Text

Submittal of Changes to Technical Specifications Bases
	ML19106A161
Person / Time
Site:	Three Mile Island
Issue date:	04/12/2019
From:	David Helker; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TMl-19-042
	Download: ML19106A161 (228)
	v • d • e

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v Exelon Generation 200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com TS 6.18.d TMl-19-042 April 12, 2019 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRG Docket No. 50-289

Subject:

Submittal of Changes to Technical Specifications Bases In accordance with the requirement of Three Mile Island Nuclear Station (TMI), Unit 1 Technical Specification 6.18.d, Exelon Generation Company, LLC hereby submits a complete updated copy of the TMI, Unit 1 Technical Specifications and Bases. The enclosed Technical Specifications and Bases include changes through the date of this letter.

If you have any questions or require further information, please contact Frank J. Mascitelli at 610-765-5512.

Sincerely, David P. Helker Manager, Licensing and Regulatory Affairs Exelon Generation Company, LLC

Enclosure:

Three Mile Island Nuclear Station, Unit 1 Technical Specifications and Bases cc: USNRC Administrator, Region 1 (w/o enclosure)

USNRC Senior Resident Inspector, TMl-1 (w/o enclosure)

USNRC Project Manager, TMl-1 (w/ enclosure)

R.R. Janati, Pennsylvania Bureau of Radiation Protection (w/o enclosures)

COPY

Bases 7_';_

.Section 1.25 establishes the limit for which the sp~cified time interval for Surveillance-Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consid~ration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or

. other ongoing*surveillance or maintenance activities. It also provides flexibility to accommodate the length of a fuel cycle for surveillances that are specified to be performed at least once each REFUELING INTERVAL. It is not intended that. this provisjon be used repeatedly as a convenience to extend survei 11 ance intervals beyond that specified for surveillances that are not performed once each REFUELING INTERVAL Likewise, it is not the intent that REFUELING INTERVAL surveillances be performed during power operation unless it is consistent with safe plant operation. The limitation of Section 1.25.is based on engineering judgement and the recognition that the most probable result of any particular .surveillance being* performed is the verification of conformance with the Surveillance Requirements. This provi~ion-is sufficient to ens-ure that the reliability ensured through

. surveillance activit~es is not significantly degraded beyond that obtained from the specified surveillance interval.

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1-9 Amendment No. ii, ttt, i56, (ii, 175

SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.

2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, axial power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the plant.

Objective To maintain the integrity of the fuel cladding.

Specification 2.1.1 The combination of the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points established in Figure 2.1-1.

If the actual pressure/temperature point is below and to the right of the line, the safety limit is exceed_ed.

2.1.2

The combination of reactor thermal power and axial power imbalance (power in the top half of core minus the power in the bottom half of the* core expressed as a percentage of the rated power) shall not exceed the protective limit as defined by the locus of points (so.lid line) for the specified flow set forth in the Axial Power Imbalance Protective Limits given in the Core Operating Limits Report (COLA). If the actual-reactor- thermal-power/axial-power-imbalance point is above the line for the specified flow, the protective limit is exceed~d.

Bases To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under *normal operating conditions. This is accomplished by operating within the nucleate boiling regirne of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coplant temperature. The upper boundary of the nucleate boiling regime is termed, departure from nucleate boiling (DNB). At this point there is a sharp reduction of the heat transfer coefficient, which could result in excessive cladding temperature and the possibility_ of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of a critical heat flux (CHF) correlation.

The BHTP (Reference 1) and BWC (Reference 2) correlations have been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The BHTP correlation applies to Mark-8 fuel with zircaloy or M5 HTP intermediate spacer grids and the BWC correlation applies to Mark-B fuel with zircaloy or M5 intermediate spacer grids (non-mixing vane). The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, accounting only for DNBR correlation uncertainty, during steady-state operation, normal 2-1 Amendment No.17, 142,157,184,233,247,26 2

GENERAL CONTROLLED COPY operational transients, and an~icip.ated transients is limited to 1.132 (BH°TP) and 1.18 (BWC).

Correspondin'g Statistical Design Limits account for all uncertainties considered with the stiltistical core design methodology (Reference 4). A DNBR of 1.132 (BHTP) or 1.18 (BWC) corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core *outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits.

(

The curve presented in figure 2.1-1 represents the conditions at yvhich the minimum allowable DNBR or greater is predicted for the limiting combination of thermal power and number of operating reactor coolant pumps .. This curve is based on the nuclear power peaking factors given in Reference 3 and the COLR which define the reference design peaking condition in the core for operation at the maximum power.* Once the reference peaking condition and the associated thermal-hydraulic situation has been established for the hot channel, then all other combinations of axial flux shapes and their accompanying radials must _result in a condition which will not violate the previously established design criteria on DNBR. The flux shapes examined include a wide range of positive and negative offset for steady state and transient conditions.

These design limit power peaking 'factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion, and form the core DNBR design basis.

The Axial Power Imbalance Protective Limits curves in the COLR are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod bowing:

a. The DNBR limit produced by a total nuclear power peaking factor consisting of the combination of the. radial peak, axial peak, and position of the axial peak that yields no less than the DNBR limit.

b. The maximum allowable local linear heat rate that prevents central fuel melting at the hot spot as given in the COLR.

Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the axial power imbalance produced by the power peaking.

2-2 Amendment No. 17, 60, 90, 126, 142, 167, 184, 247, 262

GENERAL CONTROLLED COPY The specified flow rates for curves 1, 2, and 3 of the Axial Power Imbalance Protective Limits given in the COLR correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.

The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3. The curves of Figure 2.1 ~3 represent the conditions at which the .DNBR limit is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operqtion or the locai qu~lity.at the point of minimum DNBR is equal to the maximum quality limit approved for the applicable CHF correlation;whichever *I condition is more restrictive. Th~ curves of Fig1,1res ~.1 ~ 1 and 2.1-3 were developed assuming a reactor coolant design flow rate of 102% of 352,000 gpm.

The maximum thermal power for each reactor coolant pump operating condition (four pump, three pump, and one pi.Jmp in each loop) given in the COLR is due to a power level trip produced by the flux-flow ratio multiplied by the minimul)1 flow rate for the given pump combination plus the maximum calibration and instrumentation error.

Using a local quality limit at the point of minimum DNBR as a basis for curves 2 arid 3 of Figure 2.1-3 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.

The DNBR as calculated by the BHTP or BWC correlation continually increases from the point of minimum DNBR, so that the exit DNBR is always higher and is a function ofthe pressure.

For each curve of Figure 2.1-3, a pressure;..temperature point above and to the left of the curve would result in a DNBR greater than the Statistical Design Limit (SOL) or a local quality at the point of minimum DNBR less than the maximum quality limit approved for the applicable CHF correlation, for the particular reactor coolant pump situation. Curve 1 is more restrictive than any other .reactor coolant pump situation because any pressure/temperature point above and to the left of this curve will be above and to the left of the other curves.

REFERENCES (1) BHTP DNB Correlation Applied with LYNXT, BAW-10241 P-A, Framatome ANP, Inc.,

Lynchburg, Virginia, July 2005.

(2) BWC Correlation of Critical Heat Flux, BAW-10143P-A, Babcock & Wilcox, Lynchburg, Virginia, April 1985 (3) UFSAR, Section 3.2.3.1.1.3 - "Nuclear Power Factors" (4) BAW-10187 P-A, "Statistical Core Design For B&W-Designed 177 FA Plants," B&W Fuel Company, Lynchburg, Virginia, March, 1994.

2-3 Amendment No. 17, 29, 39, 50, 120, 126, 142, 150, 157, 184, 238, 247, 262

CONTROl ! ID COPY t:.;7')

2.2 SAFETY LIMITS - REACTOR

.~/ SYSTEM PRESSURE.

App1 i cabi 1 i ty Applies to the lim it on rea cto r coolant system pressure Objective To maintain the in teg rit y prevent the rel ea se of sig of the rea cto r coolant system and to ac tiv ity . ni fic an t amounts of fis sio n product Sp ec ifi ca tio n 2.2 .1 The re ac to r coolant system pressure sh all not exc psig when there are fuel* assemblies in the rea ctoeed 2750 r ve sse l.

Bases The re ac to r coolant system preven~ rad ion uc lid es in the(Reference 1) serves as a ba rri er to atmosphere. In the event rea cto r coolant from reaching coolant system is a ba rri er of a fuel cladding fa ilu re , the the ag rea cto r Es tab lis hin g a system pressu ain st the rel ea se of fis sio n products.

the re ac to r coolant system re lim it helps to assure the in teg rit y of in the re ac to r coolant system . The maximum tra ns ien t pr pressure vessel under the essure allowable Section II I, is llO t of de ASME Code, tra ns ien t pr es su re allowabsign pressure (Reference 2) . The maximum valves, and fit tin gs un_der le in the rea cto r coolant system piping, pr es su re. Thus, the sa fet y ANSI Section 831.7 is 110% of design design pr es su re) has been lim it of 2750 ps ig (110% of the 2500 psig es tab lis he d (Reference 2) .

se tti ng s fo r the rea cto r hig The maximum pr es su riz er code sa fet y va h pressure tri p (2355 ps1g) and the accordance with ASME Boile lves (2500 psig) have been establ1shed in Ar tic le 9, Winter, 1968 to r and Pressufe Vessel Code,Section II I, pressure sa fe ty lim it is noassure th at the rea cto r coolant system was conducted at 3125 ~sig t exceeded. The in it. ial hy dr os tat ic te st in teg rit y of the react*or co(125% of design pressure) to ve rif y the the re ac to r co ola nt system olant system. Add1tional assurance th at is provided by the presence pressure does not exceed the sa fe ty lim it va1ve {Reference 3) . of a pr es su riz er ele ctr om ati c re lie f

/

(References (1) UFSAR, Section 4.0 - "Reacto r Coolant System" (2) UFSAR, Section 4.3.10

- "Safety Limits and Conditio ns"

{~) UFSAR, Table 4.2 "R eactor Coolant System Press ure Se tti ng s"

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2-4 Amendment No. tl, 2S, li, 4S, qa, ilS, 157

r GENE RAL CONTROLLED COPY I

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2300 - - - - - - - - - - - ,_ - - - - - - - - - - _,_ - - - - - - - - - - - - - - - - - - - - - "' - - - - - - - - - - - ~ - - - - - - - - - - -

ACCEPTABLE OPERATION

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"[ 2100 ---------- L- -------- ---------~-- ---------~-- ---------

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UNACCEPTABLE OPERATION I

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1500 ..J.,--- ------- ~------ ----~-- --....-- --~--

590 600 610 620 630 640 650 Reactor Outlet Temperature (°F)

CORE PROTECTION SAFETY LIMIT TMl-1 FIGURE 2.1-1 2-4a Amendment No. 60, 142, 167, 214, 238, 247, 262

INFOR MATIO N ON THIS PAGE HAS BEEN DELET ED

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2-4b Amend ment No. 17, 29, 39, 45, 60, 120, 126, 142, 167, 184 278

GENERAL CONTROLLED COPY I

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- - - - - - - - - II II '

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1500 ...J-----r----.---'---,------'T----,------'-~--;

590 600 610 620 630 640 650 Reactor Outlet Temperature (°F)

Pump Operation -+- 3-Pump Operation _...,..2-Pump Operation RC Pumps Reactor Coolant Flow Power Pumps Operating (Type of Limit)

. (gpm) 4 359,040 112% Four Putnps (DNBR Limit) 3 SeeCOLR SeeCOLR Three Pumps (DNBR Limit) 2 SeeCOLR SeeCOLR One Pump in Each Loop (DNBR Limit)

CORE PROTECTION SAFETY BASES TMl-1 FIGURE2.1-3 2-4c Amendment No. 50, 126, 142, 167, 184, 214, 238, 247, 262

CONTROU ED COPY 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTION INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, axial power cool~nt system pressure, reactor coolant outle t temperat~re, imbalance, reactor pumps in operation, and high reactor building pressure .. flow, number of Objective To provide automatic protection action to prevent any combinatio variables from exceeding a safety limit . n of process Specification 2 .3. I The reactor protecti6n system trip setti ng limit s and the perm bypasses*fqr the instrument channels shall be as stated in Tableissible and. the Protection System-Maximum Allowable Setpoints for Axial 2.3-1 Power Bases Imbalance as given in the COLR.

