4). 3. The reactor shall not be critical without a steam bubble in the pressurizer. 4. The reactor shall not be critical during inservice leak or hydrostatic testing of the reactor coolant system. C. Residual Heat Removal System 1. The residual heat removal system (RHRS) must be isolated whenever the reactor coolant system pressure exceeds 600 psig or the temperature exceeds 4500F. D. Reactor Coolant System Low Temperature Overpressure Protection 1. The power operated relief valves, aligned for the low pressure set point, and the RHR s'pring relief valves shall be operable for RCS overpressure protection whenever the RCS is less than the minimum pressurization temperature and the RCS is not vented. Exceotion: One power operated relief valve or RHR spring relief valve may be inoperable FJr 7 days. Remedial Action: If tne conditions of D.1 are not met, the RCS shall be depressurized and vented within 8 hours. 3.4-2 10V 3 01984

2. No more than one HPSI pump may be energized at PCS temperature below 2200F. Exceotion: A second HPSI pump may be energized for up to 5 minutes for the purpose of rotating operating equipment. 3. Reactor Coolant Pumps may be started (or jogged) only if there is a steam bttble in the pressurizer with a maximum level of 80% or 0 the steam generator temperature is less than 100 F above the reactor coolant temperature. Basis: The heatup and cooldown limit curves (Figures 3.4-4 through 3.4-7) are composite curves which were prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup or cooldown rates of up to IOOoF per hour. Linear interpolation is permissible. The heatup and cooldown curves were prepared based on the beginning of life RTNDT at the reactor vessel, and include adjustments for possible errors in the pressure and temperature sensing instruments. The reactnr vessel materials opposite the core have been tested to Appendix G of 10CFR50 to determine their RTNOT. Reactor operation and resultant fast neutron (E greater than 1 Mev) irradiation will cause an increase in RTNDT. As a result of irradiation tests of actual vessel materials, the shift in RTNDT can be determined at the critical 1/4t and 3/4t locations from Figure 3.4-4. The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradittion surveillance specimens installed near the inside wall of the reactor vessel in the core area. The pressure-temperature limit lines shown on Figures 3.4-4 through 3.4-7 for normal operation and inservice leak / hydrostatic testing, as well as the limits on criticality have been provided to assure compliance with the reauirements of Appendix G to 10CFR50. The maximum NDTT for all reactor coolant system pressure retaining materials, with the exception of the reactor pressure vessel, has been determined to be 400F. The Lowest l Service Temperature limit line shown on Figure 3.4-1 is based upon this i NOTT since Article M3-2322 (Summer Addenda of 1972) of Section III of the ASME Boiler and Pressure Vessel Code, requires the Lowest Service i Temperature to be RTNDT + 1000F for piping, pumps and valves. In addition, a 600F margin is added to this for conservatism. Below this l temperature, the system pressure must be limited to a maximum of 25% of l this system's design pressure of 2485 psig. The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASE Code requirements. i ~' NOV 3 01981

2 a

n o& as9 N" TABLE 3.4.1 LIMITS OF OPERATION FOR Tile REACTOR COOLANT SYSTEM M STEMI C!!'?.RATOR Reactor Limit vessel Pressurizer Primary Side Secon.lary St.le haximum lleatup Rate ( F in any.one hour perio,d) 100 100 100 diaximum Cooldown Rate ( F in any one hour period) 100 200 100

w linimum Pressurfzation E

Temperature ( F) 200 70 70 100 !!aximum Pressure Below Minimum Pressurization Temp (psig) 621 500 500 230 !!aximum TemgIerature Dif ference 340 Between Operating Loops ( F) C, ' e 2 w

j FIGURE 3./.-1 \\ u.. ..r.. s ~O ..s 'I. REACTOR COCLET

s. n..,.

...r i ..s m LIMITS CF OPEP.ATION ..r ..._t......._... 2200 1... ..2 f .._i........_......_..... _.... _ _ _...... ......2.... . _...r.._...i._..... _....... 2000.__._..__...._.._..........__.._.._...._. g o._...._.._.._...._.............._....._................r....._.._..... ... _....... _............a.... .~ 1800 .. _ _ _. _... _.. _. _. _... _. _... _..... _.. _........ a ,_....n...._.._. _. _ _ _.................._.r._............. _..... ........_i...._.i_.............. _. _. _.... _. _.. _., _... ......o. ..._2..... .. _....i... _.. ...._...__.2...._._.... -. .... _... _. _. _.. _.... _ _ _ _.. _.... _ _ _. _. _.. _.....___.m 160 ... _...... _.. _.. _. _ _ _... _.___..._._r__ ...._.._......_...._.._.i_._......_._.._.__._...._.__..__.__._._.,...-..... _ -. _....., ... _. ~.. _. _ _ _ _ _.. _.. _ _... _.. _ _. _. _. _............ _.... ..... _.. _... _ _.. _ _... _. _ _.. _... _.. _. _.... _ _ _.. _. _ _ -.. _, _. _.._...._.__.__._..t.._._..- ..__s. _ _.._. _ _._._. _ ....._.s .._......_._...__z..__._.. _.. _. _. _.... _ _.. .___._.___.__..._._._._________.____.._.i... 00.._.._.___._._._....__._.__._....... _..... _...... _.. . _..... _... _ _........x.._....... _... _. _ ._.a.._. m t.n 93 ... _..... _............._..___...__..._....._._......._..s.. ... _. ~... __... a..__ _..- . c.,, 1200 ......t..__.._. ."".'".".'1 _a.._. u .._ _.i_ _ _. __-_.__.._.. _....... _. .......__.____-_i._.__..__._.__......_..__..___..... __ . _ = p 7_..___...__._.._._._._... _ _ _ _. -. _ _. _ _,.... m ia ..t. g _...... _ _. _.. _ - _. _..... _.. _. _..._._s. .__.r._...__...._i. ..1___._.._..._____.._._.._-_.._.. 1000._.__._..______.___.._.___.__. .......i._......__...____._........i_._. r._......__..__._.... _. _.. _ _.. _.. _......_4_____. _ _... _ _ _ _.. _.. _ _... _ _ _ _. _. _...._r.__._._._..._.._i___..._.._._... .._.._._._._a......_....._.._..........._...._.-i.._..___._.m....... .._.._:._.. -........_.._.a._._..._._._._.... ............._...i._.._..._.. .t.._..._.... _.. _ _ .._.__..._.r._.._......_.._...__........... ...... _. _........_......_._._._..i..__ _... ___.__._.._..._..___......n._..._........ .......... _. _. _...._._........i._._.__.... _... .. _..... _....._.:..._...____.i._._..._. ......i.._._.......i..... ._._i.. 1 .. _.. _... _.. _.......,......._.....i..._.._:.._........... i. _..i.... t l ..i...._.. .i..._............., . _....f_.._i..... ._.._...._.....t.............,. ...i..... I .....l..._ 8 i. .. i. .. ~. ......i 00_ i....

e s-x.... _. _ _. -

i . ~.. _. _... e -.. _. n. .......___.1...,._.s_.._..i r .. -. i. ..i g 9 . 9 .g.. ..g .g ~ - .t ~ 200 - L _. 6 = f ...... _..... _. _ _ = i 1 i... ..i..._.. l i i -i os .r.1_ NOV 3 01981 TEMPERATURE (V) i

FIG.?. ? 3.4-2 FL*CUCE VS. 32..:? TOR 1/4T AND 3/t.! ?CSITIONS .s .v.c. v. r_ ._r..e r_ sa - = - r m M o a 3 3 i _a. I .1 sl y 'i l 1 f ,o a 's I 10-' .t -/ __.=---__ _ _. __.==E s - - ~ __ p f -= -,/ / 3 p. 1 _.z s. A 3 w i / . <,t-4 N n u ,o. .s i /n S e ,I w ,a Z

/

u.: .i , / I i ] ./ / e n e i

n.

/ 6 f. 18 ,f 10 f 8P A' 3 s 2 D i m S H ---f- "E 2

1_

l i ,I ,i e e i j t I e I f I 10 '1 ,i i e 8 1 2 3 4 5 6 7 8 WEth x 10 NOV 3 01981 3.4-6 Amendment No. U, 58 JUL 101981 (.....-..... P ,=w-g urw1-t-a%-g-.9,,g,_ ,_y,m,

3 c,:c E w*,w 4 = I( ' i !fi ', Il ( Il lt l l.ga l q :. ll p L d 5 qlp ' __ j' .ili:' I U ,i' l' l lj' [ )- '!l i! l y, i l-fl 4 l. 9 1 3 0 1 9 9 1 t n D17 P. 2 i F T 7 l a r 9 r e n Gr1 F i r YSl it ? I f l W nNI t I i l' r WtuA y. T T J i 1 If I I A fl i p!i il l T fq! 9 j l ii l i I li b! 8 I -l 7 6 ) V. 5 iM A F. C 1 N M l 4 E F. i l p T l l. ( y F l 0 4, . 3 1 2 'I F 0 n j R O 1 /a N N n T O P T I 2 R P C C T N W t J iS E l k U / F I I l 1 gi, ,I: l f3l I, ,i i'l@ 9 ,i!

i' 1I
i' l

l !i

l l

,.I! - ld - j-I! IU 8 i ] II. ,i, i i i'l 7 } i; i f

i. 5 7

i. 1 l 4 0 I l 1 gt l]l{ I i I r 3 2 0 o g 0 0000 0 0 g 0 0907 6 5 o, 0 2 1 s =

g. - [2l l

n C i f - r@ - ?. e ,m gr ogS w.*4 t c a ma< C O E,-

FIGU2E 3.4-4 f l,. .l... _. 1.. p 2400 i 'I

b i J

BEGIMI::G OF LIFE HEATUP LI!'ITS

j. #--

r {': FOR ::CF".AL O?F. RATIO:: l ' i '. ~- rz H i ii r. . l... a. r-3r...>:. - u.. _, D,., y u. _;. \\._ _a r: .n 2'500 "- ~ ' ' i. NOT

.i..._: !.

f..:l [ 8-8 "i -30 F RT = 8-J. J __,j _ y 8:-, l - [ :.l J .j.; 1 2000 l j . l i

..f.=.

.i I I I j .i

=_ -

-...J:.3.:. f. I. }. 'r_.%. *. .S-. ,}(...._j 1800 i'- - i'- " i ? i 1

'::I sin -

..; it m-a : :-

.;: = l i.;2 7 -

J;;_ j;iri

f. C-l j

[ =Tl. n;.

,m: T.j g.::.:q ;.. ; ;,.... :. :

2: ; iq.,,. {,: -,. j -..;i p. ujg:g .........:.a =:

.: : - :i 2.r: :. I. f.. j. -

.. j- ~ j;;. ~ ~ ' ' _:: :.:_iii... 1600 p.ni:'._'. a; r w, = r.-.:::;.. _._ _:.. i 1 i l----l-1. -. -- -:.., :ina u ;. :: -i.b-2. n. =..n-- - --- : : ;... :: 2 :.---r. : r :-

=._; u..=. :.;:.. m.: :;. ;n;-.. :.i..= :.:. ;.:--: - :.l

y. '"i 95flibi5fi:E ;5:i: Fili 1:Juc :iin jEh/7 f h hu!~:'i~.Eli.i:ia -

.......'gID'-i. O l f' '.-~'I / :i : .......': n~! I" 1400 5'~#"ii:+ II"!#:ifSii "" :-P2 -i": Fl ......- : - - T. . 8 : _n. ......n :. :n..=. r _.n..c. _=.:n. -.u. __n.=._.=_.=_n.....=. t .T K"-5-f.-! ".." 2/ ::':1:i- : I-iiE: :' aiti :- "f' _- : - I d!".'_E#d-Ji.... b: f.- . :.j .__n.;n-j u-es =. r. : - - ::: : ;..- :p. ::.:.:f.J - ;_:-- - :-- :l. o. ., -,. i..

. =_. - i _:..,.......... ;. -- !'

~~---".........l-... -""!. ~- V . _............./ ;

= i ::=

1200 __s _...:, :_=istis : :_ r = = l@

. : :,.._.. =_. 3,...... =..=.. :.. _ =..- : :.. =... ::._: :.-......... u. n.

...__...y.r.u......_..- p ei :i'. Pi.i i' :'.!E.:9. 3...iti :ii.'..l.i..;u.. P 2=ijT._fii_=__'..lir..1.T_=. i.. i.i.-ii. _._iT. ii_ 9. T..-i.:../.,.: us g .b_ _4.___.- .- _...: :.:= -- :- :ru r-r :- m. - ru ; _;;

..-: i..=m r z.n. =,t==l.= -

- -- ::= = =. .a . :=- a: .=r- . p. .:;. ; f-;:7.i ;.p;ii.;;i=[:;-_::iin.:i _ a:j_ ::=.ii=ji:j.:.:=_.'.-4==...}. ;in_-i,.--. :- 1000 ,-__.:.._.:.-..g...-.........,._:n_ n_ a.. :. : ._..e

=,.=_ y.=_ =.:..._=... :.. :..: r. mu:. ::=_.-. ::.:. g y_ ~.. :

_: i.E.:8.i.;i 7:ii,F..-i :-fi.i::i'..!.-G.E_.I.h:.!!y ; -.t -f .m iT..1..~it".li.i!r ". - Z 2; N! INN 5.ili.E ~il:I!.*z"I-I~ I. *.,, " -I.l ~i.

M I

siM :ii"ti:luiti!=i-i;i? "H & 8,T, , ~, - - -~' iiI Mil" - l

g, '.:

800 =: 9n.

= u. -

..:..a 4.. "O

j.. _,,:,,,

4 1 i -r . o. .#.,, i O

=

a:nc ;-. ::;. _ :: t..... - _cv cp,le j8: i -le. ..: "l.' ET"31 ;ii'i.uq;i:+:ej i iNj - Y O 70 .l. = 2...r. j u.r..=.. :. .~: :.. u., o..N \\.-

r. l-o t:

x Co 4 i. . p .I e

.: ?

. /n - O. 600 .=...:u.d... .l.

i...

o .~- ,,_. _, e .;;:n..}.. .i. .] ..e-a l v l:. . i. i, l.ui -i a. :p it i i /* - l'

l:

l-l 400 6

q.

8..

j..._. !

___.8 ~ v _ g._.:. _. l 1 .- - r u _h.u;..j:.. r- ..i. _.-; :.;._. a;4 i .a i i i i y 200 - - t l --~.._:.. ..).._. . i-i - :I - s ; - ,.a...... .a I. t q l = _ -. i..

s.. l..u__.q; l

t ' .l. ...l ,l i-i o ' 0' 100 20'0 300 d.J/ -200 -100 3 y. 3

N0% 01981 mf m ATURE (oP)

I

FIGUFl 3.4-5 'j_, i [i6l~ ..i..... i...:..~.I :.... l :.: !

I

t 2400...

BEGINNING OF LIFE C00LDOC LIMITS !*.I FOR NOTU*A'. OPERATION E ....::-ij [j.i I:.- BAS r D G.,. , _. D u... A, RT.nT = -30 F 0 5"- ..l 2'_nD

i J _.

i .a

g.. _

,y. ..j_ ca... .s t 1:: - n. .2 2030 ' ~

i!

- - - -. im ii-ri. -r - :.:.:2 ;..

t.. r.. n_ :...

.-d! : 1-::!: : j '. --i i M' I!.2i: i'. .. ; : = := R. t_ t.- 1800... .. -r.. i l 1 f jh.i... I :Mi 5.~ _:.4 t. _.... 'i :*. ' : i. U ! ' '.!! *'. :. I. : :. :.'. (' d.._ .t... ij"i!:4: :pi....i.::: _;.: El ; ;;;di':-.:i - ,1:i f ---,-- 3 -,p.. ; _ryn;:_ ;. .i::f='.:--[.Q J.ji". j.: i=. '!...r-:= r.-?:i .. _in i=. :=- ' =l: ' " i r - 1 = 2=r= '. 1600 ....... : :c .

i '=-fi. I MI i

i i 5553-P=5Ei I=i. :*:-) ID. T: riai-si24 - ~~ . :2. u.. r..: : : _., -.....l _:.:....... _.... -.-. ::.._ e. ! =.. =. _....r _ _.u..=.: -j=sT.[ 2 "r - 7 ;--if ' (if=!ji';ji, - ).:.apjip" -ici ;i s 1400

5' E

I SIE~ *i # 5 M = .. ;._-,.. _ :...... _.. _ :.._. @:.... i

it !-" = !!

r:- -!} ' 3 }i5-5. L.= i -di'.; ~_ - 7 .x. q u..a.... -., - -..,..__......:.. r,._, O ..=:.._..... ,e.e ..... ~.2 ~.Mi"... O

7 ; *.' ; {
*..: . 8._ :

c. = n na m. :2 :- 2_...

-2 12 00 i=..=. '. _=. k-..-:.==..=i.t.,..--e.=.=.===..'a=.=.l.....~..==..=..=....=.:===9.

~s y c:

=j a l ::. :;;a..:::. :===::::2 7.--.r.-aa n a, ::: 1. :. u -_ _ '..'== = : '*--- l' C

l-2'2' . ; :. :c. _;.: n: ... t ... _ _ =..,. -. _. _.... -. _.