I The react or protection system consists of four each of several selected plant conditions which instru will ment channels to monitor one of these conditions deviates from a pre-selected cause opera a reactor trip if any ting degree that a safety limit may be reached. range to the The trip setti ng limit s for protection system Table 2.3-1. These trip setpoints are setti ng instru limit mentation are.l isted in s on the setpoint side of the protection system bista ble comparators. The safet upon these protection system instrumentation trip set ypointanalysis has been based s plus calib ratio n and instrumentation error s.

Nuclear Overpower A react or trip at high power level (neutron flux) to the fuel cladding from reac tivity excursions toois rapid provided to prevent damage to be detected by pressure and temperature measurements.

During normal plant operations with all react or coolant pumps trip is initi ated when the reactor power level reaches 105.1% operating, react or Adding to this the possible variation in trip set points due of rated power.

instrument error s, the maximum actual power at which a trip to calib ratio n and could be 112", which is the value us*ed in the safety analysiswould be actuated (Reference 1).

2-5

Amendment No. 11, 17, iJ, JiJ, 1Ji, JJ7, 184,

GEN ERA L CON TRO LLE . . . D COP Y t'"!fn;~\

Jj

a. Overpower trip based on flow and imbalance The power level trip set point produced by the reactor coolant system flow is

. based on a power-to~flow ratio which has been establish,ed to accommodate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power, Analysis has demon*strated that the specified power to flow ratio is adequate to prevenfa DNBR bfless tt,an the Statistical Design Limit should. a low flow condition exist due to ~my malfunction.

The power level trip set pornt produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump operation.* For.every flow rate there is a maximum permissible pbvver level, and for every power level there. is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.3-1 are given in the COLA.

The fiux/flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives *a conservative indication of the RC flow.

No penalty in reactor coolant flow through the core was take'n for an open core vent valve because of the core vent valve surveillance program during each refueling outage:

For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limit.s are either power peaking Kw/ft limits or DNBR limits.

The axial power imbalance (power in the top halt of the core minus power in 2-6 Amendm ent No. 13, 17, 25, 28, 39, 50, 126, 142, 157, 184, 247, 262

GENE RAL CONTROLLED COPY

/ 7~::\ the l:mttom half of core) reduces the power level trip produceq _by the power-to-flow ratio so that

' l the boundaries of the Protection System Maximum Allowable Setpoints for Axial Power Imbalance in the COLA are produced.

b. Pump Monitors The redunda.nt pump monitors prevent the minimum core DNBR from decreasing below the Statistical Design Limit by tripping the reactor due to the loss of reactor coolant pump(s). The pump monitors also restrict the power level for the number of pumps in operation.

c. Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip setpoint is reached before the n.uclear overpower trip setpoint. The trip setting* limit shown in Figure 2'.3-1 for high reactor coolant system pressure ensures that the sys~em pressure is maintained below the safety limit (2750 psig) for any design transient (Reference 2). Due to calibration ar\d instrument errors, tne safety analysis assumed a 45 psi pressure error in the high reactor coolant system pressure trip setting.

As part of the post-TM1.:2 accident modifications, the high pressure trip setpoint was lowered from 2390 psig to 2300 psig. (The FSAR Accident Analysis Section still uses the 2390 psig high pressure trip setpoint.) The lowering of the high pressure trip setpofnt.and raising of the setpoint for the Power Operated Relief Valve (PORV), from 2255 psig to 2450 psig, has the effect of reducing the challerige_rate tq the PORV while maintaining.ASME Code Safety Valve

.* capability.

A B&W analysis completed in September of 1985 concluded that the high reactor coolant system pressure trip setpoint could be raised to 2355 psig with negligible impact on the fr13qu_ency of opening of the PORV during anticipated over-pressurization transients (Reference 3). The high pressure trip setpoint was subsequently raised to 2355 psig. The potential safety benefit of this action is a reduction in the frequency of reactor trips.

The low pressure and variable low pressure trip setpoint were initially established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction (References 4, 5, and 6). The B&W generic EGGS analysis, however, assumed a low pressure trip of 1900 psig and, to

establish conformity with this *analysis, the low pressure trip setpoint has been raised to the more conservative 1900 psig. The revised low pressure trip of 1900 psig and the variable low pressure (16.21 Tout - 7973) trip setpoint prevent the minimum core DNBR from decreasing below the Statistical Design Limit. Figure 2.3-1 shows the high pressure, low pressure, high temperature and variable low pressure trip setpoints.

2-7 Amendment No. 17, 28, 39, 45, 78,126,135, 142,157,184 ,247,262

CONTROLLED CO PY

d. Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (618.BF) shown in Figure 2.3-1 has been established to prevent exc:essive core coolant temperature in the operating range.

The calibr~ted range of the temperature channels of the RPS is 520° to 620°F.

The trip setpo int of the channel is 618.BF. Under the worst case environment, power supply perturbations, and drift, the accuracy of the trip string is_ 1.2°F. This accuracy was arrived at by surnming the worst case accur acies of each modu le.

This is a conservative method of error analysis since the normal procedure is to use the root mean square-method.

Therefore,- it is assured that a trip will occur at a value nb higher than 620°.F even under worst case conditions. The safety analysis used a high temperature trip

. set point of 620°F.

The calibrated range of the channel is that portion of the span of indication which has been. qualified with rega.rd to drift, linearity, repeatability

, etc. This does not imply that the equipment is restricted to operation within the calibrated range .

. Additional testing has demonstrated that iri fact, the tempe rature channel is fully

. operational approximately 1_0% above the calibrated range

Since it has been established that the channel will trip at a value of RC outlet temperature no higher than 620°F even in the worst case, and since the channel is fully operational approximately 10% above the calibrated range and exhibits no hysteresis or foldover characteristics, it is concluded that the instrument design is acceptable. * *

e. Reactor building pressure \

The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

2-8

0} Amen dmen t No. 45, 78, 135, 142, 24 7

CONTROIJ ED COPY

~:),~ f. Shutdown bypass

)

In order to provid physics te sti ng s, eandforstacortu ntrol rod drive te st s, zero power for bypassing ce rta in segme p proceduies, th er e is provision system; The re ac to r prote nts of the reactor .prot~ction bypassed are shown in Ta ction system segments which can be when the by~ass is used: ble 2.3-1. Two conditions are imposed

1. By administrative contr point must be reduced tool va the nuclear overpower tri p during re ac to r shutdown. lu e< 5.0 percent of rated se t

-

power

2. A high re~ctor coolant 1720 psig is automaticasys tem ~ressure tri p se t point lly imposed. of The p~rpose of the 1720 ps to prevent normal operatio ig high pressure tri p se t point is pr ot ec tio n sy~tem bypassed.n with ~a rt of the re ac to r is lower than the normal low This high pr~ssure tri p se t point the re ac to r must be tripped pressure tri p se t point so th at The overpower tri p se t po before the bypass is in iti at ed .

. sig ni fic an t re ac to r power int of < 5.0 percent prevents any the physics te st s. Su ffi cie from being produced when pe rforming

~vai1able to remove 5.0 pe nt natural cir cu lat io n would be re ac to r coolant pumps were rcent of ra ted power if none of the ly ~) operating.

/ References' (1) UFSAR, Section 1. 4. 6 -

"Criterion 6 - Reactor Core Design" (2) UFSAR, Section 14 .1 .2 .2

~ "Startup Acciden t"

(3J "J us tif ic at io n fo r Ra Pr es su re ," BAW-1890 Re ising Setdoint for Reactor 1 v. 0, Babcock and Wilcox, Trip on High September 1985.

(4) UFSAR, Section 14 .l. 2.

Control Rod Accid*nt" 7 - "Stuck-Out, Stuck-In, or Dropped (5) UFSAR, Section 14 .1 .2 .9

- "Steam Line Break" (6) UFSAR, Section 14.3 Re 177-FA Lowered Loop 1 NNS," ference 28 - "ECCS Analysis of B&W's Wilcox, Lynchburg, Virgin 8AW~l0103-A, Rev. 3 1 Babcock and ia, July 1977.

(7} UFSAR, Section 14 .1 .2

.6 - "Loss of Coolant Flow" 2-9 Amendment No. YJ', t1, 28, ai, 4o, 1B, 90, t26, 1!6, !4l. 157

';>*/p~-~ l

' TA REAC TOR PROTECTION SYSTEv'I TRIP SETTING LIMIT S (5)

Four React or Coola nt Three Reactor Coola nt One React or Coola nt Pump s Opera ting Pumps Oper,1ting. . Pump Opera ting in*

(Nomi nal Opera ting) (Nominal Qpe1*ating) Each Loop (Nomi nal

Shutd own Powe r-100 % Power - 7~:%, Opera ting Power - 49%) Bypas s

1. Nuclea r power, max.% 105.1 of rated power 105.1 105.1 5.0(2)

2. f\Juclear power based on Power /Flow Setpo int in Power /Flow Setpoi nt in Power /Flow Setpo int in flow (1) and imbala nce COLA times flow minus .Bypa ssed COLA times flow minus COLA times flow minus .

max. of rated power reduction due to reduction due to reduction due to imbala nce imbala nce imbala nce

3. Nuclea r power based (4) NA*

on pump monito rs max. NA 55% Bypas sed

% of rated power

4. High reacto r coolan t 2355 system pressu re, psig 2355 2355 1720(3 )

max.

5. Low reactor coolant 1900 1900 1900 Bypas sed system pressu re, psig min.

6. React or coolan t temp. F.,

618.8 618.8 max. 618.8 618.8

7. High React or Building 4 4 pressu re, psig max. 4 4

8. Variab le low reacto r (16.21 Tout - 7973)( 6) (16.21 Tout - 7973)(6) (16.21 Tout - 7973)( 6) Bypas sed coolan t system pressure, psig min.

( 1) React or coolan t system flow, %

(2) Admin istrativ ely contro lled reduct ion set during reactor

. shutdo wn.

(3) Autom aticall y set when other segme nts of the RPS (as specified) are bypassed.

(4) The pump monito rs also produ ce a trip on: (a) loss of two reactor.coolant pumps in one reactor coolan t loop, o*r two reacto r coolan t pumps during two-pu mp operation. and {b) loss of one (5) Trip settings limits are limits on the setpoi nt side of the protection system bistable connectors.

(6) Tout is in degree s Fahre nheit (F).

2-10 Amen dment No. 45, 78, 90, 126, 1aa, 142,1 84,24 7,262

GENERAL CONTROLLED COPY 2500 i '. - 1*-----. *1 I

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i UNACCEPTABLE' OPERATION I

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I I I I I I II I I I I I

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I I I 1500 540 I

560 I

i I I 580 600 620 640 Reactor Outlet Temperature, °F PROTECTION SYSTEM MAXIMUM ALLOWABLE SETPOINTS TMl-1 FIGURE 2.3-1 2-11 Amendment No. 13, 17, 28, 30, 16, 78,126,135, 142,167,247 ,262

INFORMATION ON THIS PAGE HAS BEEN DELETE D

2-12 Ame ndm ent No. 17, 29, 39, 40, 45, 50, 120, 126, 142, 167, 184 278

SECTION 3.0 LIMITING CONDITIONS FOR OPERATION

3. LIMITING CONDITIONS FOR OPERATION 3.0 GENERAL ACTION REQUIREMENTS 3.0.1 When a Limiting Condition for Operation is not met, except a~

provided in action called for in the specification, within one hour action shall be initiated to place the unit in a condition in which the specification does not apply by placing it, as applicable, in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the action requirements, the action may be taken in accordance with the time limits of the specification as measured from the time of failure to meet the Limiting Condition for Operation. Applicability of these requirements is stated in the individual specifications.

Specification 3.0.1 is not applicable in COLD SHUTDOWN OR REFUELING SHUTD OWN.

BASES This specification delineates the action to be taken for circumstances not directly provided for in the action requirements of individual specifications and whose occurrence would violate the intent of the specification.

The NRC approved TS Amendment 98 on 08/07/84 which incorporated Genera l Specification 3.0.1 above. The TMI amendment request and NRC approval specified that this General Specification was incorporated only where it was determined to specifically apply by stating "specification 3.0.1 applies."

3-1 Amendment No. 98, AR 4048309

3.1 REACTOR COO LAN T SYSTEM 3.1.1 OPERATIONAL COMPONENTS Applicability Applies to the operating status of reactor coolant system components.

Objective To specify those limiting conditions for oper ation of reactor coolant system components must be met to ensure safe reactor operation which s.

Specification 3.1.1.1 Reactor Coo lant Pumps

a. Pump combinations permissible for given power levels shall be a~ shown in Specification Table 2.3.1.

b. Pow er operation with one idle reactor cool ant pump in each loop shall be restricted to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the reactor is not returned to an acceptable RC pum p operating combination at the end of the 24-hour period, the reactor shall be in a hot shutdown condition within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

C. The boron concentration in the reactor cool ant system shall not be reduced unless at least one reactor coolant pump or one decay hea t removal pump is circulating reactor coolant.

3.1.1.2 Steam Generator (SG) Tube Integrity

a. Whe nev er the reactor coolant average tem perature is above 200°F, the following conditions are required:

(1.) SG tube integrity shall be maintained.

(2.) All SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Prog ram. (The Steam Generator Program is described in Section 6.19.)

ACTIONS:

NOTE--

Entry into Sections 3.1.1.2.a.(3.) and (4.),

below, is allowed for each SG tube.

(3.) If the requirements of Section 3.1.1.2.a.(2

.) are* not met for one or mor e tubes then perform the following:

3-1a Amendment No. 12, 17, 28, 47, 98, 261 278, 279

With one or more SG tubes satisfy ing the tube plugging criteria and not plugged in accordance with the Steam Generator Program:

a. Verify within 7 days that tube integrity of the affected tube(s) is maintained until the next refuelin g outage or SG tube inspection AND ,

b. Plug the affected tube(s) in acc ordance with the Steam Ge ner ato Program prior to exceeding a rea r ctor coolant average tem per atu of 200°F following the next refuelin re g outage or SG tube inspection

. (4.) If Action 3., above, is not completed within the specified com times, or SG tube integrity is not pletion maintained, be in HOT SH UT DO within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in COLD SH WN UTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

3.1.1.3 Pressurizer Safety Valves

a. The reactor shall not remain crit ical unless both pressurizer cod valves are operable with a lift set e safety ting of 2500 psig +/- 1%.

b. When the reactor is subcritical, at least one pressurizer code saf ety shall be operable if all reactor coo valve lant system openings are closed except for hydrostatic tests in acc ,

ordance with ASME Boiler and Vessel Code, Section Ill. Pressure 3-1b

~* Amendment No. 2&1-, 279

The limitation on power operation with one idle RC pump in each loop has been imposed ECCS cooling performance has not been calcu since the lated in accordance with the Final Acceptan Criteria requirements specifically for this mod ce e of reactor operation. A time period of 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed for operation with one idle RC pump s is in each loop to effect repairs of the idle pum to return the reactor to an acceptable combina p(s) and tion of operating RC pumps. The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for mode of operation is acceptable since this mod this e is expected to have considerable margin for peak cladding temperature limit and since the the likelihood of a LOCA within the 24-hour perio considered very remote. d is A reactor coolant pump or decay heat remo val pump is required to be in operation before concentration is reduced by dilution with-mak the boron eup water. Either pump will provide mixing will prevent sudden positive reactivity changes which caused by dilute coolant reaching the reactor.

decay heat removal pump will circulate the One equivalent of the reactor coolant system volum one-half hour or less. e in

The decay heat removal system suction pipin g is designed for 300°F and 370 psig; thus, can remove decay heat when the reactor cool the system ant system is below this temperature (Referen and 3). ces 1, 2, Management of gas voids is important to DHR System OPERABILITY.

Both steam generators must have tube integ rity before heatup of the Reactor Coolant Syst to insure system integrity against leakage unde em r normal and transient conditions. Only one*

steam generator is required for decay heat removal purposes. Refer to Section 3.1.6.3 allowable primary-to-secondary leakage. Refe for r to Section 4.19 for Bases for Steam Generato tube integrity. r One pressurizer code safety valve is capable of preventing overpressurization when the reac not critical since its relieving capacity is grea tor is ter than that required by the sum of the avail sources which are pump energy, pressurizer able heat heaters, and reactor decay heat. Both pres safety valves are required to be in service prior surizer code to criticality to conform to the system design capabilities. The code safety valves prevent relief overpressure for a rod withdrawal or feedwate break accidents (Reference 4). The pressuriz r line er code safety valve lift set point shall be set psig +/-1 % allowance for error. Surveillance at 2500 requirements are specified in the INSERVIC PROGRAM. Pressurizer code safety valve E TES TING setpoint drift of up to 3% is acceptable in acco with the assumptions of the TMl-1 safety anal rdance ysis (Reference 5).

References (1) UFSAR, Tables 9.5-1 and 9.5-2

{2) UFSAR, Sections 4.2.5.1 and 9.5 * "Decay Hea t Removal"

{3) UFSAR, Section 4.2.5.4 - "Secondary System"

{4) UFSAR, Section 4.3.10.4- "System Minimum Operational Components" (5) UFSAR, Section 4.3.7 * "Overpressure Prote ction" 3-2 Amendment No. 4+ (1m 2/7B ),+a 7,~. 26:t -.2e i, asa, 290

3.1.2 PRESSURIZATION HEATUP AND COOLDOW N LIMITATIONS Applicability Applies to pressurization, heatu p and cooldown of the reactor coolant system.

Objectives To assure that temperature and pressure chang es in the reactor coolant system do not caus e loads in excess of design for reactor coolant syste cyclic m components.

To assure that reactor vessel integrity by maint aining the stress intensity factor as a result of operational plant heatu p and cooldown conditions and inservice leak and hydro test conditions below values which may result in non-ductile failure.

Specification 3.1.2.1 For operations until 50.2 effective full powe r years

, the reactor coolant pressure and the system heatup and cooldown rates (with the excep tion ofthe pressurizer) shall be limited in acco rdanc e with Figures 3.1-1, 3.1-2, and 3.1-3 and are as follows:

Heatup/Cooldown Allowable combinations of pressure and temperatu re shall be to the right of and below the limit line in Figures 3.1-1 and 3.1-2.

Heat up and cooldown rates shall not exce ed those shown on Figures 3.1-1 and 3.1-2. When the core is critical, allowable combinations of pressure and temperatu re shall be to the right of the criticality limit curve shown on Figure 3.1-1.

lnservice Leak and Hydrostatic Testing .

Allowable combinations of pressure and temperatu re shall be to the right of and below the limit line in Figure 3.1-3. Heatup and cooldown rates shall not exce ed those show n on Figure 3.1-3.

3.1.2.2* The seco ndary side of the steam generator shall not be pressurized above 200 psig if the temp eratu re of the steam generator shell is below 100°F.

3.1.2.3 The press urize r heatup and cooldown rates shall not exceed 100°F in any one hour.

The spray shall not be used if the temperature difference between the press urize r and the spray fluid is greater than 430°F.

3.1.2.4 DELE TED 3.1.2.5 DELE TED

~ 3-3

~ Amendment No. 29, 1:34,176, 208,2 34 278,2 81

Bases All reactor coolant system components are designed to withstand the effects of cyclic loads due to system temperature and pressure changes (Reference 1). These cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations.

The number of thermal and loading cycles used for design purposes are shown in Table 4.1-1 of the UFSAR. The maximum unit heatup and cooldown rates satisfy stress limits for cyclic operation (Refere nce 2). The 200 psig pressure limit for the secondary side of the steam generator at a tempera ture less than 100°F satisfies stress levels for temperatures below the Nil Ductility Transition Temper ature (NOTT):

The heatup and cooldown rate limits in this specification are based on linear heatup and cooldown ramp rates which by analysis have been extended to accommodate 15°F step changes at any time with the appropriate soak (hold) times. Also, an additional temperature step change has been included in the analysis with no additional soak time to accommodate decay heat initiation at approximately 240°F indicated RCS temperature.

The unirradiated reference nil ductility temperature (RTNor) for all Linde 80 welds were determined in accordance with BAW-2308, Rev. 1-A and Rev. 2-A,. and 10 CFR 50, Appendices G and H. For the beltline plate and forging materials, 10 CFR 50 Appendices G and H were used to calculate the unirradiated reference nil ductility temperature (RT Nor). For other beltline region materials and other reactor coolant pressure boundary materials, the unirradiated impact properties were estimated using the methods described in BAW-10046A, Rev. 2.

As a result of fast neutron irradiation in the beltline region of the core, there will be an increase in the RT NOT with accumulated nuclear operations. The adjusted reference tempera tures have been calculated as described in Reference No. 5.

The predicted RTNOT was calculated using the respective predicted neutron fluence at 50.2 effective full power years of operation .and the procedures defined in Regulatory Guide 1.99, Rev. 2, Section C.1.1 for the plate metals and for the limiting weld metals (WF-70, WF-8, and SA-1526).

Analyses of the activation detectors in the TMl-1 surveillance capsule s have provided estimates of reactor vessel wall fast neutron fluxes for cycles 1 through 17. Extrapo lation of reactor vessel fluxes and corresponding fluence accumulations, based on predicted fuel cycle design conditions through 50.2 effective full power years of operation are described in Reference 6 with effective full power years clarified in Reference 7.

3-4 Amend ment No. 29, 134, 167,17 6,208,2 34,281

Based on the predicted RT Nor after 50.2 effective full power years of operation, the pressure/

temperature limits of Figures 3.1-1, 3.1-2, and 3.1-3 have been established by AREVA calculation, Reference No. 8, in accordance with the requirements of 10 CFR 50, Appendix G. The methods and criteria employed to establish the operating pressure and temperature limits are as described in BAW-10046A, Rev. 2 and ASME Code Section XI, Appen dix G, as modified by ASME Code Case N-640 and N-588. The protection against nonductile failure is provided by maintaining the coolant pressure below the upper limits of these pressure/temperatu re limit curves. The minimum temperature for core criticality is determined to satisfy the regulatory require ments of 10 CFR Part 50, Appendix G.

This limit is shown on Figure 3.1-1.

The pressure/temperature limit curves in Figures 3.1-1, 3.1-2, and 3 ..1-3 have been established considering the following:

a. System pressure is measured in RCS "A" loop hot leg. RCS "A" is most conservative and bounds use of "B".

b. Maximum differential pressure between the point of system pressure measurement and the limiting reactor vessel region for the allowable operating pump combinations.

The spray temperature difference restriction, based on a stress analysis of spray line nozzle is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.

Temperature requirements for the steam generator corres pond with the measured NOTT for the shell.

REFERENCES (1) UFSAR, Section 4.1.2.4 - "Cyclic Loads" (2) ASME Boiler and Pressure Code, Section Ill, N-415 (3) DELETED (4) DELETED (5) AREVA Docum ent No. 32-9035343, "ART Values for Three Mile Island at 60 Years."

(6) AREVA Document No. 86-9162844, "TMl-1 Cycles 16 and 17 Fluence Analysis Summ ary Report."

(7) AREVA Document No. 51~9035228, "TMl-1 Reactor Vesse l Embrittlement Limits at 60 Years."

(8) AREVA Document No. 32-9177176, "TMl-1 Corrected P-T Limits at 50.2 EFPY."

3-5 Amen dment No. 29, 1a4, 167,1 76,20 8,234 ,281

~

)>

3 Figure 3.1-1 CD a.

Reactor Coolant System Heatup and Criticality Limitati'ons 3

[Applicable through 50.2 EFPV]

CD 2400 HU Limits Adjusted Reference 280, 2396 f r J 320. 2398 Criticality Limits 2200 Temp Press Temperature -

2000 60 101 575 575 '-

Beltline 1/4T Circ. Weld 234.5 "F 270, 2057 J '

J 310. 2057 Temp Press 263 263 0

1111 ,__

-I\J -.

C) u, 1800 120 150 180 59-9 662 ....

Axial Weld 184.7 ClF Beltfine 3/4T Circ. Weld 178.5 "F l

J 300, 1m 1 2.72 280 1235 1360 ,__

co C.