.~.w a ar.

p =a = = := :=a = =.-- = =-- r.=- d,= = =,=;,_.~ y ; g =;,;_;= h =-u _ :;:;-__ _;__=.-! gn -f u..._. in =3iE"N Ei~! 25! 'iui:- J=isI'i 41 -!:." / liil'}; O-1 n I 5"=55f-! E=:=.ei.s..=_..4.:._.i..i. =;=..i.iz: i:2iis.. !.'. <:.iirl.=if. :- = =. ie.:.=. u=;.E.ij 1000 r 2== :r- =.=-

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

__,==.=====..=. :=;_ _; : :. :,=.. c.n/==iiMrti inie=i- = ;-

., = -

= =- :.- = =.. : -- - x.

EiEN=u + =l=-ii>ii= =iri=F=ii.p:i =-2}i-iii iis!!;_i.:=i:4.ji.. cj : "pl.V-. s; setj=j ;.4:. : ri.d-! t 2 lig==. :fre ;a.iic :.i:: iii -:=.1 i.i),/ ;ii=r iii-j:. n = ;3.;;;. 3gg =Md"==ii.E:!= ;=:-k;;" i:!;. t= d - -!r=*i'i',:t i ii"+i+ i4n - [- u;;;;; u .. q.._ i.a :.... + . r. til .iiiiiic :. = ;: = y=... -i_n_.. - - .- it; p;..;. ! - !~-e+ 600 "iiii 'SU i ~ =l: <_... j. i I~1 .ii:!:ml ' .i.. 4 ..... -] a r s..p., g. -. /?. l' - [. ag. :. ". I'

: n 1 -.$.@-~~

././-- ~-i.'.l. u t.. t . F 400 -.ai f p . J:

p _.j:...

g \\, ..q.._'.. g. ._.:I.. u :l. u.... - _ \\.$. _. \\ .. j. ..@oS / .i.. .... = _ .u. i' 200 i. -- 0 4'\\ - b@,... .2 I l. [ s.. o o i . ~l - - sDv-o ..a w I - [. ..l : - .l. - : -

s..

~l. l 5' kN* b$ l e 0. i-3 V-7 I . NOV 3 01981 soO -100 o 1 200 TEMfER ATUM. 6) 00 i J

.~ FIGU?.E 3.c.-6 I s .._.....t..... l. j I 2!.00 I " c4.s....e 0r tm... u....a. ~m. ~~ . n. ~u~s 4.1 .j TOT. liYDROSTATIC TE5 3 i. _r..__s s.__.L c.. i. 4 s..... n 2.,C0 ETp r = -30 7 g' -{' r. _ J._ '_. - -i) +- ii. E-j H i _..a._. i%8 L._ j i l !-~ ~;-~. ' " ~ ".- ~ 1 ,f ji . ~ - - - 2000 c-r .r .- h. i ..:u.c... - [. j. j . ;.. :g v~ -O.;:j. .7 .up. !f yi I "" N" 9 Ni i ' F "" I 1800 I .435

t4 2::- 25.lii -

Up: 2": i J ~-[' .b-. 2- - Ili ~.. nri.;i:=, g=.. i;.24.:.;,,..... . g 2.._- c ;_. :,;. i....

t__,

I: -- j. :3.-.l.i i}ii: ;]i:- - -..j:. lu:'i-i :. i ' '- - - -li :,- :'::: Q,,-. : j :J n- :.r :aan e.cn.:: - - : _: :.n-....... r- .... e : ::: = n e-:!.:.. _J. 1600 ' :~~ ... ~ ' '. ~ _.: . ;.hj..: 'u ::'. & = M' =;.[~=.i-

=~.

t:_.__-!.-

.!!s4'4= dei;$la ~;JO.M. =' ;i-+1g :: j I P:iuni::Mi.-d!

, 'i~ilN.=".;.3 : "rhi;i;' N "*:. i~:f-- i!!..::/5:1751".i4 - _ Ej q 1J.00

.in.71.
Leili.c=.: p ' i.-P.-i ~ '.=== 1

.j /.i: : -40.1:.ii. m s. .=4 -. q... _ r. --. .:._. r,.___-a.: n.=..:. . : n.. -.

==. _.:=.... l. m ... __, e. i -4_;i:-- i - :.::.'.;_ r_ ;..x.._:.a.;. .-f g. :

_ _u.: ;t;=.:_=

E s=}i; j:i:jn:..: q.: .ii.. :.:. i a:.. :. - :.i!:..: y;p:_.;:jj m_ Ei i*f;i' Nr : p t r

t.

- 4 * :e --Jh. = ~" *'i M T 12 ..... "......E--/ rr :-- M 1200 U g _.... &=....:.l=..--..--tr : -===!=ir =::==Li a, / k. ! ~.:: r. : ~.a-l.. r.r _. _. l... .. g. - %. ::-,;;- sg n, EEfr-i===r t.i. 0-oi c.-7 O

.*n c.._n:.- n_ :.._nz.n.... g.

._2 c.g ;, f, m.,o l 2.Ci iC i;." "'tiyp5, n:_y.jF ' [ l,

h. @ t n-{:p( 'il 1000

. _. n.. _ : n.. =. _..m.. n_:n.=...i..._..... j. _........:n.e.g-u n. p.u... r: -p _.: =.=. :. :. :..:... :, __}, . e rrn.. __;g:p.g3 u [:.==gn.-

.npptr_;;j..: :

Jiniii s-hi :12il7"te-i rl-fi.2

i.!! r *n!":n-

a i

l 800 i : := ::.' : ; 1-i.~~

n-"

il' 2.:e --.[:. ".l.:i.: l'-.

..t

"-((-

/ :-i-

1:: tin:l ; - :l g ; Jl un; ::

- - :.t.2 : r ....3 t. ......:n. t.n. r jii! t :.j : . j..

l

-- b+ r i" i.. ' "t. 600 ~ j' i.. i i.:..l -i-

1

.j i:i ' -- - l t----- r--r -

I

~ .l.. .j.... .. [.. g .i .a i ._ 4.. q...; _a _.a... _..,.J_.__. --8 l I g ,.1 . - s .t 400 l ._ a. _.. _:_.. _. - us:-1._a. ..._...:.2. i l.- .a..... ,r ! I i: !m : 'i i e ( ..l ..s t 200. ---a. ~ - + - a..p.

_ _.. ; _... ;...... _ !. ;.... _j..: .

9 .. 4.:. .I l. .... i :_.. :. _.._. __:n ._. ;.. _.aa _ _.i... A fr. J / i i i 0 M-* l NOV 3 01981 3co 2oo 200 TmeYwruutr$00

FIGURE 3.4-7 l t-4 8-

I.

j-a 2400 - BEGI$'C.*:0 OF LIFE C00LD0 !'; LI.MITS -

Tv^ r'. 'rT.. D'.'...<. 2.~.'I C 1 r S. S

.i

~!- Bn.S D 0...._.D . AL c. r s :. -20,r

.l:.

n. -.

= ..ia 2200,- i

...n a

.s .,. l l-._. _.. _ _.. _ .-l .:;..; _..__2_..l.. _ ::p

i.

. _... ___ 3 2. : 2000 - 4 ./ i i .. _" i. ..r..... v ,.g ..g.., __...__"l 1800 _'" ,cr - ~

't + ' -

~ -- f

F "-" '

I;'ih r..l.l. ':...i 9.

.: ...J

. -J -.:i , ::9:, "I 7: :n.!.i q. < ! -- - -i : ";;;!-i,+:i-ig:3.:;[ - " - [...

i;:.2 :.i:!

7 u [;._:; ; niin.;; %C ..._.a::...:..x .. c. -. ;, :.u:........ 1600 ' :."--!:Mi:e:.. '-it :lj_ji;E - NNi.I:l --fl iiiiii[-[hlt, ; ti Eisii4==!:=hipt-==i=.i =i =:=M=hi -i=#4=4=+- E-l 4 =. i ch- _.je rir .:9 -- /. ... -1.; i ?/ =".. 1."J; .ii.; iiff.. 3 ri.:

"nr;'=n-i

"' " mi =f = =: ~ :Cr : i ir: 33;r I

^ 1400 so

__ ' ~~!. iani, :i.;-d Y

isN;;.:.1-i.-4;

..i:4+:rp-I

r

. : ; u.. :.. -.j c. _......- - - 3. =; 2:.=.. =;;.:; u..j.. _. :.:.. = -- 4 v ". _.:i_f _.".,. ? :.+r "---i - -.-!. n:.... .. :.-!..i. i..*:r.:..h. -j W g g 1200 ... :,..: a;.r....... - ;... ;...,. us . 1 g 2_.;..g: ua_ a= .:n..g l . _ _.1.. -. -.. u. _=....:.t. r _.._-.: ::..uz..,. ;. -. h+ - - :. : ; =-== :23 w ~ ' - - ' - ' - - ~ ' ~ * ' ' * ~ ~ g A* .".~]iii -y " --i ::::lih::i :.f.- !!- F-!= ' ':;iiiiii.ip: =

::.1. s.._..

=. :=:+.....i..

s = : - -i-.- - F.. n I......

===. ....1 --fii !!.! :-- !. /.09i : ' 8. in-f::id::t:iiri: =. :.::.a..u:.:-i 1000 T=.=g-' Ej +21::9;i ~i i-ji: - N!- / j F53 -(p : :.r.;--.-= :r - ' "I :: i.~. i_.. ::0!:fi.n-?i :: .n. !;& F.,,. - ! E ;i."li :.X l E--E5 ' iia }ii ir.tri- ~? 7d * *-. *], li - i:htsli --l i:N.:.iu nt - .....y.. -/ :1 .. :=:l.: r i:.:: : 9.:=:

:..=

800

j: iluiij.: - i j e -- i-i. N :-

g . f / _;.i-J: -.ij e i:!. . r.. i,.: : :-- j.... =... ;=... Fi.

n. i n. : =,9_.

,.i.. I. - .r .. 4

t.

5

:..::L a.

- - - - ~ 2.._.4. 9.. /.. . y. ..7!- -l-600 .i +': /.i -, - - i-i..: ,..O(_

_. _; - )..

8 i).. 4 : ! nl... I l fil tO j

l O

.,-e - t --- - j---j ' 400,. .1 .r., a z-e .i t ~. g._l[g<1

...: ...j.

9.: :.._.__.j p.4.+_, t I -.8..-~~-.. -. -g ~

l.: ._.l 8

7.: -" e . :..l.. 1 200 i .l.. 3 t .i ' g. .) ...q.... )._ _- l. _g . I..... l."q. :. . !.d... s......:..!...... l .g a i .... j. __..... j o NOV 3 01981 ~ -200 00 0 100 200 Amendment No. 31 3.4.-1 TEMPERgTUIE OF)

3.5 CHEMICAL AND VOLLE CONTROL SYSTEM Acolicability: Applies to the operational status of the chemical and volume control systeni when there is fuel in the reactor and the Reactor Coolant System is not in a cold shutdown baron concentration. Objective: To specify those limiting conditions for operation of the chemical and volume control system which must be met in order to ensure adeouate boration capability is available. Soecification: A. Whenever there is fuel in the reactor and the Primary Coolant System is not at cold shutdown boron concentration the boric acid storage tank or the refueling water storage tank shall contain sufficient boric acid solution to bring the Reactor Coolant System to the cold shutdown baron concentration. Solution temperatures shall be maintained at least 100F above the concentration saturation temperature but not less than 400F. 8. Whenever there is fuel in the reactor there shall be at least one operable path for boron injection consisting of system pumps, piping, heat tracing, valves, instrumentation and controls' operable as to assure the capability of boron injection at a rate in excess of 250,000 ppm-gals / min. Into the reactor coolant system. C. Whenever the reactor is critical there shall be at least two independent operable paths each meeting the recuirements of B above. Exception: The reouirements may be modified during operation with an isolated loop to permit operation with one operable flow path for a period not to exceed 24 hours. Remedial Action: With the reactor critical and only one of the boron

injection flow paths recuired in C above OPERABLE, restore at least two baron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours.

~ Basis: The chemical and volume control system provides control of the reactor coolant baron inventory. Reduction of concentration is accomplished by dilution with unborated primary grade water or by baron removal through ion exchange. An increase in concentratton may be accomplished by using either of the two operable charging pumps which have separate suction lines from the refueling water storage tank. An increase may also be accomplished using the auxiliary charging pump taking suction from the boric acid storage tank. Each of the three operable pumps can be lined up to discharge into the reactor coalant system through a separate flow path. Thus there are three operable flow paths normally available during operation. However, during periods of two loop operation, the loop fill header and the auxiliary charging pump may not be available, reducing the 3.5-1 NOV 3 01981

number of available flow paths to two. The exception provides time to restore redundancy should one flow path become inoperative. The rate specified is adecuate to bring the reactor to a cold shutdown condition. It precludes the possiblity of the lower capacity auxiliary charging pump and the lower concentration refueling water storage tank being taken together as an available flow path. The allowable out-of-service periods ensure that minor component repair or corrective action may be completed widhout undue risk to overall facility safety from injection system failures during the repair period. e l l l l I l 3.5-2 NOV 3 01981

3.6 EMERGEbrY CORE COOLING APO CONTAIbMENT SPRA SYSTEMS Acolicability: Applies to the operating status of the emergency core cooling and containment spray systems. Cb iective: To define the conditions under which components of the emergency core cooling and containment spray systems must be operable. Soecification: A. The following equipment must be operable wher ever the reactor coolant system temperature and pressure exceed 2100F and 400 psig: 1. Two safety injection tanks set for automatic initiation. Each tank shall contain 11,200 + 500 gallons of water boratea to at least 1720 ppm and pressurTzed with nitrogen to 230 psig + 10 psi, - 25 psi. 2. One operable ECCS train consisting of the following subsystems of

the train. Each subsystem includes the manual valves that are aligned and locked in the position required for safeguards operation, the automatically operated valves set for automatic operation or aligned and locked in the position required for safeguards operation, the controls set for automatic initiation where appropriate, and a pump powered from an engineered safeguards bus.

a. One service water pump subsystem b. One component cooling pump subsystem c. One low pressure safety injection pump subsystem d. One high pressure safety injection pump subsystem f. One containment spray pump and RHR heat exchanger subsystem 3. Station service power in accordance with Technical Specification 3.12.A supplying the same operable ECCS train as in (2) above. 4. The refueling water storage tank filled and available in accordance with Technical Specification 3.7. 5. The fill header motor operated root valves to two non-isolated loops. Exception: The recuir ments may be modified with regard to the position of controls and valves during periods of hydrostatic testing. Remedial Action: Restore recuired limiting condition within four hours then follow Spec. 3.0.A. 3.6-1 ZOV 3 01981

B. Whenever the reactor coolant system baron concentration is less than ** that reouired for Hot Shutdown condition, two high pressure safety injection pump subsystems shall be operable. Exceation: See Specification C, Exception 1. C. The following eculpment must be operable whenever the reactor is in a power operation condition. 1. Three safety injection tanks set for automatic initiation and subject to the conditions specified in A.1 above. 2. Two operable and redundant ECCS trains, each train consisting of the subsystems specified in A.2 above. 3. Station service power in accordance with Technical Specification 3.12.B. 4. The refueling water storage tank and the spray chemical addition tank filled and available in accordance with Technical Specification 3.7. 5. The fill header motor operated root valves to three non-isolated loops. Exceptions: 1. If any of the component subsystems specified in B and C.2 above becomes inoperable continued power operations is permitted for a maximum of 72 hours provided the component subsystem performing the same function in the other train are operable. In this situation the operable subsystem and its diesel generator shall be tested within two hours after discovery of the outage. 2. If any of the fill header motor operated root valves becomes inoperable continued power operation is permitted for a maximum of 72 hours provided both of the remaining root valves are tested

operable within two hours after discovery of the outage.

3. One safety injection tank may be isolated for a period not to exceed one hour.' 4. If one of the safety injection tanks is found not -to be within '+ specifications it shall be restored to specification within four hours. Remedial Action: Restore recuired limiting condition within grace period specified then follow Specification 3.0. A. Basis: Adecuate core cooling and containment spray is provided for the entire break spectrum up to and including the design basis accident. This protection covers all modes of operation from shutdown to full power. 3.6-2 NOV 3 01981

At full power minimum reouired safety injection includes three (3) operable safety injection tanks, and two complete ECCS trains consisting of the subsystems specified in A.2. The accident analysis considers that only 2/3 of the capacity of the operable eouipment is effective for core cooling. Containment peak accident pressure is maintained below design pressure and subsecuent containment cooling recuirements are adeouate if one of the two containment spray pumps is operable. Specification A provides a pressure and temperature limit above which ECCS must be operable. It recognizes the greatly decreased probability of a loss of coolant accident and the negligible amount of energy stored in the primary coolant. Specification B ensures that a sufficient Quantity of boron can be injected by the ECCS to maintain the reactor subcritical following the most limiting main steam line break accident with the concurrent failures ** of the highest worth CEA stuck out'of the core and the failure of one ECCS train to function. 3.6-3 NOV 3 01981

3.7 BORON AND SODIlN HYDROXIDE AVAILABLE FOR THE CONTAIPNENT SPRAY SYSTEM Acolicability: Applies to the concentration and volume inventory of barated water and spray chemical-vater. Objective: To ensure the availability of barated water for baron injection, core cooling and containment spray and the availability of sodium hydroxide solution for iodine absorption. O Specification: A. Whenever the core cooling or containment spray systems are specified to be operable, the refueling water tank shall contain not less than 300,000 gallons, and shall have a boron concentration of between 1720 and 1900 ppm, the spray chemical addition tank shall contain not less than 15,400 gallons of sodium hydroxide solution at a concentration of between 8 and 11 percent.. Solution temperature shall be maintained at least 100F above th,e concentration saturation temperature but not.less than 400F. B. Whenever the core cooling or containment spray systems are specified to be operable, the total boron available for mixing in the containment sump shall be limited accordirg to the following ecuation: C M +C M +C,M3 is less than or ecual to 1890 ppm 1 1 99 M +M +M +M4 1 2. 3 where C1 boron concentration in refueling water storage tank, = ppm C2 baron concentration in reactor coolant system, ppm = C3 average boron concentration in safety injection = tanks, ppm minimum mass of licuid transferred from refueling M1 = water storage tank = 1.67'x 106 lbs M2 mass of licuid in reactor coolant system = 4.7 x = 105 lbs mass of licuid in safety injection tanks, lbs M3 = minimum mass of licuid transferred from spray M4 = d,emical tank = 7.5 x 104 lbs. C. Whenever the refueling water storage tank is specified to be operable for baron injection it shall contain sufficient water, at a minimum baron concentration of not less than 1720 ppm, to bring the reactor coolant system to a cold shutdown boron concentration. 3.7-1 NOV 3 01981

Remedial Action: If A, B or C above are not met, restore canpliance with

the LCO within twenty-four hours then follow Speci?ication 3.0. A.