200 778:

&90 Axial Weld 126.8 "F 255, 1657 ) 290 300 154S 1777

.'190, Cl) 1600 I- ,__

w lo. 220 1079. Closure Head 60 "F 310 2057 I

01 ~ 1548 u, 240 1360 Nozzles 60 "F

)

l 320 2398 II) u, 1400 Cl) 255 1657 ~

- 2i°, 136/ 28~, 1360 270 2057 11.

U) 1200 280 2398 272, 1235 0

a:::

"O 1000 220.1J79 J 1 I (263, 11 11 Cl) 1a ,200, 890/ (

(J "O

800 180,778 C

600 ' - - 60, 575 Cl 101,575

... 120,599 150,662

,, ,I 400 ,,

Criticality Limits I 200 263,0 See Following Notes 0 ....

0 50 100 150 200 250 300 350 400 450 500 Indicated RCS Inlet Temperature, °F

Notes to Figure 3.1-1 1 - Temperatures:

All Temperatures are the indicated values in the operating RC pump(s) Cold Leg.

Except:

When the DHRS is operating without any RC Pumps operat ing, then the indicated DHRS return temperature to the Reactor Vessel shall be used.

2- Heatup:

50F/Hr or 15F/18 Min. Steps 3 - RC Pump Combinations for Heatups:

T s 100 No RC Pumps Operating 100 <Ts 199 Any 1 or 2 Pump Combination (2/0, 0/2, 1/1) 200 < T s 349 Any Pump Combination except 2/2 T ~ 350 Any Pump Combination 4 - Criticality Limits:

When the core is critical, allowable combinations of pressu re and temperature shall be to the right of the criticality limit curve.

a~

~

3-5b Amen dment No. 281

l>

Figure 3.1-2 3

(D Reactor Coolant System Cooldown Limitations a.

3 [Applicable through 50.2 EFPY]

(D 2400 I t' 265, 2400 CD Limits Adjusted Reference I Ji 2200

. 2000 Temp Press (F) 70 (psig) 519 Temperature Beltline 1L4T Circ. Weld 234.5 °F I . 260 2220 101 519 Axial Weld 184.7 °F

-I\J C) en 1800 130 160 574 698 Beltline 3L4T Circ. Weld 178.5 °F I l 250, 1909 e::,.. 1600 Q.

0:, 190 921 220 1276 Axial Weld 126.8 °F Closure Head 60 °F /1240, 1619 240 1619 Nozzles 60 °F en

(,J I

01 a.-

~

u, 1200 1400 250 1909 260 2220 265 2400 t 220, 1276 V

(')

0

~ 1000

-~-

Q)

/1,90,921 ca 800 "C

- C 600 70,519 = .1

_,, ~o, 69!

130,574 400 101,519 200 See Following Notes o

o 50 100 150 200 250 300 350 400 450 500

Indicated RCS Inlet Temperature, °F

Notes to Figure 3.1-2 1 - Temperatures:

All Temperatures are the indicated values in the operating RC pump(s) Cold Leg.

C Except:

When the DHRS is operating without any RC Pumps operating, then the indicated DHRS return temperature to the Reactor Vessel shall be used.

2 - Cooldown:

T > 255F 1OOF/Hr or 15F/9 Min. Steps T s 255F 30F/Hr or 1SF/30 Min. Steps 3 - RC Pump Combinations for Cooldowns:

T s 100 No RC Pumps Operating 100 <Ts 199 Any 1 or 2 Pump Combination ( 2/0, 0/2, 1/1) 200< Ts 349 Any Pump Combination except 2/2 T ~ 350 Any Pump Combination 3-Sd Amendment No. 281

~

3 CD Figure 3.1-3 3

a. Reactor Coolant lnservice Leak Hydrostatic Test CD

-z

, * [Applicable through 50.2 EFPY]

2400 I I ,I, I I Adjusted Reference

[ 260, 24~0 ISLHUmits**

2200 Temper ature Temp Press Beltline 1L4T J 255, 2244 {F) (psig)

Circ. Weld 234.5 °F 2000 J 60 671 Axial Weld 184.7 °F 65 724 C) Beltline 3L4T enC. 1soo Ctrc. Weld 178.5 °F '240., 1848 90 120 724 759 I\)

CX>

...... e"' 1aoo

J Axial Wefd 126.8 "F erasure Head 60 °F I 150 180 894 1067 wI en u,

a., 1400 D.

Nozzles 60 OF I

1210. 1335 210 240 255 1335 1848 2244 01 CD U) 1200 260 2400 0

c::

'C 1000 a., ~

/. 180, 1067 1u ~

-(.)

'C 800 65,724 1~ --

Jiir 150,894

-C 600 60,671 90,724 120, 759 400 200 See Follow ing Notes 0

0 50 100 150 200 250 300 350 400 450 500 Indica ted RCS Inlet Temp eratur e, °F

Notes to Figure 3.1-3 1 - Temperatures:

All Tempe ratures are the indicated values in the operating RC pump(s) Cold Leg.

Except:

When the DHRS is operating without any RC Pumps operating, then the indicated DHRS return temperature to the Reactor Vessel shall be used.

2 - Heatup:

SOF/Hr or 1SF/18 Min. Steps 3 - Cooldown:

T > 255F 1OOF/Hr or 1SF/9 Min. Steps T s 255F 30F/Hr or 1SF/30 Min. Steps 4 - RC Pump Combinations for Heatups / Cooldowns:

T s 100 No RC Pumps Operating 100 <Ts 199 Any 1 or 2 Pump Combination (2/0, 0/2, 1/1) 200 <Ts 349 Any Pump Combination except 2/2 T ~ 350 Any Pump Combination 3-5f Amend ment No. 281

3.1.3 MINIMUM CONDITIONS FOR CRITICALITY AQplicability Applies to reactor coolant system conditions required prior to criticality.

Objective

a. To limit the magnitude of any power excµrsions resulting from reactivity insertion due to moderator pressure and moderator temperature coefficients.

b. To assure that the reactor coolant system will not go solid in the event of a rod withdrawal or startup accident.

c. To assure sufficient pressurizer heater capacity to maintain natural circulation conditions during a loss of offsite powe r.

Specification 3.1.3.1 The reactor coolant temperature shall be above 525°F except for portions of low power physics testing when the requirements of Specification 3.1.9 shall apply.

3.1.3.2 Reactor coolant temperature shall be above OTT+ 10°F.

~

,f:JJIJ 3.1.3.3 When the reactor coolant temperature is below the minimum temperature specified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Specification 3.1.9 shall apply, the reactor shall be subcritical by an amount equal to or greater than the calculated reactivity insertion due to depressurization.

3.1.3.4 Pressurizer 3.1.3.4.1 The reactor shall be maintained subcritical by at least one percent delta k/k until a steam bubble is formed and an indicated water level between 80 and 385 inches is established in the pressurizer.

(a) With the pressurizer level outside the required band, be in at least HOT SHUTDOWN with the reactor trip .

breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in COLD SHUTDOW N

within an additional 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.1.3.4.2 A minimum of 107 kw of pressurizer heaters, from each of two pressurizer heater groups shall be OPERABLE. Each OPERABLE 107 kw of pressurizer heaters shall be capab le of receiving powe r from a 480 volt ES bus via the established 1 manual transfer scheme.

3-6

Amen dmen t No. 78, 157 278

( a) With the pressu rizer inoperai:>le due* to one Cl) inoper *ble

emerge ncy power supply to the pressu rizer heater s either .J rest~r e the inoper aole emergency power supply witn.in 7 days or be in at least HOT STANDBY within the next 6 nours and in HOT SHUTDOWN within the follow ing .12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(b) Wii:h the pressu rizer inoper able due to two (2) inoper able emerge ncy power suppli es to the pressu rizer heater s eicner resto.r e the inoper able emergency power suppli es within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or .>e *in at least HOT *s.rAi-lDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the follow ing 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

J. l.3. 5 Safety rod groups shall be fully witndr awn .prior to lny ocher reduct ion in shutdown mar 6 in by debora tion or regula ting rod withdr awal during the approa ch to critic ality with the follQV -

in6 except ions: .

a. Inoper able rod per 3~5.2. 2.

b. Phvsic s restin g per 3.1.9.

c* Shutdown margin inay not be reduce d below 1% A k/k per 3. 5. 2.l._.

d. Exerci sing rods per 4.1.2.

Follow ing safety rod withdr awal, the regula ting Tods shall be positio ned within their positi on limits as define d by Speci fi-cation 3.5.2. 5 prior to decora tion.

3-oa Amendment No. 78

CONTROLI 8) COPY

. ~'1"~~-,~~~.'.::> ;,_~ ~

t, Jl Bases At the beginning of life of the initial fuel cycle, the moderator temperature coeff icient is expected to be slight ly positive at operating temperatures with the opera ting configurafion of control rods~

Calculations show that above 525°F the positive moderator coeffi cient is accep table .

Since the moderator temperature coefficient at lowe r tempe rature s will be less negat ive or more positi ve than at operating temperature, startu p and opera tion of the react or when reactor coolant temperature is less than 525°F is prohib ited except wher e necessary for low power physics tests. ,, * * * .

The potential reactivity insertion d1:1e to the mode rator pressu re coefficient that could-result from depressurizing the coola nt from 2100 psia to satyra tion press ure of 900 psia is approximately 0.1 percent delta k/k.

During physi cs tests, special opera ting precautions will be takeh . In addition, the strong negative Doppler coeffi cient and the small integr ated delta k/k would limit the magn itude of a power excursion result ing from a reduction of mode rator densi ty.

The requirement that the reactor is not to be made critical below DTT+10°F provi des increased assurances that the prope r relationship between primary coola nt pressure and temperatures will be maintained relati ve to the NOTT of the prima ry coolan t syste m. Heatup to this temperature wil_l be accomplished by operating the reacto r coola nt pumps.

If the shutd own margin required by Specification 3.5.2 is maint ained , there is no possibility of an accidental criticality as a result of a decrease of coolan t pr~ss ure.

The availability of at least 107 kwin pressurizer heate rcapa bility is sufficient to maintain prima ry system pressure assum ing normal system heat losses.

Emer genc y power to heate r groups 8 or 9, supplied via a manu al transfer scheme, assur es redun dant capability upon loss of offsite power.

The requirements that the safety rod groups be fully withdr awn before criticality ensures shutdown capability durin g startup. This requir emen t does not prohibit rod withd rawa l when the reactor will remain more than .1 % dk/k shutd own with the rod(s ) withdrawn (e.g. , rod latch verification).

The requirements for regulating rods being within their rod posit ion limtts ensu res that the shutdown margin and ejected rod criteria at hot zero power are not violated.

3-7 Amen dmen t No. 78, 1 57, ECR TM 04-00911

3.1.4 REACTOR COOL ANT SYST EM (RCS) ACTIVITY 3.1.4.1

LTh'HTING CONDITION FOR OPER ATiO N RCS DOSE EQUI VALE NT I-131 and DOSE EQUIVALE NT Xe-133 speeific activity shall be limited to:

a. Less than or equal to Q.35 microcuries/gram POSE EQUI VALE NT 1-131. and

b. Less than or equal to 797 microcuries/gram DbSE EQUI VALE NT Xe-I 33.

3.1.4.2 APPLICABILITY: At all times except REFU ELIN G'Sli UtDO WN and COLD SHUT DOWN.

3.1.4.3 .. *. ACTION:

.' ::~**

MODES: At all times excep t REFUELING SHUT DOW N and COLD SHUT DOW N

a. l With DOSE EQUIVALENT l-131 not within limit, perfo nn Jhe sampling and i:tnalysis requirements of Table 4.J .3 until the RCS DOSE EQUI VALE NT l- 131 is restored to within limit, AND a.2 Verify tl~at DOSE EQUI VALE NT 1-131 is Jess than or equal AND . . to 60 microcuries/gram, a.3 Restore DOSE EQUIVALENT 1-131 to within Limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

a.4 If the requirements of a. I, a.2 or a.3 canM t be met, be .in at least HOT SHUT DOW N within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT DOW N within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months b.1 With DOSE EQUJVALEN T Xe,-133 not within limit, restore

. DOSE EQUIVALENT Xe-,133 to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , b.2 If the tequirements .of b.1 cannot be met, be in at least HOT SHUT DOW N wHhin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT DOW N w'ithin 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

LCO The iodin e specific 'activity in the reactor coolant is limite d to 0.35 µCi/gm DOSE EQU N ALEN T 1-131, and the noble gas specific activity in the reactor coolant is limite d to 797 µCi/gm DOSE EQUIVALENT Xe-133.

The lilliits on specif ic .activity ¢nsure that offsite and contro l room dose~ will. meet the appropriate 10CFRI00.11 (Ref. 1) and 10CFR50 Appendix A GDC 19(Re f. 5) acceptance criteria.

The SLB ~nd SGTR accident analyses (Refs. 3 and 4) show that the calcu lated doses are within acceptable limits. Violation of the l.CO may result in reactor coola nt rad_ioactivity levels that could, in the event of a SLB or SOTR , lead to doses that exceed the IOCFRl00.

11 (Ref. 1) a;nd lOCF R50 Appendix A GDC I9 (Ref.

5) accep tance criteria.

3-8 Amen dmen t No. 108. 117,2 04,27 2

Bases (continued)

APPLICA131LITY lq all MODES other than REFUE LING SHUTD OWN and COLD SHUTD OWN, oper~tion within the LCO limits.for DOSE EQUIV Al.ENT 1-131 and DOSE EQUIV ALENT Xe-133 is necessary to limit the potential consequences of a SLB or SGTR to within the I OCFR I 00.11 accepta nce criteria (Ref. 1) and I OCFR50 Appendix A GDC 19 acceptance criteria (Ref. 5).

Jn the REFUELING SHUTD OWN and COLD SHUTD OWN MODES , the steam generators are transitioning to decay heat remova l and primary to secondary leakage is minima

l. Therefore, the monitoring of RCS specific activity is not required.

ACTIONS LCO 3.1.4.3. a.l, a.2, and a.3 With the DOSE EQUIV ALENT I-1 :31 greater than the .LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to _demonstrate that the specific activity is~ 60.0 gCi/gm. The Comple tion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample. Sampling is continued every 4 hour~

to provide a trend.

The. DOSE EQUIV ALENT 1-131 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of48 hours is accepta ble since it is expected that, if there were an .iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

LCO 3.1.4.3. b.l With the DOSE EQUIV ALENT Xe-133 greater than the LCO limit, DOSE EQUIV ALENT Xe-133 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The allowed Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expecte d that, if there were a noble gas spike, the normal cool~t noble gas concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

LCO 3.J.:4.3.a.4 and 3.1.4.3.b.2 If the Required Actions of 3.1.4.3. a and 3.J .4.3.b are not met, or if the DOSE EQUIV ALENT 1-131 is >

60.0 µCi/gm, the reactor must be brought to HOT SHUTD OWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTD OWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , The allowed Completion Times are reasonable, based on operatin g experience, to reach the required plant conditio ns from full power conditions in an orderly manner and without challenging

. plant systems.

SURVE ILLAN CE REQUI REMEN TS Table 4.1-3, Item La

Table 4.1-3, Item l .a.i requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant at least once every 7 days. This measure ment is the sum of the degasse d gamma activities and the gaseous gamma activities in the sample taken.

This Surveillance provide s an indication of any increase in the noble gas specific activity.

Trendin g the results of this Surveil lance allows proper remedial action to be taken before reaching the LCO limit under norma:i operatin g conditions. The 7-day Frequen cy conside rs the low probability of a gross fuel failure during this time.

The 7-day Frequen cy is adequat e to trend changes in the xenon activity level.

The Frequency, between 2 a

and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months af!:er power change > 15% RTP within a I hour period, is establis hed because the xenon levels peak during this time followin g iodine spike initiation; samples at other times would provide inaccurate resµlts.

3-9 Amend ment No. 108, 117, 204, 272

Bases (Contiriued)

If a specific rioble gas nuclide listed in the definition of DOSE EQUiV ALENT xe~l3 3 is not detected, it should be assumed to be present atthe 111inimum detec table activity.

Table 4.1-J, Item l .b The TableA.1-3, Item l .b surveillance for isoto*pic analy sis for DOS E EQUIVALENT 1-13 I concentration':is perfonned to ensure iodine *specific activity remains within the LCO 1imit during normal operation and following fast- power. changes wh~n iodin e spiking is more apt to occur. The 14-day Frequency is adequate to trend .changes in the iodine activity level., considering noble gas activity is monitored every 7 ,clays. The Frequency. ~twe en2 and 6 'hours after a pi;,wer change > 15 % Rll within a l. hour ~rio d, is esta.bHsned because the iodine levels peak during this time iolloi.ving iodine spike initiation; samples*.at oth~r times Would provide inacc urate results.

REFERENCES I. IO CFR 100.1 I.

2. Standard Review Plan (SRP) Section 15. l.5 Appe ndix A (SLB) and Section 15.6.3 (SGTR).

3. FSAR. Section 14.1:2.9.

4. FSAR, Section 14.l.4._IO.

5. 10 CPR 50 Appendix A, General Design Criteria 19

\

3-9a Ame ndme nt No. 272

CONTROUED OOPY

3. 1. 5 CHEMISTRY A pp lic ab ili ty Applies to ac ce pt ab le operation of th e re concentrations of im pu ac to r. rit ie s fo r continuous Objective To pr ot ec t th e re ac to r coolant system from th e ef fe ct s of im pu rit ie s.

Sp ec ifi ca tio n

3. 1. 5. 1 If the concentration of exceeds O.l ppm durin oxygen in .the primary g po coolant sh al l be in iti at ed with wer op er at io n, co rr ec tiv e ac tio n le ve ls to ~ 0.1 ppm. in ei gh t hours to re tu rn oxygen

3. 1. 5. 2 If th e concentratio exceeds 0.15 ppm dunrinof g ch lo rid e in th e primar y coolant po sh al l be in iti at ed with wer op er at io n, co rr ec tiv e ac tio n le ve ls to ~ 0.15 ppm. in ei gh t hours to re tu rn ch lo rid e

3. 1. 5. 3 If th e concentratio exceeds 0.10 ppm fonlloof wi flu or id es in th e primar y coolant primary system involvin ng m od ifi ca tio ns or re pa g welding, co rr ec tiv e ir to th e be in iti at ed within ei gh ac tio to 5 0.10 ppm. t hours to re tu rn flu or n sh al l id e le ve ls

~(9/'-') 3. 1. 5. 4 r~;,,

If th e concentratio flu or id e given in 3.n1.lim its for oxygen, ch lo rid e or not re st or ed within 24 5. 1, 3. 1. 5. 2, and 3. 1. 5. 3 above are sh al l be placed in a hohours of de te ct io n, the re ac to r th er ea ft er . If the norm t shutdown co nd iti on w ith in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months re st or ed w ith in an ad di al operational lim its ar e not sh al l be placed in a co tio na l 24-hour pe rio d, the re ac to r hours th er ea ft er . ld shutdown co nd iti on w ithin 24

3. 1. 5. 5 If th e oxygen, ch lo primary co ol an t systerid m e, or flu or id e co nc en tra tio n of the be brought to the hot exceeds l.O ppm th e re ac to r sh al l shutdown procedure and shutdown cond1tion using normal the system to within noac tio n is to be taken to re tu rn normal op er at in g sp ec ifirmal operation sp ec ifi ca tio ns . If 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , th e re ac to r w ca tio ns have not been reached in shutdown co nd iti on . ill then be brought to a cold

Bases By maintaining the ch lo rid the re ac to r co ol an e, flu or id e, and oxygen re ac to r coolant system t w ith in the sp ec ifi ca tio ns , th e co nc en tra tio n in in te gr corrosion at ta ck (Refer is protected ag ai ns t po te nt ia l st reity ss of the ences i and 2}.

3-10 Amendment No. 157

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/j1'\1 The oxJgen concentratio expected to be below den in the reactor coolant system is no used when the re ac to r te ct ab le lim its si nc e dissolved hydrogrmally concentration not exce is cr it ic al . The requirement th at th en is assurance th at st re ss ed 0.1 ppm during power operation is ead oxygen corrosion cracks w ill de not occur (Referenced 3)

If the oxygen, ch .

can be taken to cplorrrid e, or flu or id e lim its deminera1izer, replacec t the condition {e .g ., ~re exceeded, measures switch to hydrogen concentratione the ion exchange re si n, or increathsee ~pare in the makeup tank)~ the Because of the tim from ch lo rid es , flueorde pendent nature of any adverse ef fe limits~ and because th es , or oxygen co nc en tra tio ns in exctcesssar is in g id e condition can be co to shutdown immediately

. rr ec te d, 1t is unneof the cessary The oxygen, ch 1or i de, order of magnitude below or f1 uor i de 1imi ts sp ec to m at er ia ls found in concentrations which ifi co ed are at 1ea st an uld re su fo r an extended period the re ac to r coolant system even if ltmaiin damage ei gh t hours t~ in it ia te of time (Reference 3) . Thus, the perio ntained to perform co rr ec tiv e co rr ec tiv e ~ction and the pe d of ac tio n to re st or e the conc~ntratio rio d of 24 hours lim its have been es ta bl n w co rr ec tiv e ac tio n allo is he d. Th e eight hour period to ith in the are co rr ec t and to w s tim e to as ce rta in th at the chem in it ia te ac tio n has not been loefca te the source of contamination. Ifica co l an al ys is re ac to r coolant system fe ct iv e at the end of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , th rr ec tiv e within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months th er ea w ill be br ought to th e hot shutd en th e the normal operational ft er an d co rr ec tiv e ac tio n w11l co ow n co ndition 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period th e re lim its ar e not re st or ed w ith in .a n ~t in ue . If within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months th er ea ac to r sh al l be placed in cold shutd ad di tio nal ft er . own condition The maximum lim it of concent.ration th at will l ppm for the oxygen, ch lo rid e, or

. values have been shown not be exceeded was se le ct ed becaflu us or id e prudent to re st ri ct op to be safe at 550°F (Reference 4) . It these e 1imi ts ar e reached. eration to hot shutdow is n co nd iti on s, if th ese

-'.'1 REFERENCES

{l} UFSAR, Section 9.2

- "Chemical Addition an d Sampling System 11 (2) UFSAR, Table 9.2-3

- "Reactor Coolant Qua lity" (3) Corrosion and Wear Handbook, D.J. DePau1, Editor

{4) St re ss Corrosion of Metals, Logan 3- ll Amendment No. 157

3.1.6 LEAKAGE Applicability Applies to reactor coolant leakage from the reactor coolant system and the makeup and purification system.

Objective To assure that any reactor coolant leakage does not compromise the safe operatio n of the facility.

Specification 3.1.6.1 If th*e total reactor coolant leakage rate exceeds 10 gpm, the reactor shall be placed in hot shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of detection.

3.1.6.2 If unidentified reactor coolant leakage (excluding normal evaporative losses) exceeds one .gpm or if any reactor coolant leakage is evaluated as unsafe, the reactor shall be placed in hot shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of detection.

3.1.6.3 If the primary-to-secondary leakage through any one (1) steam generator. exceed s

150 GPD, the reactor shall be placed in hot shutdown within S hours, and in cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . '

3.1.6.4 If any reactor coolant leakage exists thrc;>ugh a nonisolable fault in an RCS strength boundary (such as the reactor vessel, piping, val.ve body, etc., except the*stea m

generator tubes), tt:ie reactor shall be shutdown, and a cooldo,wn to the cold shutdown condition shal,lbe initiated within.24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of detection.'

3.1.6.5 If reactor shutdown is required by Specification 3.1.6.1, 3.1.6.2, 3.1.6.3, or 3.1.6.4, the rate of shutdown and .the conditions of shutdown shall be determined by the safety evaluation for each case.

3.1.6.6 *- Action td evaluate the safety implication ofreactoi' coolant leakage st,all be initiated within four hciurs of detection. The nature, as weil as the magnitude, of the leak shall be considered in this evaluation. The safety evaluation shall assure that the exposur~tof offsite personnel to radiation is within the dose rate limits of the ODCM.

3.1.6. 7 If reactor shutdown is required per Sp.ecification 3.1,6.1, 3.1.6.2, 3.1.6.3 or 3.1.6.4, the reactor shaUnot be restarted until the leak is repaired or until the problem is otherwise corrected.