Basis: The 300,000 gallons in the refueling water storage tank is based on allowing a minimum of 200,000 gallons to be transferred to the containment via spray and core cooling before recirculation is manually established. Automatic transfer to recirculation will occur after at least 200,000 gallons has been transferred from the tank leaving a minimum of 100,000 gallons which will insure adecuate NPSH recuirements for the engineered safeguards pumps. The concentration of 1720 ppm is the highest value used in any of the safety analyses. By specifying this concentration the safety of the plant shown in Section 14 of the FSAR is assured. Analysis of loss-of-coolant incidents shows that 200,000 gallons will be sufficient to limit core temperatures and containment pressure for the full spectrum of pipe, ruptures. These analyses are discussed in Section 14.14 of the FSAR. The 15,400 gallons of sodium hydroxide solution is based on hydrostatistically balancing a full refueling water storage tank. The minimum and maximum sodium hydroxide and boron concentrations are based on maintaining the pH of the initial spray solution; the sump water at the start of recirculation, between 8.5 and 11. This will assure that the containment spray system will effectively remove iodine from the containment atmosphere. The twenty-four hour grace period is necessary in order to allow time to obtain representative sample and/or to achieve recirculation. 3.7-2 NOV 3 01981

3.8 REACTOR CORE ENERGY RB40 VAL Apolicability: Applies to the operating status of plant components for removal of reactor core energy. Objective: To specify conditions of the plant eculpment necessary to ensure the capability to remove energy from the reactor core. Soecification: A. Whenever there is fuel in the reactor, at least one of the following cooling mechanisms shall be in operation with a second mechanism operable: 1. RHR Train A 2. RHR Train B 3. Steam Generator No. 1 4. Steam Generator No. 2 5. Steam Generator No. 3 6. A minimum of 23 feet of water above the top of the core with the reactor head removed. Exceotions: 1. The RHR$ may be secured for a period not to exceed six hours to facilitate special maintenance, refueling functions or tests. During such periods reactor coolant temperatures shall be continuously monitored and initiation of core cooling shall be continuously available. 2. For purposes of inservice inspection testing, the RHRS may be secured provided that reactor coolant temperature is continuously monitored and two cooling mechanisms are continuously available. B. The following conditions must be met for a steam generator to be considered operable for decay heat removal. 1. The reactor coolant system must be closed and pressurized to 100 psi above saturation pressure. 2. The steam generator must have both the cold and hot leg stop valves fully open. 3. The steam generator water level must be above the top of the tube bundle. / 3.8-1 NOV 3 01981

1 4 An inventory of cver 100,000 gallons of primary grade feedwater must be available. 5. A feed pump must be operable. C. The steam generators shall be demonstrated operable in accordance with specification 4.10 before the reactor coolant system T. Ave. can be increased above 2100F. D. The reactor shall not be in a power operation condition which generates steam at a rate in excess of the on-line steam generator relieving capacity 1n' accordance with figure 3.8-1. E. The reactor shall not be maintained in a power operating condition unless the following conditions are met to assure post shutdown heat removal capability. 1. Two steam generator auxiliary feed pumps are operable. 2. An inventory of over 100,000 gallons of primary grade feedwater is available. Exception: If either steam generator auxiliary feed pump becomes inoperable continued power operation is permitted for a maximum of seven days. In this s!.tuation the operable feed pump is to be tested once a day. Basis: Specification A assures that. decay heat removal capability is always available. A single steam generator is capable of removing core decay heat by natural or forced circulation provided the conditions specified in B are met. A single cooling mechanism is sufficient to remove decay heat but single failure considerations recuire that two mechanisms be OPERABLE. Specification C assures the structural integrity of the ' steam generator tubes which are a fission product barrier. Specification D assures suffluent relieving capacity during either two loop or three loop power operation. l A reactor shutdown ~from power requires removal of core decay heat. Immediate decay heat removal recuirements are normally satisfied by the steam bypass to the condenser. Therefore, core decay heat can be continuously dissipated via the steam bypass to the condenser as long as feedwater to the steam generators is available. Normally, the capability to supply feedwater to the steam generators is provided by operation of the feedwater system. In the unlikely event of complete loss of electrical power to the station, decay heat removal is by steam discharge to the atmosphere via the main steam safety valves or the atmospheric steam dump valve. Either steam generator auxiliary feed pump can supply sufficient feedwater for removal of decay heat from the plant. 3.8-2 , NOV 3 01981

F2' o< ,m + g 1 O \\ G 9 100 5 ~ / 15 b-j 90 g 14 o h 13 1 6u_ g, l W $' 12 %RATEDPOWER m J g 11 a -i vs.

70

{ g '0 ~ ~, STENi GENERATOR SAFELY VALVE ,) w r. 60

a

/ RELIEVING CAPACITY l g 9 f 0 ~ o 8 50 w :s i m W a, s < 40 m i 30 .s. 20 114ERE SHALL PE A filNIf1JM OF THREE ON-LINE 10 SAFETY VALVES PER OPERATING STEAM GENERATOR i e i i i i i i i i 1 2 3 4 5 6 7 8 9 10 11 12 a 13 14 6 SRN1 CNJWOR S/FDY VALVd 10 IEVIIS CAPACITY (LB./HR. x 10 ) i j F I G 0 R E 3.8-1

3.9 OPERAliONAL SAFETY INSTRUMENTATION, CONTROL SYSTEMS, AND ACCIDENT MONITORING INSTRLNENTATION Apalicability: Applies to plant instrumentation system. Objective: To specify the conditions of the plant instrumentation and control systems ~ necessary to ensure reactor safety. Soecification: The operability of the plant instrument and control systems shall be in accordance with Tables 3.9-1, 3.9-2 and 3.9-3. A. Power operation shall be permitted to continue with the limits as stated in Table 3.9-1 column entitled " Minimum Operable Channels" except as conditioned by the column entitled " Bypass Conditions". Remedial Action: In the event that specification A above is not met,

the plant shall be placed in a hot shutdown condition within 6 hours.
8.

Whenever automatic initiation of engineered safeguards is recuired, the number of operable sensors shall not be less than the minimum specified in Table 3.9-2. Exception: One subsystem can be removed from service during periods of maintenance or on-line testing for a period of 24 hours. Remedial Action: In the event that specification ', including the exception, is not met, the plant shall be placed in a hot shutdown condition witi in 6 hours. C. Whenever the reactor is at power the minimum Accident Monitoring Instrumentation listed in Table 3.9-3 shall be operable. Remedial Action: In the event the nucher of operable accident monitoring instrumentation channels falls below the Minimum Channels Operable recuirements in Table 3.9-3, either restore the inoperable channel (s) to operable status within 48 hours or be in at least hot shutdown condition in the next 12 hours. Basis: Reactor safety is assured by the instrumentation channels, logic circuitry, trip modules, and other eculpment :.ecessary in the reactor protective system. Selected nuclear steam supply system conditions are monitcred and a rapid reactor shutdown is initiated if any one or a combination of conditiuns deviates from a pre-selected range. This system automatically initiates appropriate action to prevent exceeding esteblished safety limits. Safety is not compromised by continuing operation with certain instrumentation channels or initiation circuit 4t 3.9-1 NOV 3 0198f

of service since provisions were made for this in the plant design. This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the reactor protection system when any one or more of the channels or circuits are out of service. In the rcactor protective system, four independent and redundant channels monitor each safety parameter. If any one of the four channels deviates from a pre-selected range, a trip signal is initiated. For any safety parameter, a trip signal from any two of the four protective channels will cause a reactor trip. If one of the four channels is taken out of service for maintenance, the protective system for that parameter is changed to a two out of three coincidence for a reactor trip by bypassing the removed channel. When a second channel is taken out of service, the trip module for that channel is placed in the trip mode, and the resultant logic for that parameter is one out of two. Thus, with one or two channels removed from service for that parameter, protective action is initiated when recuired and the effectiveness of the reactor protection system is retained. The operating recuirements for the reactor protective system are shown in Table 3.9-1. Redundant sensors and logic are provided for i.he initiation of all engineered safeguards systems. In both the contairment isolation and containment spray systems, two identical subsystems are used in each system. In the safety injection actuation systems diverse sensors are used for the initiation of two identical subsystems. Each of these three engineered safeguards systems may be operated as shown in Table 3.9-2 without jeopardizing safeguards initiation. One subsystem may be removed from service for a limited time for purposes of mairitenance or testing. Although no credit is taken for the high rate-of-change-of-power channel in the Maine Yankee accident analysis, operebility of this channel at lov power levels provides back up assurance against excessive power rate increases. Temperature fee &ack effects protect against excessive power rate increases at higher power levels. f The minimum number of operable channels for the accident monitoring instrumentation is given in Table 3.9-3. The accident monitoring instrumentation is used to evaluate and aid in mitigating the consecuences of an accident. l 3.9-2 7 "301B85 L

TABLE 3.9-1 Instrumentation Operatina Reouirements for Reactor Protective Sys Big Minimum Operable No. Functional Unit Channels (a) Bypass Conditions 1 Manual (trip buttons) 1 set Nona i 2 High Rate-of-Change Power 2(c) Below 10-4% and Above 10% of Rated Power (b) 3 High Power Level 2(c) None 4 Thermal Margin / Low 2(c) Below 10% of Rated Power (b) Pressurizer Pressure 5 High Pressurizer Pressure 2(c) None 6 Low Reactor Coolant Flow 2(c) Below 2% of Rated Power (b) 7 Low Steam Generator Water Level 2(c) None 8 Low Steam Generator Pressure 2(c) 190 psi Above the Trip Setpoint 9 High Containment Pressure 2(c) None 10 Axial Flax Offset 2(c) Below 15% of Rated Power (b) (a) The minimum degree of redundancy is one, except for manual trip which has a minimum degree of redundancy of zero. l (b) As indicated on Nuclear Instrumentation Channels. 1 I (c) Providing one of the inoperable channels is placed in the trip positions, otherwise 3 channels is a minimum. 3.9-3 9N 3 01981

TA81.E 3.9-2 Instrumentation Operating Reouirements for Engineered Safeguards Systems Init'ation Minimum Operable Sensors Bypass i No. Functional Unit Per Subsystem Canditions Set Points 1 Safety Injection: A. Manual 1 8. High Containment Pressure 2(a) less than 5 i psig C. Low Pressurizer Pressure 2(a) greater than 1585 psig 2, Containment Spray: 1 A. Manual 1 B. High Containment Pressure 2/ set (b) 1ess than 20 psig 3 Containment Icolation: A. Manual 1 8. Containment High Pressure 2/ set (b) 1ess than 5 psig (e) Subsystem initiated by two out of four pressure sensor dhannels. The minimum degree of redundancy is one. (b) Each subsystem is initiated by two out of three pressure sensors. The minimum degree of redundancy in each subsystem is one. Reactor coolant pressure less than 1685 psig. 3.9-4 NOV 3 01981 g 3 --,. - - - ,m-g ---9 g-ryn------'%9 g, v- .--c

TABLE 3.9-3 Accident Monitoring Instrumentation Instrument Minimum Channels Operable 1. Pressurizer Water Level 1 2. Auxiliary Feedwater Flow Rate 1perSteamdenerator 3. Reactor Coolant System 1 Subcooling Margin Monitor 4. PORV Position Indicator 1/ valve j ( Acoustic Flow Sensor) 5. Safety Valve Position Indicator 1 ( Acoustic Flow Sensor) f NOV 3 01981

3.10 GA CROUP, POWER DISTRIBUTION, MODERATOR TEMPERATURE COEFFICIENT LIMITS AND COOLANT COPOITIONS Applicability: Applies to insertion of CEA groups and peak linear heat rate during operation. Cbjective: To ensure (1) core suberiticality after a reactor trip, (2) limited potential reactivity insertions from a hypothetical CEA ejection, and (3) an acceptable core power distribution, moderator temperature coefficient, core inlet temperature, and reactor coolant system pressure during power operation. Specification: A. CEA Insertion Limits 1. When the reactor is critical, except for physics tests and CEA exercises, the shutdown CEA's (Groups A, B and C) shall be fully withdrawn and the regulating CEAS (groups 1 through 5) shall be no further inserted than the limits shown in Figure 3.10-1 for 3 loop operation. 2. CEA's shall be considered fully withdrawn when positioned such that: the rods are inserted to 4 steps from their upper electrical ** a. limit when the RCS baron concentration is greater than 100 ppm or b. the rods are at their upper electrical limit when the RCS baron concentration is less then or eaual to 100 ppm. 3. When the reactor is critical, the shutdown margin with one CEA stuck out will not be less then that shown in Figure 3.10-7. During low power physics testing at the beginning of a cycle, CEA insertion is permitted such that the minimum shutdown margin is no less than 2% in reactivity. 4. Operation of the CEA's in the automatic mode is not permitted. B. Power Distribution Limits 1. The peak linear heat rate with appropriate consideration of normal flux peaking, measurement-calculational uncertainty (8%), engineering factor (3%), increase in linear heat rate due to axial fuel densification and thermal expansion (0.3% for Types E, G, H & I only) and power measurement uncertainty (2%) shall not exceed: 3.10-1 "0V 3 01981 f

Fresh Fuel 13.5 kw/ft X greater than 0.50 and CAB L less than or eaual to 792 MWD /MTU 14 kw/ft X_ greater than 0.50 and CAB L greater than 792 MWD /MTU 16 kw/ft X_ less than or ecual to 0.50 L Exposed Fuel: 14.0 kw/ft X_ greater than 0.50 L 16.0 kw/ft X_ less than or eaual to 0.50 L where X_ is fraction of core height and CA8 is cycle L average burnup. Should any of these limits be exceeded, immediate action will be taken to restore the linear heat rate to within the appropriate limit specified above. 2. The total radial peaking factor, defined as Fd = F$ (1 + To), shall be evaluated at least once a month during power operation above 50% of rated full power. 2.1Fkisthelatestavailableunroddedradialpeak determined from the incore monitoring system for a condition where all CEAs are at or above the 100% power insertion limit. To is given by the following expression: To=2 (Pa-Pc)2 + (Pb-Pd)2 Pc+Pd)2 y (Pa+Pb + Pi = relative cuandrant power determined from incore system for cuandrant i, when the incore system is operable and by Specification 3.10.8.4 otherwise. 2.2 If the measured value of F% exceeds the value given in Figure 3.10-4, perform one of the following within 24 hours: l 1. Reduce symmetric offset pre-trip alarm and trip band (Figure 2.1-2), thermal margin / low pressure trip limit (Figure 2.1-1 and Tech. Spec. 2.1), and excore LOCA monitoring limit (Figure 3.10-3) by a factor: t > FE measured Fh(Figure 3.10-4) or 2. Reduce THERMAL PCWER at a rate of at least 1%/ hour to bring the combination of THERMAL power and % increase in 3.10-2 NOV 3 01981 t

i Fk to within the limits of Figure 3.10-5, while maintaining CEA's at or above the 100% power insertion limit; or 3. Be in at least HOT STAND 8Y. 3. Incore detector alarms shall be set at least weekly Alarms will be based on the latest power distribution obtained, so that the peak linear heat rate does not exceed the linear heat rate limit defined in Specification 3.10.8.1. If four or more coincident alarms are received, the validity of the alarms shall be imediately determined and, if valir', power shall be immediately decreased below the alarm setpoint. 3.1 If the incore monitoring system becomes inoperable, perform one of the following within 4 E.F.P.H. a. Initiate a power reduction to less than or ecual tofP at a rate of at least 1%/ hour where P (% of rated Power) is given by: P = 0.85 (Linear heat rate permitted by Specification 3.10.B.1) x 100 Latest measured peak linear heat rate corrected to 100% Power while maintaining CEA's above the 100% power insertion limit and monitor symmetric offset once a shift to insure that it remains within + 0.05 of the value measured at the time when the above equatinn is evaluated. This procedure may be e,tployed for up to 2 effective full power weeks, or 2. Comply with the alarm band given in Figure 3.10-3. If a power reduction is recuired, reduce power 'at a rate of at least ]%/ hour. 4. The azimuthal power tilt, To, shall be determined prior to operation above 50% of full rated power after each refueling and at least once per day during operation above 50% of full rated power. To is given by the following expression: To = 2 (Da-Dc)2 + (Ob-Dd)2 3 (Da + Cb + De + Dd)2 Di = signal from excore detector channel 1. To shall not exceed 0.03. 4.1 If the measured value of TQ is greater than 0.03 but less than or ecual to 0.10, or an excore channel is inoperable, assure that the total radial peaking factor (Fk) is within the provisions of Specification 3.10.8.2 once per shift. 4.2 If the measured value of To is greater than 0.10, operation may proceed for up to 4 hours as long as Fk is maintained within the provisions of Specification 3.10.8.2. Subsequent 3.10-3 NOV 3 01981

operation for the purpose of measurement and to identify the cause of the tilt is allowable provided: 1. The THERMAL POWER level is restricted to less than or equal to 20% of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination, and 2. Reduce setpoints in accordance with Specification 3.10.B.2.2. 5. The incore detector system shall be used to confirm power distribution, such that the peaking assumed in the safety analysis is not exceeded, after initial fuel loading and after each fuel reloading, prior to operation of the plant at 50% of rated power. 6. If the core is operating above 50% of rated power with one excore nuclear channel out of service, then the azimuthal power tilt shall be determined once per shift by at least one of the following means: a. Neutron detectors (at least 2 locations per cuandrant), b. Core-exit U1ermoccuples (at least 2 thermocouples per cuandrant). 7. The pre-trip limits of Figure 2.1-2 constitute Limiting Conditions of Operation. C. CEA Drop Times 1. At operating temperature and 3 pump flow, the requirement for the maximum drop time of each CEA shall be not greater than 2.7 seconds from the time the holding coil is de-energized until the rod reaches 90% of its full insertion. D. Misaligned, Inoperable, Slow or Dropped CEA 1. A CEA is considered misaligned if it is out of position from the ( reminder of the bank by more than 8 inches. Every 24 hours, except during physics tests and CEA exercies, if a CEA is misaligned, linear heat rate and total radial peaking factors must be shown to be within design limits as specified in 3.10.8.1 and 3.10.8.3 using the latest unrodded radial peaking factors. If the CEA deviation alarms from both the computer pulse counting system and the reed switch indication system are not available, individual CEA positions shall be logged and misalignment checked every 4 hours. j 2. A full length CEA is considered inoperable if it cannot be i tripped. A CEA that cannot be driven in shall be assumed not able l to be tripped until it is proven that it can be tripped. No more l than one inoperable CEA is permitted during power operation, except during physics testing, or CEA exercises. The shutdown 3.10-4 lNOV 3 01981 ,-,,-.p ,-e-,- -, a w w-- , ~