3.1.6.8 When the reactor is critical and above 2 p~rcent power, two reactor coolant leak detection systems of different operating principles shail be in operation for the Reiictor Building with one of the* two sys~ems sensitive to radioactivity. The systems sensitive to radioactivity may be out-of-service for no "'!ore than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months p~ovide da sample is taken ofthe Reactor Building atmosphere every eight hpurs and analyze d

for radioactivity and two other means are available to detect leakage.

3-12 Amendment No. 47, ~ . 180, 24e, 281, 271 (12-22-78)

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3.1.5 .9 Loss o, react or coola nt through r,1ct or coola nt pump s11ls and system v1lv1 s to c:onnec:tfng syst11111 wtlfch v1nt wtlf~tl coola nt c:1n be retur ned to tht r11ct or coola nt to. tht gas vent he1d1r and from consf ~i"ld II react or coola nt lt1k1 g1 and shill not bf*systt

shall not bt cons. fdtra tfon* of, Sp1e fftc1 tfon1 l.1.5 .1, l.1.5 ~2, l.1.5subJ tct to the 3.LI .S~ 3.1.1 .5* or 3.1.S .7, 11c1 pt that suc*11 lout s .3,.3 .1.s. ,,

111t1g1 shall nat 1xc11d 30 gpm. If l*1k~ ge plus losse whtn 1ddtd to ttte react or shill be p11c1d fn HOT SHUTDOWN* wfthf n 24 s 1xc11 ~s 30 gpm, ho*urs .of dtttc tfon.

3.1.6 .10 Op1r1tf ng cond ftf O"- of POWER OPERATION, STARTUP and to tht Qp1r1tfor11l s~1tu1 of ttlt higll press ure fsoll tion HOT SHUTDOWN apply tht prf111f"y coola nt 1y1t1

and tht low pr1ss ure fnjec v1lv1 i betw tn tfon syst1M.

1. Du,.1ng 111 op,,.a tfng cond ftfon s fn thfs specf f-fc1 tfon, fsola tfon v1lv11 lfstt d fn T1bl1 3.1.5 .1 that art .loc1 111 Dl"'ISSurt tht prf1111ry coola nt system and the LPIS shall funct ion t1d bltw en fso11 tfon dtvfc ts 11ce pt 11 sptc ffftd fn 3~1.5 .10~6 . 11 prtssul'"t shill not 1xc11d the amount fndfc 1t1d fn Table 3.1.5 Yalvt leaka ge

.1.(a )

b. In the tvtnt th*t f nttgr fty of any hfgh P"H u" fsola valve s spec fffed f n Table 3.1.1 .1 cann ot bl dima nstri tfon check oper ation nay contf1111 provf dld that 1t leas t t'NO* valve tad, react a,.

high prtss ure lfnt* tlavfng a non-f unctf o,..1 valve ' " s in 11ch fn, t111 IIIDdl corre spond ing to the fsola tld cond ttfon fn and re*t n

. (b)

c. If Spec fffca tfon 3.1.5 .10.a or 3.1.1 .10.b cann ot bt 1111t, an order shutdown shall be acco11plfshtd by 1ctlf1¥f n9 HOT SHUTD ly hours and ca.a SHUTDOWN wf thf n an addtt tonal 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months OWN w1th fn 6 81111 Any 111k of radfo actfv e nutd , .tteth er fro* th* react syste

prf* ry boundary o,. not, can bl a strf ous problor coola nt respe ct to tn-p1 1nt ,.adfo actfv e conta afnat fon and requ. . wf tll o,., fn the case of react or cool ant, ft could d1,11 op 1nto f"d clean up

..,,.. se,.fo us prob lta.a nd, there fore, tht ffrst tndtc atfon a stf 11 leaka ge wf11 bt follow ed up 11 soon 11 pr1ct fc11. Ttle s of such systt ll has the capa bflf ty to *keu p consf dlrab ly mre unft 's 111keup than 30 p of react or coola nt l11k191 plus 101111.

Vattr fnvtn tory balan ces, mnft orfng 1qufp lllilt, ,.,dfo actfv I

borfc acfd crys tallf nt dtpo sfts, and physi cal fnspt ctfon e traci ng, s can df sclos t react o,. coola nt 111ts .

cal For tht pur-pose of thfs spec fffc1 tfon, fnt19,.f1;y to havt betn dtmn strat ld by -tfn g Sptcf ffc:at fonfs 4.2.7 cons fdtre d (b) Motor oper 1tN valve s shall bt placi d fn tht close I

d posf tfon 111d powe,. supp lfts dNnt rgf zed.

Am1ndlllnt No. >', W , 141

Bases <Continued>

Although some leak rate s on the order A,~} tole rab le from a dose point of vl_ew, of gallons per minute may be the order of drops per minute through it ts recognized that leaks in

,/

primary system could be indicative of any of the barriers of the stre ss corrosion cracking. If depres materials fail ure such as by other saf ety *measures are not taken pro surization, iso lati on, and/or could develop 1ntci mtich larger leaks, ~ptly, these small leaks rupture. Therefore, the nature and loc possibly int o~ gross pipe the magnitude of the leakage. must be ation of the leak, as well as evaluation. considered in the saf ety When rea cto r coolant leakage occurs to ulti ma tely conducted to the Reactor Buithe Reactor Building, 1t 1s rea cto r coolant is. safely contained, lding sump. Although the escape to the . . Reactor Building atmosp the gaseous components in it components beco~e a potential hazard here . There, the gaseous inspection tours within the Reactor Bui to pla nt personnel, during pubJic whenever the Reactor Bui lding, and to the general purged to the environment. lding atmosphere is per iod ica lly When rea cto r coolant leakage occurs to col lec ted in the Auxiliary Building sumthe Auxiliary Building, it is rea cto r coolant l~akage within.the Au p. The gases escaping from col lec ted in the Auxiliary and Fuel Han xiliary Building will be ven tila tio n system and discharged to dling Building exhaust Auxiliary and Fuel Handling Buil~ing the environment via the uni t's thi s leakage occurs within confined, ven t. Since the majority of within the Auxiliary Building, it inc sep ara tely ven tila ted cubicles pe.rsonne 1. urs very lit tle hazard to pla nt In' .r.egard to the surveillance spe cifi valves may be tes ted at a reduced pre cat ion 4.2 .7, the iso lati on Franklin Research Center Report titl edssu re~ ~ accordance with the I

Pressure Iso lati on Valves for TMI-1" "Primary Coolant System October 24, 198 0, Section 2.2 .2. (FRC Task 212> dated Hhen rea cto r coolant leak.age occurs to cooling water system, the leakage, both the nuclear services closed contained because the nuclear services ** gaseous and 11qu1d, 1s surge tank is a. closed tank tha t is maiclosed cooling water: system pre ssu re. The leakage would be detect ntained above atmospheric

closed cooling water system monttor ed by the nuclear services both of which alarm in the control room and by purge tan~ liqu id lev el, closed cooling water system's only pot . Since the nuc.lear ser vic es coolant ts in the sample coo lers , \t ential contact with rea cto r hazard. However, If reactor coolant ls co_nsidered not to be a occurred and the surge tan k's rel ief leakage to thi s receptor gases could be discharged to the environ valve discharged, rad ioa ctiv e aux ilia ry and fuel handling building ment via the uni t's vent.

Order dtd. 4/2 0/8 1 3-13a

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Amendment no 149

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When reactor coolant leakage occurs to the intermediate cooling cl.osed cooling water system, the leakage is indicated by both the intermediate cooling water monitor (RM-L9) and the interme diate cooling closed cooling water surge tank liquid level indicator, both of which alarm in the control room. Reactor coolant leakage to this receptor ultimately could result in radioactive gas leaking to the environment via the unit's auxiliary and fuel handling building vent by way of the a~mospheric vent on the surge tank.

When reactor coolant leakage occl!rs to either of the decay heat closed cooling water systems, the leakage is indicated by the affected system's radiation monitor (RM-L2 or RM-L3 for system A and B, respectively) and surg~ tank liquid level indicator, all four of which alarm in the control room. Reactor coolant leakage to this receptor ultimately could result in radioact ive gas leaking to the environment via the unit's auxiliary and fuel handling building vent by way of the atmosp heric vent on the surge tank of the affected system.

Assumi ng the existence of the maximum allowable activity in the reactor coolant, a reactor coolant leakage rate of less than one gpm unidentified leakage within'the reactor or auxiliary building or any of the closed cooling water systems indicated above, is a conserv ative limit on what is allowable before the dose rate limits of the ODCM would be exceeded.

When the reactor coolant leaks to the secondary sides of either steam generator, all the gaseou s components and a very small fraction of the ionic components are carried by the steam to the main condenser. The gaseous components exit the main condenser via the unit's vacuum pump which discharges to the condenser vent past the condenser off-gas monitor. The conden ser off-gas monitor will detect any radiation, above background, within the conden ser vent.

However, buildup of radioactive solids in the secondary side of a steam generator and the presenc e of radioactive ions. in the condensate can be tolerated to only a small degree.

Therefore, the appearance of activity in the condenser off-gas, or any other possible indications of primary to secqndary leakage such as water inventories, condensate deminer alizer activity, etc., shall be considered positive indication of primary to secondary leakage and steps shall be taken to determine the source and quantity of the leakage.

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'0J9 3-14 Amendm ent No . .:t-+, ~. 7+, +a9, +49, 4-00, 246

Bases (Continued)

If reactor coolant leakage is to the containment, it may be identified by one or more of the following methods:

a) The containment radiation monitor is a three channel monitor consisting of a particulate channel, an iodine channel, and a gaseous channel. All three channels read out in the Control Room and alarm to indicate an increase in containment activity.

The containment particulate channel is sensitive to the presence of Rb-88, a daughter product of Kr-88, in the containment air sample. Since this activity originates predominantly in the Reactor Coolant System, an increase in monitor readings could be indicative of increasing RCS leakage. The sensitivity of the particulate monitor is such that a leakrate of less than 1 gpm will be detected within one (1) hour under normal plant operating conditions.

b) The mass balance technique is a method of determining leakag e by stabilizing the Reactor Coolant System and observing the change in water inventory over_a given time period. Level decreases in the Makeup Tank may also serve as an early indication of*

abnormal leakage.

c) The Reactor Building Normal sump receives leakage from system s inside containment. Sump level readings are checked and recorded regular ly for rate of water accumulation. High accumulation rates alert the operat ors to increase their surveillance of possible leak sources. One half inch of level corresponds to a volume of approximately 4Y2 gallons.

d) Deleted.

The leakage detection capability provided by the above methods can be used to determine potential pressure boundary faults. Such leakage, while tolerable from a dose point of view, could be indicative of material degradation which if not dealt with promptly, could develop into larger leaks.

If 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is exceeded, manual samples will continue at a hour intervals and an IR will be written to determine what additional actions need to be taken based on the plant conditions at that time. The evaluation to determine additio nal actions will consider current RCS leakage and trends, availability of other leakag e monitoring instrumentation (i.e., mass balance, flow balance, RB sump instrum ents) and existing NUREG-1430 guid~nce to appropriately limit the timefra me.

This specification is concerned with leakage from the Reactor Coolant System (RCS) and Makeup and Purification System (MUPS

). The metho ds discussed above provide a means of detecting, as early as possible, leakage which could be the result of a fault in the reactor coolant system pressure boundary. The primary metho d used at TMl-1 for quantifying RCS and MUPS leakage is the mass balanc e

technique.

3-15 Amend ment No. 17, 141, EGR TM 04 00601, EGR TM 06 00206, AR 4048309 (6 18 76)

Bases (Continued)

The unidentified reactor coolant leakage limit of 1 gpm is established as a quantit y which can be accurately meas_ured while sufficiently low to ensure early detection of leakage

. Leakage of this magnitude can* be reasonaqly d~tected within a matter of hours, thus providing confidence that cracks *associated with such leakage will *not develop into a crltical size before can be faken... . mitigati ng actions Total reactor coolant leakage is limited by this $pacification to 10 gpm. This)im 1tation provides allowanc~-fbr aliinited qiTIOqnt Of leakage ffomknown sources whose presence wrll not interfere with the detection of unidentifi~d leakage. . . .