~. t margin limitation specified in 3.10. A.3 must also be met by enough [ boration to compensate, if necessary for the inoperable CEA within i 2 hours. 3.. A CEA is considered to be a slow CEA if it does not meet the drop time reouirement. Should a CEA cxceed the required drop time, then the shutdown margin limitation specified in Specification 3.10. A.3 must be met by enough boration to compensate, if necessary, for the eouivalent of 1.5 times the negative reactivity insertion which is delivered after 2.5 seconds. 4. Except during physics testing, in the event of a dropped or misaligned CEA which cannot be corrected withJn 4 hours of its identification: a. The remainder of the rods in its group will be aligned within 8 inches of the misaligned or dropped CEA while maintaining i the allowable CEA sequence. b. Following realignment, the peak linear heat rate will be shown to be within the limit specified in 3.10.8.1 and the total radial peaking factor will be shown to be within the limit specified in 3.10.B.3 using the latest unrodded radial peaking factor. E. Moderator Temperature Coefficient (MTC) shall be: 1. Less positive than 0.5 x 10-4 delta rho /oF whenever Thermal power is less than or equal to 70% of Rated Thermal Power. 2. Less positive than 0.0 whenever Ulermal Power is greater than 70% of Ratej lnermal Power. i F. Coolant Conditions 1. The reactor coolant pressure and the reactor coolant temperature at the inlet to the reactor vessel shall be maintained within the limits of Figure 3.10.6 under steady-state 3 loop operation. l 2. The reactor coolart flow rate shall be maintained at or more than a nominal value of 360,000 gpm (indicated) during steady-state 100% power operation. Basi s_: The CEA insertion limit shown in Figure 3.10-1 assures that the individual CEA worths used for the CEA ejection analyses are not exceeded. The CEA insertions used for the CEA withdrawal accident are also not exceeded by this insertion limit. In addition, the limit ensures that the reactor can be brought to a safe hot shutdown condition even with the highest worth CEA not inserted. This restriction provides more shutdown margin than is recuired ht BOL, since the modertor temperature coefficient is more negative at EOL. For this regulating group insertion limit, the peak l l linear heat rate will be well within the design values. 3.10-5 00V.3 0.198(

The limit applies also to two loop operation, in which case the power coordinate is rescaled to 100% of the rated two loop power. This ensures that the CEA induced peaking will not lead to worse thermal conditions than for 3 loop operation since the flow to power ratio is greater for two loop operation. This CEA insertion limit may be revised on the basis of physics calculations and physics data obtained during plant startup and subseauent operation. Incore detector alarms are set based on the latest power distributions obtsined from incore detector analyses. These techniaues reflect actual rLdial and axial power distribution which exist in the core. Should the system become unavailable, continued operation is permitted under either the more conservative excore symmetric offset pretrip (alarm) envelope or at a power level consistent with maintaining a 15% margin to the peak linear heat rate assumed in the LOCA. Both these functions ensure that operation is within the limiting peak linear heat rates assumed as initial conditions for the Loss of Coolant Accident (LOCA). Further, since rod position information is not available to this excore system, this function l assumes the most limiting radial power distributions permitted at each j power level, t The split excore detectors monitor the axial component of the power distribution. The signal generated from the excore detectors is provided as input to both the Symmetric Offset and Thermal Margin / Low Pressure. Trip Systems. Limiting Safety System Settings (LSSS) are, therefore, gen'erated as a function of the excore detector response. The radial component of the power distribution is monitored as a Lin,iting Condition of Operation (LCO) by Tednical Specification 3.10.B.3. Therefore, the intent of Technical Specification 3.10.B.3 is to monitor the radial component of the ~ power distribution and to ensure that assumptions made in the generation of Reactor Protective System (RPS) LSSS remain valid. The LCO on the radial power distribution is specified in Fi steady-state unrodded total radial peak (Fk)gure 3.10-4 in the form of a and provides indication that the core power distribution is behaving as predicted. Figure 3.10-4 includes 10% for calculational uncertainties. The measured steady-state value of Fk, augmented by 8% for measurement uncertainty, is compared to this limit on a monthly basis. Should the measured steady-state unrodded total radial peak including uncertainties exceed the limit of Figure 3.10-4 at any time in the cycle specific action is to be taken to assure that the LSSS remain valid. The specific action includes a) the reductionofRPSLSSSandLCObytheratioofFkmeasured/F (Figure 3.10-4) to directly compensate for the higher radia peaks, or b) the imposition of additional restrictions on power and CEA position (Figure 3.10-5) to assure that the assumptions made in establishing the RPS LSSS and LCO remala valid. Figur? 3.10-5 in conjunction with restricted CEA insertion allows for an ir. crease in the steady-state unrodded total radial peak above the limits of 3.10-4 without a modification of the RPS LSSS. The allowed increase in radial peak is derived from the difference between the radial peaks assumed in the RPS setpoints for rodded conditions at reduced power and the radial peaks reflected in the CEA insertion limit at 100% power. This assures that the radial peaking factors vs. power assumed in the RPS LSSS remain valid. The power distribution in the core can be determined in two ways. The normal method is through analysis of the fixed and movable neutron detector signals with the on-line computer. The alternative is to NOV 3 01981

2 determine the radial and axial peaking factors by hand. The radial peaking factor can be determined from the core exit thermocouples, the fixed incore detectors or the movable incore detector traces. The axial peaking factor can be determined from the fixed inc-detectors, the movable incore detector traces or the excore de' The reouirement' 2. that the core power distribution be shown to be wit:iin the design limits after every refueling not only ensures that the reactor can be operated safely but will also provide reasonable verification that the core was properly loaded. The requirement for operability of incore instrumentation in the instance of an excore detector channel being cut of service ensures that an unobserved cuandrant power tilt will not occur. The maximum CEA drop time specified is consistent with the values used in the safety analysis. For a full length CEA, with misalignment up to 8 inches from.the remainder of the bank, the peaking factors will he we)) within design limits. If a CEA is misaligned, the peak linear heat rate will be shown to be within design limits every 24 hours. The 24 hour time limit is short with respect to the probability of an independent incident occurring. The recuirement that no more than one inoperable CEA is allowed and that the. shutdown margin is maintained ensures that the reactor can be brought to a safe shutdown condition at any time. Shutdown margin is assured within the required CEA drop time by conservatively barating to compensate for a slow CEA during operation, if necessary. CEA drop times, CEA core height vs. time, and CEA worth measurements are all made after initial loading and each refueling. Should a CEA drop time be in excess of 3.10.C.1, then the core height on that CEA at 2.5 seconds would be conservatively determined. Reactivity worth of tF-CEA from the above core height to the bottom of the core would the' :e determined. Sufficient boron would thereafter be added, if necessar) furing power operation to compensate for 1.5 times the above measured :.>..ctivity in order to maintain adecuate shutdown margin. The recuirement to align the dropped or misaligned CEA with the remainder of its bank assures thet operation will not be under conditions which violate the assumptions used in the generation of the RPS setpoints. The moderator temperature coefficient, coolant pressure, flow rate, and temperature specified are consistent with the value assumed in the safety analysis. The safety analysis assumes a maximum reactor inlet temperature of 5540F. The specified value includes 40F for temperature measurement uncertainties. 3.10-7 'NOV 3 01981

EH$ Ck aaa 6 0E5 rE< POWER LEVEL (% OF RATED POWER) VS. CEA WITilDRAWAL (STEPS) 00h 3 FOR TilREE LOOP OPERATION i C. " m OR (%. OF 0.63 x RATED POWER) FOR WO LOOP OPERATION O n p.n.....n..u .n"o-o.on" "" { 1 POWER STEPS CEA WITMDRAWAL LEVEL BY CROUPS 4g' , s i i 90 j f (2) 5 4 3 2 1 I y 100 135 180 180 180 180 o 80 * ' 90 90 180 ISO 180 180 q l i 80 45 153 180 180 180 n 70 0 72 180 180 180 i 4 g 70 - 60 0 36 144 180 180 n 2 ma F @m3 3 50 0 18 126 180 180 0 i 40 0 0 90 180 180 I s M o 60 -- 30 0 0 54 162 180 O 20 0 0 18 126 180 n ce gND H 10 0 0 0 90 180 e" g 50 - 0 0 0 0 54 162 i "v$ ~ g y o 40 - l r j d 30 - UNACCEPTABLE OPERATTON ACCEPTABLE OPERATION i 5 e W i '; i a WOTE: WHILE REDUCING power, CEA l k 0- ~ D INSERTION WITHIN THE CROSS-MATCNED i d RECION WILL BE PERMITTED FOR PERIODS rt i g NOT TO EXCEED ONE HOUR 10 h j / filllillllllilllifilll I lli lillllilllilllllilllllllillllillil croup i -1 caouP 3 cRote J. i 3 1 l l 1 I I I I I I I Il O 50 100 150 180 0 50 100 150 180 0 50 100 15) 180 I CROUP 2 CROUP 4 f I I I I l I l' l l l l 0 50 100 150 180 0 30 100 150 180 f F? CEA WITHDRAWAL BY GROUP (STEPS) J oE an H* NOTE: IN APPLYING THIS RESTRICTION TO WO LOOP OPERATION, THE RATED PCWER FOR WO LOOP OPERATION HUST BE USED UITHIN FOUR (4) IIOURS OF ATTAINING WO LOOP CONDITIONS i

Cycle Ase-age Burnup < 792 !M/'GU .u l l 100 { ' -- M ' l, l t a - : - -i-. J l -. ^ I l t l I-i --F !: l i& :; I i i l

i I

l I i l I . i--{.-03,87.-0.) ~

4--

- -i-. t l l l / l l .s i /. l /4 A. 30. Q :c. ' * - ~) t

..j l

r i i i/: N /. i i N- -/i i l+l - i\\ 0 . ip s.,,w.o.. l . si j. l~ .'.l.:.k .l / j. + .s. (.40,6t.2) l --l \\ . L. 60 r'.~w,0:.~) q). l .I .:.= ......j... ]... .:J.. .i ..1-

g I

I.. S& -.i. i i.:g

1. :.

~ t.- - :- - l. .s. l j... '1. '(8-l- - .I I 40 1 j r.. u l t g l l l I 30 t 20 __4 _,3 .2 .1 0 +.1 +.2 +.3 +4 +.5 Escore Monitor 5.0. Alarm Band for LOCA p].3* O 2 MAINE YANKEE Limiting Conditions of Operation When Incore. Technical Monitors are Tnoperable S;ieci fications 1.10 9 A.:*ndment No. 58 NOV 3 01981

Cycle Aver ye Bu. map > 7?2 MID/MTU I 100 I I- --- --~ .----l [ i 3 l s ~ l i t l 90 !'~! :.': l. (.06,38.5) ,(.04,88.5). i t / l 1 i j l 1 i i. I SO I [ i .(.19,79.0). - d -- l ( i i l-i 1 l t t g l

k l i

' -l 0

\\

l f i. I ,i l l g l' i ... ('). (- ;4 0,6 4 ;l) - (+.+U. M-s s' E 60

~-

.l' l. ,h t, .. ~

j l:

.::p* .....g....... ~i _ +- t l-s 50

:..l : 5:: -

l l-1 l - -:- :r. -

.t i

1 f.:: *.:'

5. l e -

l 40 l- .j l

. g. l.. :

i: 1 i. l l n. l l i a0 i 4 l l .l .i i i ! ".} 2 l 20 l l 3 + +.3 +.4 +.5 .S .4 .3 .2 .1 0 +.1 l Excore Monitor S.O. Alarm Band for LOCA Figure .M INE YANKEE Liniting Conditions of Operation When Incore 3.10 3 Technical Specifications Monitors a g operable z %0V 3 0 Ivm I Amendment No. H, 53

a No:e: 1. Pis curve inelndes 10* calculat f or.a1 uncertair.ty 2. T{ = F[

1. 03 3.

easured T{ should be aug=ented by ceasure=ent uncertainty (8%) before cccparison to this curve. ^ 1.f9 1 l a i i l l i ( 0, 1.6203 ) (.05, 1.6084 )

1. 4 j

j-"- (0.5, 1.6082 ) (1.0, 1.6015 ) (2.0, 1.6070 ) (4.0, 1.6297 ) l 1.67-(6.0, 1.6227 ) (8.0, 1.6.120 ) (10., 1.6538 ) (12., 1.6643 ) l,: l l l l 1.66!: -l I l Ii l-l l l l - 6' l 1 .l / l .1 /:- l i i-i i

1. 6..

l l [l l i .M - .i l : ff! -l 1 i/, l l ~.l I i l -i .5 - I. A I I i ...! / ;. l.; i j 3 f l .l 1_ i.. ~. A,f. ~! L 1.6.M-l l i 3 \\ 8 1.6C 0 2 4 6 8 10 12 Cycle Average Exposure (rx c/MT) .V.AINE YANKEE Allowable Unrodded Radial Peak Versus Figure Technical Cycle Average Burnup 3.10-4 i ' Specification 3.10-11 NOV 3 0 ig8j Arcanorent No. 38,10, 18, 58 ~~ . x=-o

2 t ? Note: CTA's are to be r.aintained at or abcVe the 100% power insertion limit when applying 3.10.B.2.2b ~ 113 l ,l. .... t. l I 3 i i. i i

2.7.100 C

A I ) IN l j: ; l.- ! j ,e i . Nn j +- l i N i . !.l EW l I 7.95,90 E 90 p f "f.. 4 i !? es x I i, w I ~ 1 + l o 14.4,80 8 e 80 i ii i . N i l l l i . 1G.0.75 g; i I 2 l 70 o 2 4 6 8 10 12 14 16 18 Allowable t Increase in FT (above Figure 3,10-4) R MAINE YANKEE A11cwable Power Level vs, Increase in Total Radial Figure Technical 3.10-5 Peak Specification OV 3 u min Amendment No. 38, fgg gg, 58

_ f i t t . ii l..i-j 4 h;.;.. i 4 l ... h. ...j. ". :. -

a J

j ..u.... .j... _ .s. i .o 2300 l-s 1 UNAOCEPTAB12CPqRATidN .n.e m C. (22,75,q50) -m i o 2250 I ue 1 .U. -3 Q225,.$5.0) .c, ACCEPTA3LE OPERATIDN, ? l l. l - w j. w t .1... ] ^ i l g 2200 .l m U t l 6* l A i t . l :s. i i i g t l c e . I: d' . : l.....:l :.

t. 6.5 q..:

U e i l.. .}.

i.,i. i m

2150 ". *, -.3. x i +- -.i

a e;..

i. o. . ; b.... .i ia g}i i. .i - i u i i . i 8 :! ---ii. -

{

.I g e 1.:., c 4 ' l.:- J -L 8..

l:.

{ e.. ri 1-

u..t. :.J. 41..a

~ 21.. . ;;7:'. O 8,; .' y 0 .a;l ? 'i: l.!.!i. ;'!.. :j:i.9.iM...r.d_M. Jgt,8 ' i,: ,8, ....j..._ .fl.l; '1 r .:E g g gg; .i. .i o i 2100 + - - - I....i d..ii y i i i-u l:[

. P.. !...: j

,..i-o i jp i. i e g g (d07), 54:2) "; .it ;pJ;..: i g ,g -t

l.a.

a: t. '?

i.:

s. .t i ..p.. y,.. ;.. .... ;..: r. - , f - g-pi.y.g. ?i .- iL i: i' ' ili-

in ",!r e

l l .i.. l r 2050 500 510 520 530 540 550 Norninal Cold Leg Temperature (indicated),0F l -J I 1 1 MOV 3 01981 1 1 l MAINE YANKEE Allo <atir. 3 Loop Steady State Coolant Conditions Figure Technical 3.10-6 Specificatieis 3,10-13 Amendment No.40 l l L

c. .. _..t.._ . _. _. _.. _._.4. ..._._. s .._..._J _.. _.._..__ 6.0 _. l RCS Baron Conc. Reactivity E..... . _. -.__-_~._

r.. _..

p ercen l .. _. _ ~ .._..m_. ._.4 l t \\ l [ . - _ __t _ Ce. _.. . s... _ ~ t ~- 4.0 .._t_.. ._.1.... . 1.... ._. _.%._. N.. 4._.... ..__1.._ ._..4._.._. _.. ...s. ._. A.. _....,. _...m _.... s. __ n_ x_ w __N......_ . __.t _.._.... 4._.._ _..... o .._._t._.__. ll t__ 3.O p. s I ~_m x ( u _.h.. .. __._3._.... w _..1_._... ..x _4.... u ..i _...........t..... g g a 2.0 ....t.... ._.N.,____ A... _... ___A..... .s.... -... _ N. 1.0 N...._... ..~ .i. ...o... ..i ...a... 0.0 0 200 400 600 800 1000 1200 RCS Boron Concentration - PPM' l i i MAINE YANKEE Minimum Required Shutdown Margin Figure l Technical Versus 3.10-7 Specification RCS Boron Concentration I

3.10-1+

d r e

l 3.11 CONTAIbNENT Aoplicability: Applies to the operating status of the reactor containment. Objective: To ensure containment integrity. Soecification: A. Containment integrity is defined to be operable when all the follow-ing are met: 1. All non-automatic containment isolation valves and blino flanges are closed. 2. The eculpment hatch is properly closed and sealed. 3. At least one hatch in the personnel air lock is properly closed and sealed. 4. All automatic containment isolation valves are operable or are lod <ed closed. 5. The uncontrolled containment leakage satisfied Specification 4.4 Section I.B.3. Exceptions: 1. When two containment isolation valves are in series, one may be open and inoperable for 4 hours to facilitate maintenance. 2. Exception (1) may be extended to 24 hours provided there is another valve on the same side of the containment boundary that may be closed remotely or by a person in constant attendance and in constant communication with the control rocm. 8. Containment integrity shall be maintained whenever there is fuel in the reactor and 1. The reactor coolant system is above 2100F or 2. The reactor coolant boron concentration is less than Cold Shutdown Concentration with the reactor vessel head in place, or 3. The reactor coolant boron concentration is less than refueling concentration with the reactor vessel head removed. Exception: On-line purging of containment is not a breach of containment integrity provided both valves on each line are automatically operable. 3.11-1 [OV 3 01981

C. The reactor shall not be critical if the containment internal pressure exceeds 3 psig. D. Containnent Weight of Air Monitoring System 1. The containment weight of air monitoring system shall be in operation whenever the reactor has been at power for more than 72

hours.

Exception: The system need not be operational during periods of system maintenance or calibration, periods of recharging the containment pressure, or periods of containment on-line purging and 48 hours thereafter. 2. When the containment weight of air monitoring system indicates a daily air loss greater than the following, an evaluation shall be

initiated to determine the validity of the indication.

a. eouivalent to 0.15 weight percent per day at 50 psig for seven consecutive days or I b. eaulvalent to 0.5 weight percent per day at 50 psig for four consecutive days or c. eculvalent to 1.0 weight percent per day at 50 psig for three

consecutive days 3.

The reactor shall be made subcritical within six hours if the evaluation recuired by D2: a. results in identification of the source of the leak and a determination that the known containment leak rate exceeds the equivalent of 0.15 weight percent per day at 50 psig through the containment integrity boundary or b. fails to identify the source of the leakage within ten days l and the Containment Weight of Air Monitoring System indica-tion persists at an average rate in excess of 0.15 weight percent per day at 50 psig. Basis: Specification A assures that the containment pressure boundary is defined

while permitting maintenance of components necessary to integrity.

Specification 4.4 Secticn 1.B.3 limits the uncontrolled containment l 1eakage to assure that public exposure will be maintained well within the

guidelines presented in 10 CFR 100 for the hypothethical accitent described in Section 14.18 of the FSAR.

Specification B includes a limit of 2100F on main coolant temperature assures that no steam will be generated in the unlikely event of a main coolant system rupture and hence no driving force to release any fission products from the containment. The shutdown margins are selected based upon the type of activities that are being carried out. The higher value for refueling precludes criticality under all postulated incidents involving fuel movement. The lower value with the head in place will also preclude criticality for all postulated incidents. 3.11-2 ZOV 3 01981

There is about a 5 psig margin between the calculated peak GCCident pressure and the containment design pressure of 55 psig. The 3 psig maximum operating pressure permits a positive containment pressure which is necessary for successful operation of the continuous leakage monitoring system. Specification D provides an added measure of assurance of containment integrity by specifying that the containment weight of air monitoring System be operational while recognizing the limitations Of such systems to reliably measure very small changes in air mass and its operational limitations. ( 3.11-3 . NOV 3 01981

3.12 STATION SERVICE POWER Applicability: Applies to station service electrical power systems. Objectivs: To assure an adeouate supply of electrical power during station operation. Specification: A. The following eculpment must be operable whenever the reactor coolant

system temperature and pressure exceeds 2100F and 400 psig.

1. One 115 kv incoming line in service. 2. Diesel generator DG-1A operable; 4160v er..ergency bus 5, 480v emergency bus 7, and d-c distribution cabinet 1 in service, or Diesel generator DG-1B operable; 4160v emergency bus 6, 480v emergency bus 8, and d-c distribution cabinet 3 in service. 3. 10,000 gallons of diesel fuel oil in the fuel oil tanks. Remedial Action: Restore recuired limiting condition within 4 hours then follow Specification 3.0. A. B. The following eculpment must be operable whenever the reactor is in a power operation condition: 1. One 115 kv incoming line in service. 2. Diesel generator DG-1A operable; 4160v emergency bus 5, 480v emergency bus 7, and d-c distribution cabinet 1 in service. 3. Diesel generator DG-18 operable; 4160v emergency bus 6, 480v emergency bus 8, and d-c distribution cabinet 3 in ser,vice. 4. 19600 gallons of diesel fuel oil in the fuel oil tanks. Exceptions: 1. If the 115 kv incoming line becomes unavailable, continued poser operation is permitted for maximum of seven days provided both diesel generators are operable. The NRC shall be notified within 24 hours of the plans for restoration of service. 2. If either diesel generator or its associated emergency buses or d-c distribution cabinet becomes unavailable, continued power operation is permitted for a maximum of seven days, provided one 115 kv line is available. In this situation the operable diesel generator is to be tested once a day. Remedial Action: Restore recuired limiting condition within grace period specified then follow Specification 3.0. A. 3.12-1 s0V 3 01981

O C. Under accident conditions the automatically connected load to either diesel generator shall not exceed the short time rating of 2900 kw. Basis: Availability of power to the engineered safeguards eculpment is necessary when the reactor is at power. If the loss of both incoming lines, a diesel generator or its associated emergency buses occurs, a period of seven days operation is permitted while the situation is being assessed and full redundancy is being restored. This time period is justified becaues adeouate sources of power remain available for the operation of the engineered safeguards eculpment. The fuel recuirement of 19600 gallons is made up as rollows: A. 17827 gallons is the amount that will be recuired for the maximum expected engineered safeguards load for a period of seven days; B. 10 percent of the above as a contingency for any non-engineered safeguard recuirement during this period. Specification A assures that an emergency power source is available wnenever the reactor coolant system is above the specified pressure and temperature limit. It recognizes the decreased consecuences of a loss of

coolant accident if the reactor is not at power.

The rating in specification C is determined by currently acceptable standards. l l i l l l t 3.12-2 lNOV 3 01981

3.13 REFUELING OPERATIONS Applicability: Applies to operating limitations during refueling operations. Cbjective: To minimize the possibility of an accident occurring during refueling operations that could affect the health and safety of plant personnel and the public. Specification: A. The following conditions shall be satisfied during refueling operations: 1. The containment venting and purge system, including inlet and outlet trip valves that isolate the ventilation system in response to radiation monitors, shall be operating, with the discharge filtered through the high efficiency particulate air filters and charcoal absorbers. Exception: The high efficiency particulate air filters and charcoal absorbers may be bypassed during operations which may be detrimental to these filters (welding, painting, etc.), provided that the containment purge trip valves remain trippable both manually and automatically in accordance with Spec. A.2 below. 2. a. Two radiation monitors that initiate isolation of the containment ventilation system, shall be tested and verified to both be operable immediately prior to fuel handling operations and remain coerable during fuel handling operations. The two monitors shall be located on the 5 containment fuel handling area level, shall be part of the plant area monitoring system, and shall employ one-out-of-two loolc for isolation. Exceotion: The valve trip system may be bypassed for a period not to exceed 0.5 hours daily to facilitate routine testing of the radiation monitors. Remedial Action: Should one of the area monitors become inoperable, repairs shall be affected immediately and the logic shall revert to one-out-of-one for isolation. Refueling operations may continue for a maximum of 12 hours in this mode. b. The capability of the containment purge trip valves to res-pond to a trip signal from the radiation monitors shall be tested immediately prior to fucl handling operations and weekly thereafter. 3. Radiation levels in the containment and spent fuel storage areas shall be monitored continuously. 4. Whenever core geometry is being changed, neutron flux shall be continuously monitored by at least two wide range logarithm 3.13-1 NOV 3 01981

) l monitors, with each monitor providing continuous visual indication in the control room. When core geometry is not being changed, at least one source range neutron monitor shall be in service. 5. At least one residual heat removal pump and heat exchanger 3.lall. - be in operation. Exception: This system may be shutdown for a maximum of 6 hours'. to facilitate upper guide. structure assembly removal or other special maintenance operations. During suc.h. periods the reactor coolant temperature shall be continuously monitored and the initiation of core cooling flow shall be continuously available: / 6. Both RHIS loops A and B shall be operable when the water level above the top of the irradiated fuel assemblies seated' within the 'N reactor pressure vessel is less than 23 feet. s ~ 7. During reactor vessel head removal and while refueling operations are being performed in the reactor, the refueling baron concentration shall be maintained in the reactor coolant systein, and shall be checked by sampling on each shift to insure that the boron concentration is such to maintain the core 5% delta K/K \\ suberitical.. 8. Direct commnication between personnel in the control room.and at the refueling station shall be operable whenever. Changes.in core geometry are taking place. / Remedial Acticn: If any of the conditions in Specification A are' riot met, all refueling operations shall cease immediately; work shall be initiated to satisfy the recuired conditions, and no operations that may increase the reactivity of the core shall be made. B. Whenever spent fuel is being handled in the spent fuel pit, the fuel building ventilation systems shall be in operation with the discharged air passing through a filter pack containing a charcoal filter before going to the primary vent stack. C. Prior to initial core loading and prior to each refueling a' complete check out, including a load test, shall be conducted on fuel hardling cranes that will be used to handle spent fuel assemblies. D. A minimum of 23 feet of water above the top of the core shall be maintained whenever spent fuel is being handled. E. Irradiated fuel shall not be handled for 72 hours after reactor shutdown. F. Spent fuel storage racks may be moved only in accordance with written procedures which ensure that no rack modules are moved over fuel assemblies. ~ Basis: The eouipment and general procedures to be utilized during refueling are discussed in the FSAR. Detailed instructions, the above specifications and the design of the fuel handling equipment incorporating built-in interlocks and safeguards systems provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety. n0V 3 01981 3,13_2 1 g

~ t r 4 The exception to paragraph 3.13.A.1 permits operations which may be f detrimental to the integrity of the filters to be conducted during refueling operations, thus eliminating unnecessary outage delays. Analysis n~as shown that a refueling accident occurring in excess of 72 hours after shutdown (Specificaticn l3.13.E) will not result in offsite consecuences in excess of 10CFR100-even if no credit is taken for filtration or containment isolation. The' exception tr.' paragraph 3.13. A.2 permits routine testing of the ~ ' radiation monitors without incurring unnecessary wear of the purge valve ~ resiliantsseals. Weekly testing of these trip valves is sufficient to insure their' ope:9b111ty. r Whenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumentation. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The residual heat removal flow is used tc remove core decay heat and maintain a uniform boron concentration. ' N singleicooling mechanism is* sufficient to remove decay heat but single failure considerations require that two mechanisms be OPERABLE. Thesh:ddownmarginasindicate'dwillkeepthecoresubstantially i subcritical, even if the highest worth CEA's were inadvertently withdrawn from the core without compensating baron addition. - ' Periodic checks of refueling water boron concentration insure the proper shutdown margin.' Communication requirements allow the control room - operctor to inform the refueling station operation of any impending visual ccndition detected from the main control board indicators during fuel movement. s In additico to the above engineered safeguards systems, interlocks are s utilized onring refueling to insure safd handling. An excess weight l interlock is'provided to prevent excess loading of a fuel assembly, should l it iradvertently become stuck. ~ The charcoal filter installed in the fuel handling building exhaust will handle the full 15,000 cfm capacity; of the normal ventilation flow. The f offsite thyroid dose as calcultted for the fuel handling incident is well , t;elow the 19 CFR 100' gu' Aline'dese. Valve alignment check sheets are ' completed to protect gainst sources of unbarated water or draining of the system. I'n the anal'ysis of the refueling accident conducted by the Staff, 23 feet P of water and 72 hours of decay time were used to limit exposures to 10% of l 10 CFR 100. l L Procedures are recuired'for movement of spent fuel racks to avoid unnecessary risk of spent fuel damage caused by dropping spent fuel racks. N s s y l ~

ov: o1981

,x 2 L ? f

3.14 PRIMARY SYSTEM LEAKAGE Acolicability: Applies to limiti.'g operation of the plant under varying rates and conditions of primary system leakage. Objective: To specify primary plant operability with primary system leakage. Soecification: A. When the reactor is above 2% power, two reactor coolant leak detection systems of different operating principles shall be operating, (*.th one of the two systems sensitive to radioactivity in the containment. Exception: The systems sensitive to radioactivity may be out of service for a period of up to 48 hours provided two other means of reactor coolant leakage detection are operable. B. Whenever reactor coolant system indicated leakage exceeds 1 gpn by any means available an investigation as to source and safety implications will be initiated as soon as practicable but no later than within four hours. C. Reactor coolant system leakage shall not exceed any of the Specifications 1 through 5 below. 1. Leakage into the reactor containment of any magnitude that may be an indication of a deterioration of crimary system pressure boundary strength welds or material. 2. Leakage into the reactor containment in excess of I gpm through bolted closures, valve packing, or other mechanical connections. 3 Leakage in excess of 1 gpm that is unexplained or unaccounted for. 4. Leakage in excess of 10 gpm to aerated or uncontained systems. 5. Total leakage through all steam generator tubes shall not exceed 1.0 gom. Remedial Action: If reactor coolant system leakage exceeos any of the Specifications C.1 through C.5 above, the reactor shall be shutdown within 24 hours. The reactor shall not be brought critical until the leak has been repaired or until the problem is otherwise corrected. Basis: Reactor coolant system leakage may be indicated by one or more of the following methods: 1. Primary system water balance inventory. 3.14-1 0:0V 3 01981

O 2. Containment sump level. 3. Containment air particulate monitor and/or radio gas monitor .4. Containment atmospheric humidity and/or temperature. 5. Steam generator blowdown monitor and/or air ejector effluent monitor. Leakage may be indicated and/or identified by the following routine and special plant surveillance operations. 1. Primary system hot leak test. (Involves monitoring steady-state drop in volume control Tank Lesel). 2. Direct observation from accessible locations within the containment. 3. Sampling and analysis of containment atmosphere, steam generator blowdown and air ejector effluent for radioactive and non-radioactive tracers. Reactor coolant system leakage will be maintained at the lowest practical value su that small leaks, with possible safety implications, will be more readily detected and identified. A safety evaluation of a leak shall consider its magnitude, nature, and possible consecuences. It shs11 assure that off-site radiation exposure from the primary coolant system activity is within the guidelines of 10 CFR 20. For the purposes of determining a maximum allowable secondary coolant activity, the steam break accident is based on a postulated release of the contents of three steam generators to the atmosphere using a site boundary dose limit of 1.5 rem. The limiting dose for this accident results from iodine in the secondary coolant. The reactor coolant distribution of ~ iodine isotopes with 1% failed fuel was used for this calculation. I-131 is the dominant isotope because of its low WC in air and because the other iodine isotopes have shorter halflives and therefore cannot build up to significant concentration in the secondary coolant, given the limitations on primary system leak rate and activity. The steam generators which operate at a constant programmed level contain 131m3 of water at standard conditions. One tenth of the contained iodine is assumed to reach the site boundary, making allowance for plate-out and retention in water droplets. The maximuo inhalation dose at the site boundary is then as follows: Dose (rem) = C v B(t) X/Q DCF 10 where: C = Secondary coolant sample activity 0.2 micro ci/cc = 0.2 C1/M3 V = Water volume in three steam generator = 131 M3 at standard conditions ] 3.14-2 NOV 3 01981