Except for primary to secondaryJeakag*e, the safety analyses do riot address operational leakage .. However, otller operational leakage.is related to the .safety gnalyses for LOCA; the amount of leakag~ ,.can* affect the J'>robabmty of suet, an event. .The safety analysi s for an** event resulting in steamdischarge to.the*atmospnere assumes primary to secondary l~ijkage from all steam .generators (SG's) 'oependirig on the specific acci_dent analyses. -The leakage

_ rate may increa$e (over that observed during normal operation) as a result of accident.:induc ed conditions;* .The TS*retjuiremeht to limitthe primary*to secondary leakage througt1 any one (1)

SG to*less than or equal to 150* gallons per day' is significantlt less than the conditio ns assumed in the safety an~lysis. *

The lirrlit of 150 gallons per day per SG, is based on the operc:1tional leakage perform ance criterion in NEI 97-06, :steam Generator Program Guideline~f(Ref. 1). The Steam Generator Program operational leakage performance criterion in NEI 97-06 states; "The RCS*operational primary to secondary leakage tt]to1,igh any one SG shall b~ limited to 150 gallons per day." The limit is based on operating experien*c:e with *sG tube degradation mechanisms that result in tube leakage .. The oj:>eration~I leakage rate criterion in conj~hction with* the implem entation qf the

Steam GeneratorProgram_is an effective measure for minirriizingthe frequ_e ncyof steam generator tube ruptures.

If reactor coolant leakage is to the auxiHary building, it may be igentified by one following methods: .

or more of the

a. The auxiliary and fuel handling building vent radioactive g13,s monitor is sensitiv e to very low activity levels and would show an increase in ~ctivity level shortly after a

reac,tor coolant leak developed within the auxiliary building.

b. Water inventories around the auxiliary building sump.

c. Periodic equipment fnspections.

d. In the _event-of gro.ss leakage, in excess of 4.53 gpm, the individual cubicle leak

deteGtors in the makeup and decay heat pump cubicles, will alarm in the control room to backup "a", "b", and c" above.

When the source and location of leakage has been identified, the situation can be evaluated to determine if operation can safely continue. This evaluation. will be performed by TMl-1 Plant Operations.

REFERENCES (1) NEI 97-06, "Steam Generator Program Guidelines."

3-15a Amendment No. 444, Order dtd. 4/20/-81, 246, 261, EGR TM 07 00719, 271

TABLE 3.1.6.1 PRESSURE ISOLATION CHE CK VALVES BETWEEN THE PRI MAR Y COO LAN T SYSTEM &

LPIS System Valve No. Max imum Low Pres sure Injection Allo wab le Leakage Train A CF- VSA s5.0 GPM DH-V22A S5.0 GPM Train B CF-VSB s5.0 GPM DH-V22B S5.0 GPM 3-15 b Ord er Dated 4/20/81 Ame ndm ent No. 441-, 286

3.1.7 MODERATOR TEMPERAT URE COEFFICIENT OF REACTIV ITY Ap pl ic ab ili ty Applies to maxim po si re ac tiv ity at fuum tiv e moderator temperatu ll power condition re co ef fic ie nt of s.

Objective To assure th at the mod lim its calcu1ated fo r sa erator temperature co ef fic fe operation of the re ac ie nt sta ys within the to r.

Sp ec ifi ca tio n

3. 1. 7. 1 The mode or temperatu at power le ve ls abovrat re co e 95% of ra te d poweefr.fic ie nt sh al l not be po sit iv e

3. 1. 7. 2 The mode or 0.9xlo-& de lta k/k/Fratat temperature co ef f,c ie nt sh al l be <+

power le ve ls ~ 95% of rat~d power:

Bases A ~o n- po sit iv e mode above 95% of ra te d po rator co ef fic ie nt (Refer temperatures ~i ll not we r is sp ec ifi ed such en th ce 1) at power le ve ls LOCA.analyses. Below 95% exceed the Final Acceptaat the maximum clad nce Cr ite ria sed on Cr ite ria wi ll not be ex of ra te d power the Final Acceptancba e co ef fic ie nt of +0.9 x 10ceed ed with a po sit iv e mode rat or tem pe rature analyses as reported in - 4 de lta k/k/F. A11 ot he r ac th e UFSAR have been rformed ci de nt moderator temperature co k/k/F. ef fic ie nt s including pe fo r a ran

+O. 9 x 10-* de lta ge of A no n- po sit iv e mode or ra te d power is also rat co ef fic ie nt at power le re ac to r co ol an t system re qu ire d to prevent ov er pr esvesu ls above 95% of riz at io n of e Sp ec ifi ca tio n 2. 3. 1, Ba1n th e event of a f~edwater 11ne break th (see sis C, Reactor Coolant Sy stem Pressure)~

The Final Acceptance wi ll not exceed 2200°rCr ite r1 a st at es th at post-LOCA clad temperatu (Refere~ce i. ) re RE FE----

*-~ RE.---NCES

.~

(1) UFSAR, Section 3. 2.

(2) UFSAR, Section 14 2. l.S .4 - "Moderator

- Tables 14.2-1, 14.2-1Temperature Coeffic1ent" 3, 14.2-14 3-16 Amendment N o .~ Y~Z. 157 FEB 1, 1990

3 .1 .8 Single Loop Re str ic tio ns Ap pl ic ab ili ty Applies to sin gl e loop operation of the re ac to r coolant system Sp ec ifi ca tio n

3. 1. 8. 1 Si ng le loop operatitin wh pr oh ib ite d. ile the re ac to r i~ cr iti ta l is Bases The re st ric tio n pr oh ib iti lif te d, provided th at : ng sin gl e loop operation with TMI-1 may op er at io n, (2) te st in g (1 ) analyses of TMI-1 support sin gl e loop be op er at io n, and (3 ) any on TMI-1 supports the an al ys is of sin gl e operation is in st al le d additional eQuipment necessary for sin gl e loop*.

at TMI-1. loop

/1 f{

¥ ..

3-17 Amendment No. 157

3.1.9 LOW POWER PHYSICS TESTING RESTRICTION S

Applicability Applies to Reactor Protection System requireme nts for low power physics ~esting.

Objective To assure an addjtional margin of safety during testing. . low power physics Specification

The following special limitations are placed on low power physics testing.

3.1.9.1 Reactor Protection System Requirements

a. BeJow 1720 psig Shutdown Bypass trip setting limits shall appl y in accordance with Table 2.3-1.

b. Above 1800 psig nuclear overpower trip shall be set at less than 5.0 percent. Othe r settings .shall be in accordance with Table 2.3-1.

3.1.9.2 Startup Hate Rod Withdrawal Hold (Reference 1)

Shall be operable At All Times.

3.1.9.3 Shutdown margin may not be reduced below 1%

delta k/k per 3.5.2.1.

Bases The above specification provides additional safet y margins during low power physics testing, as is also provided for startu p (Refere,nce 2.)

REFERENCES (1) UFSAR, Section 7.2.2.1.b - "Reactivity Rate Limit s"

(2) UFSAR, Section 14.1.2.2 - "Startup Accident" 3-18

CONTROWED COPY 3.1. l O CONTROL ROD OPERATION This page intentionally left blank

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(Page 3-l 8b deleted) 3-18a Amendment No. 211

CONTROU ED COPY 3.1.11 REACTOR INTERNALS VENT VALVES Applicabilitv Applies to React6r Internals Vent Valves Objective To verify that no reactor interna ls vent valve is stuck in the open position and that each valve continues to exhibit freedom of movement.

Specifications 3.1.11:1 The structu ral integri ty and operab ility of the reactor interna ls vent valves shall be maintained at a level consis tent with the acceptance criteri a in Specification 4.16 .

. !t"

~)

3-18c Amendment No.,A't,167 (8-16-78) .

3.1.12 Pressurizer Pow er Operated .Relief Valve (POR V), Block Valve, and Low Temperature Overpressure Protection (LTOP)

Applicability Applies to the settings, and conditions for isola tion of the PORV.

Objective To pr.e~entthe possibility of inadvertently over pressurizing or depressurizing the Hea ctor Coolant System.

Specification 3.1 *; 12.1 LTOP Protection If the reactor vessel hemd is insJalled and indic ated RCS temperature is s 313°F, High Pressure Injection Pump breakers shall riot be -racked in unless:

a. (v1U-V16A/B/C/D are closed with their brea kers open, and MU~V217 is closed, and *

b. Pressurizer level is maintained s 100 inches.

If pressurizer level is >: .100 inches, restore level to s 100 inch es within. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

3.1.12.2 The PORV settings shall be as follow s:

a. Low Tem pera ture Overpressure Protection Setpoint

1. When indicatect RCS temperature is s 31~° F, the LTOP system shall be ope~able as d_efined in Spec ification 3.1.12.1 and

2. The PORV wilfh ave a maximum lift setpOint of 5$2 psig.

With the PORV setpoint above the rnax:imum value, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

1. restore the setp9int.below the maximum valu e, or

2. verify pressurjzer level is s 100 inches indic ated and satis fy the requirements of Technical Specification 3.1.1 be taken out of service. 2.3 allowing th,e PORV to

b. Unless the Low Temperature Overpre!?sure Protection Setpoint is in effect, the a

POR V lift setpoint will fie minimum of 2425 psig.

With the POR V setpoint below the minimum value, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either: .

1. restore the setpoint above the minimum value

, or

2. close the associated block valve, or

3. close the PORV, and remove power from POR V

4. otherwise, be in at least HOT STANDBY withi n the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the hours. following 30 3-18d Ame ndm ent No. 56, 78, 149, 167, *186, 234, 281

3.1.12.3 When* the, indicated RCS temperature is below 313° F the PORV shall not be taken out_ of service, nor .shall it be isolated from the system unless one of the following is in effect:* *

a. High Pressure.Injection Pump breakers are rack

b. ed out.

MU.:Vt6A/B/C/D are closed with their .bre_akers open MLJN217 is'-cfosed. * , and C. Head of the_ Reactor Vessel is removed.

3.1. 12.4 The PORV Blo'ck Valve shall b:~ QRE;RABlE durin g HOT STANDBY, STARTUP, *and POWER OPERATION: .

a. With the PORV Block*Valve inoperable, within 1 e~he~

hour

1. restore the PORV Block Valve to OPERABLE statu s or

2. close thePORV (verify closed) and remove pow the.PORV . er from

3. oth~f"Jllise, be in at least HOT STANDBY within the next

. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the follo wing .30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With. th~- PORV block V?IVe inoperc1ble;*restore the inoperable vl:ilve 'to OPERABLE st~tus,prior to start up. from

. t!le next COLD SHUTDOVvN'unless the COlD:SH UTDOWN occurs within go*-Effective Full Po'We(Day~ (EFPP) t:>f the end_ of the fuel *cyc:ie .. If a COLD SHUTDOWN occu:rsw ithin this 90 day pericl~, ::restore ttie *incip~rabl.~.yalye to OPE RABLE status prior to startup for the nextfuel cycle.

Bases lfthe PORV is* removed from service whi'le the.R CS is below 313~F,: .

sufficient measures' areJncorporated to prevenf severe. overpressurization by either eliminating the high :pre_ssure sources or flowpaths or a~suring that the RCS is open to atmosphere:

The PORV setpoints are specified with tolerance s assum~d in the bases for "'

Technipal Specificat'ion 3. 1.2. Above 313°F, the PORV setpoint'has been chosen to limit the potential-for ihadve.rtent disch arge or cycllng 9f th.ePORV.

Other *acti6ns4cfra~ rem.oving the;,powerfo the PORV has the sa'rne effect as raising the .setpolnt which also satisfies*this requ irement. There :js no upper limit ori this s~tpointa:s the'Pressurizer Safety Valv es (T.S. 3.1.1;;,3) provide the required overpressure reli_ef.

Below 313°F; the 'PORV setpoint is reduced to provide tllEl-

required low temperature overpr~ssure reliefvv~

en hig!:1 pressure sources anc::f flowpath~ are *in s_E;)ivice. There is no lower limit on the pressure act~.iatioh specified as lower setpointsalso prov fJ"~\ ide this same protection.

3-18e Amendment No. 78, 149, 167, 186, 234, 281

In both cases, the ~etting is specified to refle ct .the nqniiqal value whic~ allows for normal varia tions in the temperature setpoint while m~intairiing the tolerances assu med in the bases for T.S. 3.1.2. Either* pressure actuation setpoint is acceptable above 313°F.