B (t) = Breathing rate (3.47 x 10-4 3 m /sec) X4 = 6.48 x 104 sec/m3 (corresponding to Pasquill F stability and 1 m/see wind speed) DCF = 1.48 x 106 rem /Ci I-131 inhaled The resulting thyroid dose is less than 1.5 rem. i l i i i l l t i i l l I \\ 3.14-3 "'1V 3 01981

3.15 REACTIVITY ANOMALIES Acolicability: Applies to potential reactivity anomalies. Objective: To recuire evaluation of reactivity anomalies within the reactor. Specification: Following a normalization of the computed boron concentration as a function of burnup, the actual boron concentration of the reactor coolant shall be periodically compared with the predicted value. If the difference between the observed and predicted steady-state concentrations reaches the ecuivalent-of 1% in reactivity, the Nuclear Regulatory Commission shall be notified and an evaluation as to the cause of the discrepancy shall be made and reported to the Nuclear Regulatory Commission in accordance with Technical Specification 5.9.1.6. Basis: To climinate possible errors dln the calculations of the initial reactivity of the core and the reactivity depletion' rate, the predicted relation between fuel burnup and the baron concentratidn, necessary to maintain adeouate control characteristics, must be adjusted (normalized) to accurately reflect actual core conditions. When full power is reached initially, and with the CEA groups in the desired positions, the baron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured baron concentration is compared with the predicted concentration and the slope of the curve relating burnup and reactivity is compared with that predicted. This process of normalization should be completed after about 10% of the total core burnup. Thereafter, actual boron concentration can be compared with prediction and the reactivity status of the core can be continuously evaluated, and its occurrence would be thorGJghly investigated and evaluated. The methods employed in calculating the reactivity of the core vs. burnup, and the reactivity worth of boron vs. burnup, are given in the FSAR. 3.15-1 NOV 3 01981

3.16 RELEASE OF LIQUID RADIOACTIVE WASTE Applicability: Applies to the controlled release of all liquid waste discharged from the plant which may contain radioactive materials. Objective: To establish conditions for the release of liquid waste containing radioactive materials and to assure that all such releases are within the concentration limits specified in 10 CFR Part 20. In addition, to assure that the releases of radioactive material in liauld wastes (above background) to unrestricted areas meet the low as practicable concept, the following liquid release objectives shall apply: A. The annual total cuantity of radioactive materials in liquid waste, excluding tritium and dissolved gases, shall be less than 5 curies; B. The annual average concentration of radioactive materials in liquid waste, excluding tritium and dissolved gases, shall not exceed 2 x 10-8 micro C1/ml; C. The annual average concentration of tritium in liauld waste shall not exceed 5 x 10-6 micro Ci/ml; D. The annual average concentration of dissolved gases in liquid waste shall not exceed 2 x 10-6 micro C1/ml. Soecification: A. Release Quantities and Concentrations of Radioactive Materials in Licuid Waste 1. If the experienced release of radioactive materials in liquid wastes, when averaged over a calendar cuarter, is such that these cuantities if continued at the same release rate for a year would exceed twice the annual objectives the licensee will: a. make an investigation to identify the causes for such release rates; b. define and initiate a program of action to reduce such release rates to the design levels, and; c. describe these actions in a report to the Commission within 30 days. 2. If the experienced release of radioactive material in liquid waste, when averaged over a calendar quarter, is such that these cuantities if continued at the same release rate for a year would exceed eight times the annual objectives, the licensee shall define and initiate a program of action to assure that such release rates are reduced, and shall submit a report to the Commission within 7 days describing the causes for such release rates and the course of action taken to reduce them. 3.16-1 90V 3 01981

3. The rate of release of radioactive materials in liquid waste from the plant shall be controlled such that the instantaneous concentration of radioactivity in liquid waste does not exceed the values listed in 10 CFR Part 20, Appendix 8, Table II, Column 2. B. Treatment and Monitoring 1. The equipment installed in the liquid radioactive waste system shall be maintained and operated with the intent of keeping releases within the objectives of these specifications. 2. At least one service water pump shall be in operation when liquid radioactive wastes are being released. 3. Liould waste disenarged from the test tanks shall be continuously monitored during release. The liquid effluent monitor reading shall be compared with the expected reading of each discharge batch. The monitor shall be tested daily and calibrated at refueling intervals. The calloration procedure shall consist of exposing the detector to a referenced calibration source in a controlled, reproducible geometry. The sources and geometry shall be referenced to the original monitor calibration which provides the applicable calibration curves. 4. The effluent control monitor shall be set to alarm and automatically close the waste discharge valve such that the recuirements of the specification are met. In the event of a malfunction in the monitor, the alarm shall sound and automatically close the waste discharge valve. 5. Steam generator blowdown shall be continuously monitored, except that during periods when the monitor is not operating, daily grab samples shall be taken. C. Samplina and Analysis In addition to the above continuous monitoring requirements, liquid radioactive waste sampling and activity analysis shall be performed in accordance with Table 3.16-1. Records shall be maintained and reports of the samplirg and analysis results shall be submitted in accordance with Sections 5.6 and 5.7 of these Specifications. Basis: It is expected that the releases of radioactive materials in licuid waste will be kept within the design objective levels and will not exceed the concentration limits specified in 10 CFR Part 20. These levels provide reasonable assurance that the resulting annual exposure to the whole body or any organ of an individual will not exceed 5 millirems per year. At the same time, the licensee is permitted the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in releases higher than the design objective levels but still within the concentration limits specified in 10 CFR Part 20. It is expected that using this operational flexibility under unusual operating conditions, the licensee shall exert every effort to keep levels of radioactive material in liquid wastes as low as practicable 3.16-2 NOV 3 01981

and that annual releases will not exceed a small fraction of the annual average concentration limits specified in 10 CFR Part 20. The design objectives have been developed taking into account a conbination of variables including fuel failures, primary system leakage, primary-to-secondary leakage and the performance of the various waste treatment systems. The actual magnitude of these parameters are as follows: A. Maximum expected reactor coolant corrosion product concentrations; B. Reactor coolant fission product concentration corresponding to 0.1% fuel cladding defects; C. Steam generator primary-to-secondary leak rate of 0.01 gpm; D. Hydrogenated licuid waste generation rate of 1.75 gpm; E. Aerated liquid waste generation rate of 0.475 gpm; F. Steam generator blowdown rate of 5 gpm, of which 3 gpm is diverted to the waste disposal system for processing before discharge; G. Decontamination factor of 104 for all radionuclides except tritium for the boron recovery and waste disposal evepeators; H. Decontamination factor of 10 for Cs, Sr, Mo ar.d Y for cesium demineralizer, j The application of the above estimates results in the radionuclide discharge concentrations and rates shown in Table 3.16-2. Also given in this table are the radionuclide concentrations in the reactor coolant and the secondary coolant, which are the " source terms" for releases from the primary and secondary systems, respectively. Licuid radioactive waste is mixed with service water in the plant discharge system prior to release. } With four circulating water pumps in operation, the rated capacity of ?iu system is 400,000 gpm. This is equivalent to a dilution multiple of 2.5 x 10-6 min / gal x the discharge rate in gal / min. Liquid radioactive waste i from the waste treatment system is collected and stored in tanks until a cuantity sufficient for processing has accumulated. The processed liquid waste is discharged through a recorder controller which provides a measure t and control of volume of licuid released. The volume discharged and the analysis of the proportional composite sample provide the basis for reporting the quantity and concentration of activity released. p The operating msnual will identify all equinment installed in the liquid waste handling and treatment systems and will specify detailed procedures for operating and maintaining this eculpment. The low as practicable licuid release objectives expressed in this l specification are based on the guidelines contained in the proposed l Appendix I of 10 CFR 50. Since these guidelines have not been adopted as i yet, the release objectives of this Specification will be reviewed at the 3 time Appendix I becomes a regulation to assure that this Specification is based uNa the guidelines contained therein. j 3.16-3 NV 3 01981

9 Tabic 3.16-1 j RADICACTIVr, LIOUID VASTE SAMPLING AND ANALYSIS A. Tect Tank Ref er.nes Type of Sensitivity (5)of Annlyais Scuplin <t Frcouency'. Activitv Ancivnie Each Betch _Cronc/I..I 10~7 ~!.Ci/ml One q.rch/th nth Dianolved Noble Cac,co 10-4 '/.C1/n1 Weekly Proportionci 8a-140, La-140, 1-131 10-6pi/mi Composite (1) Monthly Proportional Gr.ntna Emitters. 10'5 /.Ci/n1(2) 10-5 }f;ci/mi Oi/ml Composite (1) H-3 10-T Grons c< Quarterly Proportioncl Sr-89, Sr-90 10-by6C1/ml(0) t Comportre (-) B. 'Secondsry Plant blovdown e.nd Tenhnre Reler.nce(3) Type of Sensitivity of (( Sampling Frecuenev Activity Ane.1voto Annivcin 5) l Weekly Crona 6, I 10-7 g i/ml Da-140, LA-140. I-131 10- 6' f,,Ci /m i One Sempic/ Month Dinnolved Nobin Gnnen 10-4 's.,C1/mi Monthly Proportional Gamma Emitters 10-O '/41/ml G ) Composite (') H-3 10-J'fif/mi Crone OC 10-/'i Ci/ml Sr-89, Sr-90 10-8/ Ci/m1(0) QuarterlyProgortional Comnocite () NO*lES: O)A proportional sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged from the plant. (2)For certain m,y.tures of ganana emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sample in much greater concentrations. Under thene circumstances, it will be more appropriate to calculate the concen-trations of such radionuclides using observed ratios with those radionuclideo which are measuresble. 3.16 4 j I NOV 3 01981

e (3)Seconddry plant blevdoun and secondary plant leakage are cach subject to the scmpling and analycic requirer.ents contained in Part B of Table 3.16.1. (4)Since these potentici sourecs of liquid radioactive vaste are discharged on a continuous rather than batch bsc3s, the volume of liquid to be used as a basis for obtaining proportional sau.ples from cecondary blovdown and leakdge is that amount discharged over the period of one week. (5)These activity analyce's sensitivitics are based en the projected capabilitics of laboratory instru=entation and techniques to be employed by Maine Yankce. In order to assure that actual !!ainc Yankee operating experience is utiliced, a reevaluation will be perforced within 2 years of initial full power operation of the plant to determine whether these censitivities should be revised. (0}0ne quarterly proportional composite sample vill be collected and analy:cd for Sr-89 and Sr-90. The proportional inputs to this sacple will be from the test tank, secondary blowdown, and second,ary leakage releaces. (7)Whenever gross B,y is less than 10 c/ml proportioned composite sample -8 analysis is-not required. I l l t 1 e 6 l 3.16-5 h0V 3 01981 1 t

4-S ca o G3 g _s .m Table. 3.16-2 l RADIONUCLIDE SOURCE TEES AND DISCHAIGES Reactor Coolant Steam Generator Plant Discharge Fraction of Blowdown Concentration Concentration 10 CFR 20 Annual'Palease Concentration Isotope (pCi/ml 0 70 F) (pCi/ml @ 70'F) (pCi/n1) MFC.. (Ci/ year) I-131 2 99-1* h.22-h 2.16-9 7 20-3 1.6h I-132 1.12-1 6.25-6 8.88-11 1.11-5 6.73-2 I-133 5 02-1 2.c8-h 1.25-9 1.25-3 9.h9-1 I-13h 7.55-2 1 57-6

h. 73 -11 2 37-6 3.58-2 I-135 2.80-1 h.33-5 3 52-10 8.60-5 2.67-1 sr-89 3 11-h 5.82-7

.2.81-12 9 37-8 2.13-3 sr-90 1.89-5 3 77-8 1.82-33 6.07-7 1 381h Sr-91 1 97-h h.25-8 2 32-13 h.6h-7 1.76-h Y-90 6.13-5 5.hh-8 2.68-13 13h-S 2. 03 -h'. T-91 2.h9-3 h.70-6 2.27-11 7.57-7 1.72-2 Mo-99 1.18-1 1.06-h 5.21-10 1 30-5 3 95-1. Ru-103 1.96-h 3 61-7 1.62-12 2.28-8 1.38-3 -Ru-106 1.82-5 3 60-8 1.81-13 1.81-8 1.37-h P Te-129 2.66-3 7.2h-8 1.7h-12 2.18-9 1 32-3 E Te-132 2 51-2 2.h8-5 1 31-10 h.37-5 9.92-2 i Ba-lho h.37-h 6.92-7 3 51-12 1.76-7 2.66-3 i La-lh0 h.17-h 2 79-7 1 5h-12 7.70-8 1.17-3 Cs-13h 5 19-2 9 93-5 h.79-10 1.60-h 3.63-1 Cs-136 1.75-3 .2.78-6 1 58-12 ,2.63-8 1.20-3 Cs-337 1.h9-1 2 97-h 1.hh-9 7.20-5 1.09 Cr-51 5.25-3 9 39-6 h.7h-11 1.58-5 3.52-2 i Hn-5h 3.80-5 7.50-8 3 76-13 3 76-9 2.85-h Fe-59 2 9h-5 5.h6-8 2.75-13 5.50-9 2.08-h Co-58 6.h3-3 1.26-5 6.32-11 7 02-7. .h.79-2 Co-60 7.25-h 1.bh-6 7.22-12 2.hl-7 5.h7-3 Zr-95 1.29-6 2.h5-9 1.23-lh 2.05-10 9 33-6 1-6.62 x 10-9 {- 8.36 x lo-J f= 5 02 H-3 2 3-2 h.59-5 1.22-7 h.07-5 92 3

2 99 2 99 x 10*1 0

9 3.17 RELEASE OF GASEOUS RADIOACTIVE WASTE Acolicability: Applies to the controlJed release of all gaseous waste discharged from the plant which may contain radioactive materials. mjective: To establish conditions in which gaseous waste containing radioactive materials may be released and to assure that all such releases are within the concentration and dose limits specified in 10 CFR Part 20. In addition, to assure that the releases of gaseous radioactive wastes (above background) to unrestricted areas meet the as low as practicable concept, the following objectives shall apply: A. Averaged over a yearly interval, the release rate of radioactive i isotopes, except I-131 and particulate radioisotopes with half lives greater than 8 days, discharged at the plant stack, shall be limited as follows: i f { 01 less than or eouc1 to 800 m3 sec / (MPC)i where Qi is the annual controlled release rate (C1/sec) of { radioisotope i and (MPC)1 (micro ci/cc) is defined for radioisotope 1 l in column 1, Table II of Appendix B to 10 CFR 20. B. Averaged over a yearly interval, the release rate of I-131 and other particulate radioisotopes eith half lives longer than 8 days, discharged at the plant stack, shall be limited as follows: { 01 less than or equal to 5.6 m3 sec / (MPC)1 where Qi and (MPC)1 are as defined above. Specification: A. Release Quantities and Concentrations of Radioacative Materials in Gaseous Waste 1. If the experienced rate of release of radioactive materials in gaseous wastes, when averaged over a calendar cuarter is such that these cuantities if continued at the same release rate for a year would exceed twice the annual objectives, the licensee will: a. make an investigation to identify the causes for such release rates; b. define and initiate a program of action to reduce such release rates to the design levels; c. describe these ections in a report to the Commission within 30 days. 17~1 COV 3 01981

2. If the experienced rate of release of radioactive material in gaseous wastes, when averaged over a calendar Quarter, is such 4 that these cuantities if continued at the same release rate for a year would exceed eight times the annual objectives, the licensee j shall define and initiate a program of action to assure that such release rates are leduced, and shall submit a report to the Commission within 7 days describing the causes for such release rates and the course of action taken to reduce them. J 3. The rate of release of radioactive materials in gaseous waste from the plant (exttot I-131 and particulate radioisotopes with half 1 lives greater than 8 days) shall be controlled such that the maximum release rate averaged over any one-hour period shall not exceed: 01 = 3.1 > 104 m3/sec 7 (MPC)1 \\ 8. Treatment and Monitoring l 1. At least one exhaust fan shall be in operation when radioactive gaseous wastes are released to the stack. 4 2. During release of radioactive gaseous waste from the gasaous waste 1 decay drums to the stack, the following conditions shall be met: a. 1. The gas decay drum effluent monitor and the stack sampling uevices for halogens and particulates shall ba operable. 2. The normal response of the decay drum effluent monitor shall be verified by comparison with the prerelease sample analysis. 3. The monitor shall be tested prior to any release of radioactive gas from a decay drum. 4. The monitor shall be calibrated at refueling intervals. The calibration procedure shall consist of exposing the detector to a referenced calibration source in a controlled reproducible geometry. The source and geometry shall be referenced to the original monitor calibration which provides the applicable calibration curves. b. The gaseous waste from the decay drumi shall be filtered through the high efficiency particulate air filters and the diarcoal absorber provided. 3. a. During normal conditions of plant operation, radioactive gaseous waste from the hydrogenated waste gas system shall be provided a minimum average holdup of 60 days except for low radioactivity gaseous waste resulting from purge and fill operations associated with refueling and reactor startup. Remedial Action: Holdup time less than that specified in B.3.a above shall be covered in the special effluent report to be included in the semi-annual report recuired by section 5.7.B.(1)(a} of these specifications. NOV 3 01981 3.17-2

b. The maximum activity to be contained in one gas decay. tank shall not exceed 88,400 curies of Xe-133 equivalent. 4. During the first indication of primary-to-secondary leakage, concurrent with sufficient fuel defects, a determination of the ladine partition factor for the blowdown tank shall be made. 5. During power operation, the cont 2mor air ejector discharge shall be continuously monitored for gross radiogas activity. Whenever this monitor-is inoperable, grab samples shall be taken from the air ejector discharge and analyzed for gross radiogas activity daily. 6. Gases discharged through the stack shall be continuously monitored for gross noble gas and particulate activity. Whenever either of these monitors is inoperable, appropriate grab samples shall be taken and analyzed daily. 7. Reactor building purge shall be filtered tnrough the.high efficiency particulate air filters and charcoal absorbers whenever the average air concentration of iodine and particulate isotopes inside the reactor building exceeds the occupational NPC. C. Sampling and Analysis In addition to the above continuous sampling and monitoring reouirements, gaseous radioactive waste sampling and activity analysis shall be performed in accordance with Table 3.17-1. Records shall be maintained and reports of the sampling and analysis results shall be submitted in accordance with sections 5.6 and 5.7 of these specifications. Basis: It is expected that the releases of radioactive materials in gaseous waste will be kept within the design objective levels and will not exceed on an instantaneous basis the dose rate limits specified in 10 CFR Part 20. These levels provide reasonable assurance that the resulting annual exposure from noble gases to the whole body or any organ of an individual will not exceed 5 millirems per year. At the same time, the licensee is permitted the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable j source of power under unusual operating conditions which may temporarily result in releases higher than the design objective levels but still within the concentration limits specified in 10 CFR Part 20. It is expected that using this operational flexibility under unus.ual operating conditions, the licensee shall exert every effort to keep levels of radioactive material in gaseous wastes as low as practicable and that l annual releases will not exceed a small fraction of the annual average concentration limits specified in 10 CFR Part 20. These efforts shall include consideration of meteorological conditions during releases. 4 I l l I 3.17-3 lNOV 3 01981

The design objectives have been developed taking into account a combination of system variables including fuel failures, primary system leakage and the performance of radioisotope removal mechanisms. The values assumed for these variables include the following: A. Reactor coolant fission pruduct concentration corresponding to 0.1% fuel cladding defects; i B. Steam generator primary-to-secondary leak rate of 0.01 gpm; C. Steam generator blowdown rate of 5 gpm; D. Reactor coolant leakage to the containment building of 0.25 gpm and four containment vents per year; E. Partition factor of 1000 for iodine in aerated drains tanks; F. Gas decay drums average 60 days holdup; G. Decontaminiation factor of 1000 for iodine in the degassifier; H. Charcoal filter efficiency of 99% for iodine on the air ejector, aerated vent and gas decay drum systems. The application of the above estimates result in the radiogas discharge rates shown in Table 3.17-2. The noble gas release rate stated in the objectives is' based on a XA] value from the annual meteorological data. The dispersion factor used, 2.59 x 10-5 3 sec/m, is conservative and controls the release rate of a small fraction of 10 CFR Part 20 requirements at the site restricted area boundary (less Ulan 10 mrem per year). The I-131 and particulate release rate stated in the objectives limits the l concentration at the restricted area boundary to less than 1% of the MPC listed in 10 CFR 20. The release rate also controls the expected l concentrations at nearby commercial dairy farms to much less than l 1/100,000 of the 10 CFR 20 requirements. t The maximum one-hour release rate limits the dose rate at the site boundary to less than 2 mrem / hour even during period of unfavcrable meteorology. (Moderately stable conditions with 2 m/sec wind speed). 1 The maximim activity in a waste gas decay drum is specified as 88,400 curies of Xe-133 equivalent based on a postulated rupture that allows all of the contents to escape to the atmosphere. This specification limits the maximum offsite dose to well below the limits of 10 CFR 100. I The gaseous waste system is divided into two sections; aerated gases and i hydrogenerated gases. Low activity, aerated gaseous wastes are discharged to the aerated gas header and through a high efficiency filter to the primary vent stack. Hydrogenated gaseous wastes flow from the surge drum and through the gas compressor which discharges to the waste gas decay drum. The drum is pressurized and then isolated for decay of the gaseous wastes before discharge to the primary vent stack. The gaseous discharge I NOV 3 01981 3 17-4 t'

is continuously monitored both in the vent line to the primary auxiliary building fan suction and in the stack. Upon detection of high activity in the vent line or upon the loss of ventilation fan suction, the vent line . flow control valve will close, terminating the release of gaseous waste. The cuantity and isotopic proportions of radioactive gases released into the reactor coolant system is dependent upon several factors including fuel leakage, burnup and power level. Changes in power level will affect gaseous generation rates temporarily. Cases are released from the reactor coolant to the gaseous waste system during degassifier treatment of the letdown and leakage water and also during venting of the system. This venting may occasionally be performed to degas the system and so control plant chemistry and/or reduce coolant radioactive gas concentrations to an acceptable value for the protection of plant personnel. Gaseous waste holdup and decay occurs while it is retained in the reactor coolant system and in the surge drum of the gaseous treatment system. The gaseous waste holdup drums are of sufficient capacity to provide an additional average retention period of 60 days during normal operating conditions. The low as practictsle gaseous release objectives expressed in this specification are based on the guidelines contained in the proposed Appendix I of 10 CFR 50. Since these guidelines have not been adopted as yet, the release objectives of this Specification will be reviewed at the time Appendix I becomes a regulation to assure that this Specification is based upon the guidelines contained therein. Specifiacation B.7 above describes when the reactor building purge shall be filtered through the high efficiency air filters and charcoal absorbers. The average air concentration of particulate isotopes may be measured by the Containment Air Particulate Detectors. The average air concentration of iodine may be measured by local sampling. l l l l i CDDV 3 01981 3 17-5 l t

9 9 Tabic 3.17-1 RADICACTIVE CASEGUS tJASTE SAMPLINO AND ANALYSIS A. Cao Decay Drun Relcenen c of Scopling Type of Sensitivity (1) Anclycie Snmple 'ITpe Frequency Activity Ant.lycis 10-S ffet/ce Gas Each Drum Release Groce Ccrenrr individual Gex=a 10-4 /Aci/ce(2) Emittern 11. Contcinment Venting Relenren of Sampling Type of Sencitivity(1)' f.nc ivcis Sample Type Frcottency Activity Annivcin \\ 1 Gac Each Vent Grcu Cem.a 10-5j ci/ce Individual Gsc.m 10-4 f Ci/cc(2) Enittnr3 Dehur.iidified Ecch Vent li-3 10-Ogi/cc Sample C. Condenner Air Eiector Rclesces of Sampling Type of Senoitivity(1) Analysia r Sample Type Frecuency Activity Anniveis 10-4 h-C1/cc Gas Monthly Gross cacnc Individual Ccr.mts 10-3kCi/cc(2) E:nitters e 3.17-6 [' NOV 3 01981

9 O (' . Table 3.17-l (cont'd) D. Stack Releccca of Szcapling Type of Scnuitivity(ll An.* 1y. iia Sample Tye Frecuency Activity Analynie Cao Quartcrly Gr,oce Gr:n:n 10-6 hCi/cc ~ Individual Ga==n F.r.l ttero 10-5 h.ci/ce(2) i Dehumidified Each Decay Drum H-3 10-0),ci/cc Sewsle Rolcar.c Ciarcoal Weekly I-131, 1-133, 1-135 3 x 10-12,sC1/cc Weekly' Groan 4.Y 3 x 10-12 4,ci/cc Weekly Ba-140, I,c-140,,..- I-131 3 x 10-11 e.Ci/cc f Particulaten Monthly Compocite Grorf d f _i 3 x 10-12'pci/cc of Wechly Scrapleo Individual Gcana Emit t e r., J 3 x 10-II.t-Ci/cc Quarterly C~.3po-Sr-89, Sr-90 l~x 10-iff.C1/cc p cite of Weekly I, ,_Sumplec C.ie Ucekly Cross g 3 x 10-12j41/cc S eple/Qutreer ..x : - r l Non s: (1)'the above activity analysis sensitivitice cre based on the projected capability of laboratory instrumentation and technique.e to be cmployed by Maine Yankee. In order,to c6eur~c that cctual Maine Yenlce operating experience is utilized, a reevaluation will be performed within 2 ycars of initici full power operation of the plcnt. (2)For certain mixturco of scmma emitters, it may not be possible to mescure radionuclideo at icvels near their sencitivity limite when other nuclides arc precent in the comple at much higher levela. Under thcac circumstances, it will be more appropriate to calculate the icvela of such radionuclidec using obacrved ratioc with those radionuclideo which are measurable. (* 3.17-7 NOV 3 01981 t

7,. C3 i C4 'j O U$ af Table 317-2 GASEOUS PADIOACTIVE P2 LEASES Releaso Pato, 31Ci/sec / Reactor Coolant Concentration Containment Isotope (pCi/n1 0 70 F) Aerated Vents Air Ejector Vent Eccay Drums Total I-131 2.99-1* 1.28-h 1.09-h h.2-h 2.0-6 6.6-h I-132 1.12-1 h.83-5 1 59-6 1 96-6 5.2-5 I-133 5.02-1 2.16-h 5.31-5 79-5 3 5-h i I-13h 7.55-2 3 22-5 1.0-7 h.95-8 3 2-5 j I-135 2.80-1 1.2-h 1.1-5 1.hl-6 1 3-h Kr-85 1.0h 6.56-1 1.2+1 3.h+1 h.7+1 1.2-1 6.27-3 1.26-1 Kr-85n 1.9-1 w Kr-87 1.08-1 6.8-2 1.05-3 6.9-2 i ~ 2.06-1 6.8-3 Y Kr-88 3 26-1 2.1-1 m 7 88-2

2.~76-1 3.6-1 Ze-131m 1.25-1 1.59+1 2.hh+1 h.0+1 Xc-133

2. 52 +1 3 53 -1 h.26-2,

Xc-135 5.60-1 h.0-1

2.99-1 = 2 99 x 10-1 e

F T

3.18 REACTOR COOLANT SYSTEM OXYGEN AND CHLORIDE / FLUORIDE CONCENTRATION Acolicability: Applies to the measured maximum oxygen and chloride / fluoride concentrations in the reactor coolant system. 03jective: To ensure that the oxygen and chloride / fluoride in the reactor coolant system do not exceed concentrations detrimental to the functional integrity of the system materials. Soecification: The concentration of chloride plus fluoride in the reactor coolant system shall not exceed G.15 ppm if the oxygen concentration exceeds 0.1 ppm. Remedial Action: If the oxygen concentration and the chloride / fluoride concentration of the reactor system simultaneously exceed the limits specified above, corrective action is to be initiated immediately and continued power operation is permitted for a maximum of 24 hours. If the system is not brought to within specifications in an additional 24 hour period, the system is to be brought to a cold shutdown condition and the cause of the out-of-specification condition accertained and corrected. Basis: By maintaining the oxygen and chloride / fluoride concentration in the reactor coclant within the limits as specified above, the functional integrity of the material in the Reactor Coolant System is assured under all operatirg conditions. If these limits are exceeded, measures can be taken to correct the condition, e.g., replacement of ion exchange resin or adjustment of the h)drogen concentration in the volume control tank, and further because of the time dependent nature of any adverse effects arising from oxygen and chloride / fluoride concentration in excess of the limits, it is unnecessary to shutdown immediately since the condition can be corrected. Thus the period of 24 hours for corrective action to restore the concentrations within the limits has been established. If the corrective action has not been effective at the end of the 24 hour period, then the reactor will be brought to the hot shutdown condition and the corrective action will continue. If at the end of a further 24 hour period, the corrective action has not been effective, long term corrective action could be recuired and the reactor will.be brought to the cold shutdown condition. 3.18-1 NOV 3 01981

3.19 SAFETY INJECTION SYSTEM Aoplicability: Applies to the condition of safety injection system isolation and loop stop valves. (hjective: To define the condition of the safety injection system isolation and loop stop valves recuired during reactor operation. Soecification: A. The reactor shall not be critical unless the following conditions are met: 1. The safety injection tank isolation valves (SIA-M-11, 21, 31) shall be disabled in the open position. This shall require the following: a. The breakers shall be locked and tagged open. b. The disconnect switches for each valve power operator shall be locked and tagged open. Exception: One safety injection tank isolation valve may be closed for a period of one hour. 2. The loop isolation valves (RC-M-ll, 12, 21, 22, 31, 32) shall be disabled in the open position. This shall require the following: a. The breakers shall be locked and tagged open. b. The disconnect switches for each valve power operator shall be locked and tagged open. Exception: Under degraded loop conditions, the isolation valves l for one loop may be made operable (breakers and disconnect switches closed), following review and approval by the PORC and l l the Plant Manager. A report shall be filed with the NRC within l 7 days describing the reasons for making the valves operable, ( the guidance provided relative to operating the affected valves, and the plans for ultimate restoration of the valves to an inoperable status. 3. The safety injection header isolation valves (HSI-16, 26, 36) shall not be closed. i 4. The following ECCS check valve barriers shall have been determined to be intact in accordance with Technical Specification 4.6. A.2.f. Barrier l Loop 1 a HSI-17 and HSI-61 b LSI-12 3.19-1 NOV 3 01981

Loop 2 a HSI-27 and HSI-62 b LSI-22 Loop 3 a HSI-37 and HSI-63 b LSI-32 Exception: If any of the ECCS check valve barriers specified above do not meet the acceptance criteria of Technical Specification 4.6. A.2.f, then the reactor may be made or remain critical in accordance with the provisions of Specification 4.6.A.2.f. Basis: The position restrictions on the loop isolation valves, safety injection header isolation valves, and the safety injection tank isolation valves are necessary to assure that plant operation is restricted to conditions considered in the loss-of-coolant accident analysis. The exception to. Specification 3.19. A.2 enhances the safety of operations

with a degraded loop component, such as a steam generator tube leak (witbtn soecification) or a degraded reactor coolant pump seal, by ensutby that the affected loop can be isolated rapidly following a postulated plant shutdown necessitated by the degraded component.

The three check valves in the ECCS line to each loop provides assurance that a valve failure will not result in unrestricted flow of pressurized reactor coolant into lower pressure connecting piping outside the containment. The valve integrity testing required by Technical Specification 4.6. A.2.f assures that the rate of flow under a valve failure condition will not exceed the pressure relief capacity of the line. It further provides periodic assurance that the check valves are intact. The two check valves clo est to the loop are grouped together as a single check valve barri+r for test purposes. The first valve provides a thermal barrier preventing thermal distortion from affecting the tightness of the second valve. The third valve alone constitutes a check valve barrier. l In addition to the check. valves the ECCS line to each loop :ontJ'is a l Motor Operated Valve (MOV) which is closed except for pericalc' monthly testing. The MOV and reactor side piping is designed for full system pressure and also capable of preventing an overpressure condition of connecting piping. The exception permits time to schedule an orderly shutdown and maintenance of a defective valve while providing assurance that two separate intact barriers always exist. l '~2 000V 3 0198f

3.20 SHOCK SUPPRESSORS (SNUBBERS) Aoplicability: Applies to those shock suppressors used on the primary coolant system and on other safety related systems or components. Objective: To define tne condition of the above defined shock suppressors required for reactor operation. Soecification: A. During all modes of operation except Cold Shutdown and Refueling, all safety-related snubbers listed on Table 3.20-1 shall be operable except as noted in 3.20.8 and 3.20.C below. B. If any snubber listed on Table 3.20-1 is found to be inoperable, it must be repaired and made operable, or otherwise replaced with one which is operable within 72 hours. C. If the recuirements of specification B cannot be met, an orderly shutdown shall be initiated, and the reactor shall be in the cold shutdown condition within 36 hours. D. If a snubber is determined to be inoperable while the reactor is in

the shutdown or refueling mode, the snubber shall be made operable or replaced prior to reactor startup.

E. Snubbers may be added to safety related systems provided that a revision to Table 3.20-1 is included with the next license amendment. Basis: Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient while al'.owing normal thermal motion during startup and shutdown. The consecuence of an inoperable snubber is an increase in the probability of structural damage to piping as a result of a seismic or other event initiating dynamic loads. It is, therefore, recuired that all snubbers recuired to protect the primary coolant system or any other safety related system or component be operable during reactor operation. Because snubber protection is required only during low probability events, a period of 72 hours is allowed for repairs or replacement. In case a shutdown is required, the allowance of 36 hours to reach a cold shutdown condition will permit an orderly shutdown consistent with standard operation procedures. Since plant startup should not commence with knowingly defective safety related eculpment, Specification 3.20.D prohibits startup with inoperable snubbers. ~ E0V 3 01981

i i! 1 e3 sl : n rbro eioi b sNt b s a yJ uegr ncne ScipArO '5 D.u f o l I ea l mn sbro f i-ioi c 4 esNt e s bs a h b egr X X X X X X X X X X X X X X X X ucne \\ ncip S arO nu I D o yt sl rl t e eal v bi uo bcc= ueic n pf R SsfEi D h

n gaw S

1 eo 1 R rd 1 E At B n u B i nh U oS 1 N ri S et g 0 b an 2 D bii E ud r 3 T

au r

A SRD c L i E b R a T u n t o "0 r i A t . S av '6 e 4 l E 3 1 2 1 1 n 1 E o E i E t n ne a e n G c G e o G m L m r t m S t 'S S r k e' 1 2 3 8 4 S 6 7 1 2 3 _S 4 S 6 7 1 rb an S S S S S S S S S S S S S S S S S Mu S S S S S S S S S S S S S S S S S ~ N j i rer b e bb 1 2 3 4 S 6 7 8 9 0 1 2 3 4 S 6 7 1 1 1 1 1 I 1 1 um . g ;< u o.ec* o nu SN

l

.f i; l 1

1 1 i' l ea sl mn 6b ro cioi b sNt bs a uegr ncne Scip ArO u S D ( i o l ea 7 l mn o sb ro rioi W esNt b s a b egr X X X X X X' X X X X X X X X X X x ucne G ncip S arO nu I D o yt sl rl t e eal v bi uo bccm ueie n pf R Ssf Ei D h n gaw ieo i rd S l R At E n u B i nh B oS 1 U ri N et g 0 S ban bii 2 D ud r 3 E nau T SRD e A 1\\ l L b E a R T "2 6 Y n T o 4 3 1 / / 1 / E i "0 1 S 1 9 7 F t A a S 7 1 S v '9 2 e 6 4 2 l 4 E 8 8 0 4 2 n ) 3 t2 e e o a r n n i 1 lcp ee i i r o zn L) L) e n l o ii 2 3 e o E l L rL g g j i t n a( u np np n a e u sy i o i o I c* G dl s a g o. go i a er rI rL y o L m sv rp a( a( t t eo PS h h e R : C C f n S e a S R 2 1 3 2 1 1 2 4 3 0 0 0 0 0 0 0 0 0 r 1 1 1 1 1 6 6 6 6 ke rb 2 3 8 4 S 6 7 a m S S S S S S S S S l S S I S l S S S I S I S l i I Mu S S S S S S S S I I I I l l l I N S S S S S S S i I l. l 1 l l l. l R Ri C C C l C C __ C i l 1 C I I t I I I l' l r er 8 9 0 1 2 3 4 5 6 7 8 9 0 1 2 3 b e 1 1 2 2 2 2 2 2 2 2 2 2 3 3 3 3 3 bbum . < " O 39

u 2

SN

.) d'f 3*b ~ Tabic 3.20-1 O SAFETY PELATED SNUBBERS ,n Snubber in High Snubbers Snubbcrs

Snubbers, Snubber Ilark Location Elevation Radiation Area Especially Inaccessible Accessible Nur.ber Nu:aber During Shatdown*

Difficult to During Normal During Nor. mal Remove Operation Operation Safaty Injection l X 35 RC-IISS-102 Line (Loop 1) 4'-0" ~ l Safety Injection 36. RC-IISS-103 Line (Loop 2) X X 37 RC-11SS-104 A 38 RC-11SS-104 B X X 39 . RC-IISS-105 ~ Safety Injection 40 RC-IISS-107 Line (Loop 3) 3'-11 S/8" X X 41 RC-IISS-106 X 42 RC-l!SS-108 , Loop Bypass Relief 43 RC-IISS-602 1.ine (Loop 2) 12'-8 3/4" X 41 RC-!!SS-601 15'-8 7/8"- X Fcedwater Pipe to 4S NFPD-!!SS-201 .Stm Gen E-1-1 37'-6" X X 46 h'FPD-IISS-202 Fcedwater Pipe to X 47 NFPD-IISS-206 Stm Gen E-1-2 X 43 NFFD-liSS-205 49 r;i 'H-IISS-203 X -t-I X 50 i XF;'D-11SS-201 Feeduater Pipe to Sin GEN E-t-3 3 + ' - 6" X 51 WFPO-HSS-215 i

o J Tablo 3.20-1 SST[ Ne(( ao SAFETY RELATED SNUBBERS ~ nubber in liigh Snubbers Snubbers Snubbers Snubber Mark Location Elevation Radiation Area Especially Inaccessible Accessible Na:..b er Number Dttring Shutdown

Difficult to During Normal During Norr.a1 Remove Operation Operation Fecdwater Pipe to 52 liFPD-!!SS-214 Stn Gen E-1-3 34'-6" X

53 liFPD-HSS-216 X l 54 l 11FPD-HSS-213 X 55 liFI1-HSS-212 X X 56 [ ift-PD-HSS-211 X 57 liFPD-HSS-210 l 58 11FPD-HSS-209 X j 4 59 11FPD-HSS-208 X* 60 hTPD-IISS-207 X i 61 SHP-HSS-101 Main Stm Linc #1 67'-1" X 62 SHP-HSS-102 X 63 SHP-HSS-108 32'-9" X 64 SHP-HSS-111 47'-7" X 'I 65 SHP-HSS-104 Main Stm Line #2 67-1" X 66 SHP-H35-103 X i X 32'-9" 67 iSHP-HSS-109 L-- X 68 SHP Hss 112 47'-7"

+ l Ili e l c c. , i s i 'i n ~ rbro eioi bsNt X X X X ~ - bs a ,) uegr ncne I. Scip uf Arou D [s e l ~ ca S imn sbro [ rioi o t esNt c bs a i d . s e b egr e v d ucne t ncip S t S arO . i nu m ID bus o e yt b sl rl t e d eal v l bi uo u bccm o ueie h n pf R s SsfEi s D ae r a 1 h n n gaw i co oi S rd R At .' i t E n u aiY B inh B oS d 1 U ri a N et g r 0 S b an )* i bii h 2 D ud r g i 3 E nau , 3 i t T SI D 'h 'e A L. - n 't 6 l b - i. ' n i a R e T s'm. i n ed gn t. o "1 "9 ne t i 7 am

3. ~

r t A a h a S v 7 2 7 c e e 6 3 4 os l E t n e ec ui

d. l.

ets 3 l x _. be an cn i T e sh i_ n i o L it i h t m tf a t c~ S o'o o L e 8 t t r n sa i np a o M l d E ai a t c 'C 5 6 7 0 iR 0 0 0 1 i fN 1 1 1 1 i r d e ke S S S S dh rb S l i l Mt i i l S S S am l Mu i s y N P S i P P P t S S S 1 s r er b e ~ bb 9 0 1 2 um 6 7 7 7 go< y.g nu SN i I

DL ), .a. 2: \\ v ~ 3 o n >M 3.23 FIRES PRO 1D' TION SYSTEMS Aoplicability:. Applies _to the operatirr, stajus of the plant installed fire protection systems. s Cbjective: ?s n To define the operating status of ths' installed fire protection systems. s .t Specification: -1 w A. As a minimum, the. smoke betection instrumentation for each of the protected zones shown in Table 3.234 rhall be opyrable whenever the s eoulpe nt in that zone is required to be operable. y^ =; x. With the number of operable sensors ;in any zone. less than required by . TablF 3.2">-1: 1 , 1. W1. thin 1Mour, establish a roving fire patrol'to' inspect the S affectedczone(s) at least once 'per hour, and s. 2. Restore th^e inoperable instrument (s) to operable status within 14 days, or prepare and submit a Special Report \\to the Commission ^ with'in the~next 30 days outlining the cause of,the malfunction and plans'forsrcstoring the instrument to operable status. Exception: If the affected zone is Zone 5, the ' inspection 'i ' required in A.1 diove. shall be performed at least once per 8 ~ nours. If the affected zone is either Zone 11,(12 or 13, the ' / .insoecti' "Jired In A.1 obove shall be performed at least once ~ g'er 24 and -the RCP purtp bearing, winding and air tempert -3 shall be monitored once per hour.. B. The-fire suppression v<ater system shall be opera lejt all times Nith: two-high pressure pumps each with a capacity cof. 2500 gpm with their. discharge aligned to the fire suppression' tender, and automatic initiation ingic for each pump, and a minimum af K fL. of water in the s fire pend. 1. With less then the;above required equipment, restore the inc@ersb.le equipment to operable status'within 7 days or prepare s and st.bmit a Special Report to the Commission within the next 30 . days outlining theuplans' and procedures to be used to provide for 7 \\ ^ the ' loss 'of redundmcy in this system.y x . a ' x J-2. With no fire sucpres'sion water system opersblei ~ \\' a. Establish a! backup fire suppression water system. Remedial' Action: If 2a above -cannot-be fulfilled in 24 hours, then follow specification 3.0.A. ~ 's-s g 3.'23-1 NOV 3 01981 i

b. Notify the Commission by the telephone or telegraph within 24 hours. Prepare and submit a Special Report to 'he Commission within t c. the next 14 days outlining the cause of the malfunction and T the plans for restoring the system to opercole status, or C. The Cardox system shall be operable, alth a minimum level of 75% and a minimum pressure of 250 psig in the storage tank, whenever the eauipment in the protected areas is required,to be operable. With the Cardox system inoperable: 1. Within 1 hour, establish a roving fire watch to check the affected area (s) at least once per hour, and provide backup fire suppression eculpment for the affected area (s). 2. Restore the system to operable status within 14 days or prepare and submit a Special Report to the Commission within the next 30 days outlining the cause of inoperability and the plans for restoring the system to operable status. D. The fire hose stations in the following locations shall be operable whenever the equipment in the area is recuired to be operable: Hose Hcase #1 -- West side of RCA Storage Bldg. Hose House #2 -- Near Demin. Water Storage Tank. Hose House #3 -- South side of Turbine Bldg. Hose House #6 -- North side of PA8, near Sta. Serv. Transformers. The following fire hose stations shall be operable whenever the component cooling water systerr. or service water system is required to be operable: FS-68, 69, 72, 84, 85----Turbine Building 21 foot olevation FS-81, 82, 83-- - -- Turbine Building 39 foot elevation With a hose station inoperable, route additional equivalent capacity hose to the unprotected area from an operable hose station within 1 hour. E. Penetration fire barriers protecting safety related areas shall be functional at all times. With a penetration fire barrier non-functional, within one hour a continuous fire watch shall be established on at least one side of the affected penetration. F. The following spray and/or sprinkler systems shall be operable whenever the component cooling water system or service water system is recuired to be operable. 1. Turbine Lthe Oil Reservoir Sprinkler ~ NOV 3 01981 3.23-2

2. Seal Oil System Sprinkler 3. Sprinkler System directly under Turbine Building 39 foot elevation. G. The following spray and/or sprinkler systems shall be operable whenever the diesel generators are required to be operable: 1. Turbine Lube Oil Storage Room Sprinkler with one or more of the above required spray and/or sprinkler systems ' innperable: 1. Within one hour establish a fire watd1 patrol with backup fire suppression eculpment for the unprclected area (s) to inspect the unprotected area (s) at 1 cast once per hour. 2. Restore the system (s) to operable status within 14 days or, in lieu of any other report required by Specification 5.9.1, prepare and submit a Special Report to the Commission within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system (s) to operable status. Basis: Operability of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to, safety related eculpment and is an integral element in the overall facility fire protection program. In the event tnat a portion of the fire detection instrumentation is inoperable, the establishment of frecuent fire patrols in the affected areas is recuired to provide detection capability until the inoperable instrumentation is returned to service. The operability of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety related eculpr:,ent is located. The fire suppression system consists of the water system, sprinkler systems, CO, and fire hose stations. The collective 2 capability of the fire suppression systems is adequate to minimize potential damage to safety related equipment and is a major element in the facility fire protection program. In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting eculpment is required to be made available in the affected areas until the affected equipment can be restored to service. In the event that the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides the eajor fire suppression capability of the plant. The requirement for a twenty-four hour report to the Comnission provides for prompt evaluation of the acceptability of the corrective measures to provide adequate fire suppresion capability for the continued protection of the nuclear plant. 20V 3 01981 3 23-3

The functional integrity of the fire barrier penetration seals ensures that fires will be confined or adeauately retarded from spreading to adjacent portions of the facility. This design feature minimizes the possibility of a single fire repidly involving several areas of the facility prior to detection and extinguishment. The' fire barrier penetration seals are a passive element in the facility fire protection program and are subject to periodic inspections. During periods of time when the seals are not functional, a fire patrol is reouired to freauently inspect in the vicinity of the affected seal until the seal is restored to furctional status. 2 f ) ~# DOV 3 01981

c q TABLE 3.23-1 SMOKI DITICTION I!;STRLTENTS a .s MINU LT ZONE LdCATION OPERABLE SI!;S0?.S 1 Service Eidg. Cable Vault 1 2 Protected C.<ble Tfay Ro: 2 3 Unprotected Cable Tray Room 8 4 Contain ent Fenetration & MCC Room (Outside) 2 i 5' containment 7enetratien Room (Inside) 2 6 Protected Switchgear R,ce: 2 7 Unprotected Switchgear Roo 3 8 Diesel Genera:er T.or (D;-1A) 1 9 Diesel Generater Rc== (DO-1B) 1

10 Computer Re::

1 11 Reactor Coolant Pump F-1-1 1 12 Feactor Coolant Pump P-1-2 1 13 Reactor Coolant Pu:p F-1-3 1 O e 9 \\ O S 3.23-5 s 6 l I Acendeent No. 35 NOV 3 0 to** y -g---- ,-y

3.24 SECONDARY COOLANT ACTIVITY Aoplicability: Applies to measured maximum activity in the secondary coolant system. Objective: To ensure that the secondary coolant activity does not exceed a level commensurate with the safety of the plant personnel and the public. Soecification: The specific activity of the secondary coolant systen shall be less than or ecual to 0.10 micro C1/ gram DOSE EQUIVALENT I-131. Remedial Action: If the specific activity of the secondary coolant system exceeds 0.10 micro C1/ gram DOSE EQUIVALENT I-131 be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. Note: The secondary coolant activity surveillance requirements are given in Table 4.2-1, Item 7. Basis: The limitations on secondary system specific activity insure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part.100 limits in a steam line rupture. This dose includes that contributed by a 0.1 gpm primary to secondary tube leak in the steam generator of the affected steam line. t 3.24-1 lNOV 3 019a1

v ATTACHMENT 2 Reason For Change - Maine Yankee Technical Specification Section 3 3.0: Restructured format to be consistent with the rtst of Maine Yankee's Tech. Specs. A: New specification, incorporates requiremant of 10CFR50.36 as a LCO. A2&3: dopts a standard remediel action program for Section 3 which is more restrictive than NRC approved " Standard" technical specifications. Notes 10 CFR 50.72 (a)(5) reporting requirement. ' B: Res'tructured for clarity and modified to be consistent with 3.0. A 2 & 3. 3.1: Restructured and reworded for clarity. 3.2: Restructured for clarity. A2&B: Word "subcritical" substituted for " HOT STANDBY" which by Technical Specification " definition" was inconsistent with less than 5000F. 3.3: Restructured for clarity. A.2: Sentence dropped since a more restrictive tech. spec. applies in Section 3.8. A. C.2: Reworded to better define pressurizer spray flow operability. 3.A: Restructured format. A1: New LCO including remedial action. Prior specification mandated how limits would be derived but did not clearly require operation within limits. B2: References changes to be consistent with the lastructured format. 3.5: Restructured format and reworced to improve clarity. A: Solution temperature not less than 400F, added to be consistent with 3.7.A. C: Remedial Action added with grace period consistent with other LCO's and the CE standard technical specifications. 3.6: Changes to improva clarity and expand scope of LCO's incorporating appropriate remedial actions. 'A2: Substantially changed to better define components within ECCS train and subsystems. 10V 3 0 loa'

o a 4 A3: New LCO added (in place of DG in A2) to better define requirements. (Prior LCO required diesels to be available, but was mute on diesel fuel, MCC's, batteries etc. ) A: Remedial Action added. B: New LCO added reflecting Proposed Change 88. Cenoted by double astericks in the margin). C2&3: Changes to be consistent with A2 and 3 above. Exceptions modified to be consistent with other changes and to provide a grace period consistent with standard specification. 3.7: Restructured Format Remedial Action added. Grace period to meet LCO provides time to inject or delete chemicals as necessary followed by recirculation period to better assure accurate results. 3.8: Restructured Format C: LCO transferred from 3.21. RCS T. Ave. was increased to 2100F to be consistent with the definition of cold shutdown condition. 3.9: Restructured and reworded to improve clarity and be consistent with definitions. A&B: Renedial Actions permits 6 hours to effect repairs prior to achieving hot shutdown. _3.10: Reflects Proposed Change #88. Cenoted by double astericks in the margin). 3.11: Several changes A: New LCO repeated from definition section since definition has been interpeted to be an LCO by I & E. A: Exception permits on line maintenance of containment isolation systems consistent with other LCO's. B: Excepticn included only as a clarification of ' rior wording which was p inconsistent with definition. 01: New LCO mandates that " weight of air system" be used. D2&3: Replaces and expands upon prior LCO clarifying intent and providing remedial action steps in response to I & E concerns. OV 3 019R1

1 e e W -3 3.12: Specification expanded and clarified to be consistent with 3.6. A: New LCO added to be consistent with 3.6A. B: Modified to be consistent with A above and 3.6C. C: Revised.to currently acceptable standards (See IEEE 387/1977 Criteria for Diesel Generator Units). 3.13: The following changes are included: A1: LCO relates to refueling so was transferred from 3.17. Exception permits refueling maintenance and backfit modifications to be conducted without degrading filter medium while providing equivalent public protection. Exception was previously granted on an interim basis, see Ammendment No. 57 dated June 12, 1981. A2: Assures that radiation monitors " remains operable" during fuel handling operations. Exception permits frecuent testing of the monitors without unnecessarily wearing and degrading the large fast acting isolation valves. 3.14: No change. 3.15: No change. 3.16: No change. 3.17: The following changes are included. B7: Portion of LCO that was specific to refueling has been transferred to 3.13. B7: LCO was clarified regarding temporary localized increases in air concentration due to maintenance while assuring that the purge air release limit is within the original intent. 3.18: Wording changes for clarification Changed to correct an error and confusion-that existed throughout the specification related to oxygen and flouride/ chloride concentration limits. The change assures that the simultaneous limit of 0.15 ppm applies to chloride plus flouride. [V 3 01981 -}}

ML20033C605

Text

Navigation menu

Search