With RC.SJemperatures less than 313" F and the makeup pum ps running, the higfi pre$*sure_:injeiction valves*are close~ and press~rrzer level is m';:1intalneq less thah 100 inches*to allow time severe:overpressurization ln tfie event of any for action tb prev ent srn'gle failure. .

The POR Vblo ck valye is req1,Jired.to b~ QPE RAB LEd uririg the HOT STANDBY, STARTUP', arid POWER OPERATION in ord,e r to provide isolcition of the PORV discharge line to positively control potential RCS depressurizatior\.

For prot~ctiori from severe overpressurization during HPI testing, refer to Section 4.5.2.1.c.

3-18f Ame ndm ent No. 186, 234, 281

3.1.13 REACTOR COO LAN T SYSTEM VENTS Applicability Provides the limiting conditions for ope ration of the Reactor Coolant System Vents. These limitfr1g conditions for ope ration (LCO) ~fre applicabie only when Reactor is critical. * *

Objective To ensure that suff icie nt vent flow path s are operable during the plant operating modes mentioned above.

Specification 3.1.13.1 At least one reactor coolant system vent path consisting of at least two power operated valves in series, powered from em(;lrgency buses shall be OPERAB I-E and closed at each of the following locations:

a. Reactor ve~sel head (RC-V42 & RC-V43

)

b Pre ssu rize r steam space (RG-V28 & RC-V44}

c. Reactor coolant system high point (eith er RC-V40A and 41 A) or (RC-408 and 41 B)

Action 3.1.13:2 a. With .one of the above reactor coolant system vent paths inoperable, the inoperablevent path shall be maintained closed' with 'power *removed

.from the valve actuators in the *inoperable veil t path ..

The-. inopernble ven tpat h shall be restored to OPERAB LE status within 30 day$; .or the plant sh ail *be in QT SHU TDO WN \vith in ari additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months arid in COLD SHU TDOWN.within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . . -

b. With two or more of the above rea~tor coolant system vent paths inoperable, maintain the iii~p erable vent path closed, with power removed from the valve actu-atqrs in the inoperable vent pc1,ths, ~nd restore at least two of the vent paths to OPERAB LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN with in an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN With in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3-18g Ame ndm ent No. 97, 186 278

CONTROLLED

,;,)

.....5*.:;...;,,.,_

Bases The safety function enhanced by this venting capability For events beyond the present design basis, this venting is core cooling.

subs tanti ally increase the olants abil itv to deal with larae capa bilit y will noncondensible gas which could interfere-with natural circu ouan tit;P. ~ of core cooling). latio n (i!e .,

The reactor vessel head vent (RC-V42 &RC-V43 in serie s) capa bilit y of venting noncondensible gases provides the from the vessel head as well as the Reactor Coolant hot legs {to majo rity of the reac tor top of the outl et rtozzles) and cold legs (through vessel the eleva tion of the paths, to the elevation of thj top of the inle t nozzles). inter nals leakage routed to containment atmosphere. This vent is Venting for the pressurizer steam space (RC-V28 and RC-V been provided to assure that the pressurizer is available44 in series) has Coolant System pressure and volume control. This vent is for Reactor Reactor Coolant Drain Tank. routed to the Additional venting capability has been provided for the leg high points (RC-V40A, 8, RC-41A, B), which' normally Reactor Coolant hot through* the Reactor vessel head vent or pressurizer steam cannot be vented vents relie ve to containment atmosphere through a rupture -space vent. These pressure) *. disk (set at low The above vent systems are seismicall qual ified in accordance with the May y23,desig ned and environmentally 1980 Conunission Order and Memorandum per NUREG-0737, Item II.8. 1. The high within the scope of 10 CFR 50.49, since the vents poin are t vents do not fall not or following any design basis event (Reference 1). The relie d upon during valves (2 in serie s in each flow path) which are powered power operated buses fail closed on loss of power. All vent valves for from emergency head vent, pressurizer vent and loop B high point vent arethe reactor vessel class IE "B" bus. The vent valves for the loop A high poinpowered from the powered from the class IE "A" bus. The power operated valvet vent are in the Control Room. The individual vent path lines are s are controlled inadvertent valve opening will not cons sized so that an 50.4 6(c) (l). These design features provtitut e a LOCA as defined in 10 CFR ide a high degree of assurance that these vent paths will be available when needed, and that inadvertent operation or failu res will not sign ifica ntly hamp er the safe operation of the plant (Reference 2).

REFERENCES (1) UFSAR, Section 4.2.3 .9 - "Reactor Coolant System Vent ing" (2) UFSAR, Section 7.3.2. 2.c (16) - "Reactor Coolant Syste m Venting" 3-lSh Amendment No. 17, t 49, t(,1, 186

3.2 MAKEUP AND PURIFICATION AND CHEMICAL ADDITION SYSTEMS DELETED 3-19 Amendment No. 43, 60, 98, 1e7, 196

THIS PAGE LEFT BLANK INTENTIONALLY 3-20 Amendment No. 60, 162, 167, 168, 196

3.3 EME RGE NCY CORE COOLING, REACTO R BUILDING EME RGE NCY COOLING AND REA CTO R BUILDING SPR AY SYSTEMS Applicability Applies to the operating status of the emergen cy core cooling, reac tor building emergency cooling, and reac tor building spray systems.

Objective To defin e the conditions necessary to assu re immediate availability of the eme rgen cy cooling, reac tor building eme rgen cy cooling core and reactor building spra y syste ms ..

Specification 3.3.1 The reac tor shall not be mad e critical unle ss the following conditions are met:

3.3.1.1 Injection Syst ems

a. The bora ted wate r storage tank (BW SD shall contain a minimum of 350,000 gallons of wate r having a minimum concentration of 2,50 0 ppm boro n at a temp eratu re not less than 40°F. If the boron concentration or wate r temp eratu re is not within limits, resto re the BWS T to OPE RAB LE within 8 hrs. If the BWS T volu me is not within limits, restore the BWS T to OPERABLE withi n one hour. Specification 3.0.1 applies.

NOT ES:

1. The BWS T piping may be unisolated from seis mic Class II Clea nup path piping for a total duration of not more than 720 hour0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months s prior to the scheduled start of the Fall 2015 Refueling Outa ge and for a total dura tion of not more than 1440 hour0.0167 days 0.4 hours 0.00238 weeks 5.4792e-4 months s during the following Fuel Cycle 21 operation unde r administrative and design controls for filtration and/ or demineralization of the tank contents.

2. The BWS T piping may be unisolated from seis mic Class II Recirculation path piping for not more than 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months per wee k to perform wee kly (and after each mak eup) BWS T boron concentration surv eilla nce testi ng unde r administrative and design controls until the end of Fuel Cycl e 21 operation.

b. Two Mak eup and Purification (MU)/Hig I h Pressure Injection (HPI) pum ps are OPE RAB LE in the engineered safeguards mod e powered from inde pend ent esse ntial buses. Specification 3.0.1 applies.

c. Two deca y heat removal pum ps are OPERAB LE. Specification 3.0.1 applies.

d. Two deca y heat removal coolers and their cooling wate r supplies are OPE RAB LE.

(See Specification 3.3.1.4) Specification 3.0.1 applies.

e. Two BWS T level instrument channels are OPE RAB LE.

f. The two reactor building sum p isolation valve s (DH- V-6A /B) shall be remote-man ually OPE RAB LE. Specification 3.0.1 applies.

3-21 Ame ndm ent No. 24,. 93,4 n,-2 W,~ .~.2 2-7.

Corrected by letter dtd July 8, 1999 273 ,289

3.3 EME RGE NCY CORE COOLING, REACTO R BUILDING EME RGE NCY COOLING AND REA CTO R BUILDING SPRAY SYSTEM S (Contd.}

g. MU Tank (MUT) pressure and level shall be maintained within the Unrestricted Operating Region of Figure 3.3-1.

1) With MUT conditions outside of the Unrestric ted Operating Region of Figure 3.3-1, restore MUT pressure and leve l to within the Unrestricted Ope ratin g Region within 72 hrs. Specificat ion 3.0.1 applies.

2) Ope ratio n with MUT conditions within the Proh ibited Region of Figure 3.3"1 is prohibited. Specification 3.0.1 applies.

3.3.1.2 Core Flooding Syst em

a. Two core flooding tanks (CFTs) each cont aining 940 +/- 30 ft 3 of borated wate r at 600 +/- 25 psig shall be available. Specificat ion 3.0.1 applies .

3-21a Ame ndm ent No. ~.8 8-,4 78, ~,2 44. ~.~

. Corr ecte d by letter dtd July 8, 1999

~ . 289

3.3 EMERGENCY CO RE COOLING, REA CTOR BUILDING EMERGENCY CO AND REACTOR BUILDING SPRAY OLING SYSTEMS (Contd.)

b. CFT boron concentration shall not be less than 2,270 ppm boron.

Specification 3.3.2.1 applies.

c. The electrically operated discharg e valves from the CFT will be assure administrative control and position indi d open by cation lamps on the engineered safe status panel. Respective breakers for guards these valves shall be open and conspicuously marked. A one hour time clock is provided to open the valv remove power to the valve. Specific e and ation 3.0.1 applies.

d. DELETED

e. CFT vent valves CF-V-3A and CF-V-3B shall be closed and the breake vent valve mot or operators shall be rs to the CFT tagged open, except when adjusting tank level and /or pressure. Specific core flood ation 3.0.1 applies.

3.3.1.3 Reactor Building Spray Sys tem and Reactor Building Emergenc y Cooling System The following components must be OPERABLE:

a. Two reactor building spray pumps and their associated spray nozzles hea two reactor building emergency coo ders and ling fans and associated cooling unit each train). Specification 3.0.1 app s (one in lies.

b. The Reactor Building emergency sump pH control system shall be mai

~ 18,815 lbs and s 28,840 lbs ntained with of trisodium phosphate dodecahydrate

. Specification 3.3.2.1 applies . (TS P) .

3.3.1.4 Cooling Water Systems - Spe cification 3.0.1 applies.

a. Two nuc lear service closed cycle cooling water pumps must be OPERAB LE.

b. Two nuc lear service river .water pumps must be OPERABLE.

c. Two decay heat closed cycle cooling wat er pumps must be OPERABLE.

d. Two dec ay heat river water pum ps must be OPERABLE.

e. Two reactor building emergency cooling river water pumps must be OPE RABLE.

3.3.1.5 Engineered Safeguards Valv es and Interlocks Associated with the Specifications 3.3.1.1, 3.3.1.2, 3.3.1.3, Systems in 3.3.1.4 are OPERABLE. Specificatio applies. n 3.0.1 3-22 Amendment No. 33, 80, 98, 137, 174

,19 0,2 11, 225 ,22 7, 263 278

\

3.3 EMERG ENCY CORE COOLINGYREACTOR BUILDING EMERG ENCY COOLIN REACTO R BUILDING SPRAY S STEMS (Contd.) G AND 3.3.2 Maintena nce or testing shall be allowed during reactor operation on any compone makeup and purification, decay heat, RB emergency cooling water, RB spray, BWST nt(s) in the instrumentation, (?r cooling water systems which wilr not remove more level system from service. Components shall not be removed from service so than one train of each train is inoperable for more than 72 consecutive hours. If the system is not that the affected system requirements of ~pecification 3.3.1 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the reactor shall be placed restored to meet the a HOT SHUTDOWN condition within six hours. in 3.3.2.1 If the CFT boron concentration is outside of limits, or if the TSP baskets contain outside the limits specified in 3.3.1.3.b, restore the system to operable status amounts of TSP the system is not restored to meet the requirements of Specification 3.3.1 within within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If reactor shall be placed in a HOT SHUTDOWN condition within six hours. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the 3.3.3 Exceptions to 3.3.2 shall be as follows:

a. Both CFTs shall be OPERABLE at all times.

b. Both the motor.operated valves associated with the CFTs shall be fully open at all times.

c. One.reactor building cooling fan and associated cooling unit shall be permitted to serv1ce for seven days. be out-of-3.3.4 Prior to initiating maintenance on any of the components, the duplicate (redunda shall be verified to be OPERABLE. nt) component Bases The requirements of Specification 3.3.1 assure that, before the reactor can be made engineered safety features are operable. Two engineered safeguards makeup critical, adequate removal pumps and two decay heat removal coolers (along witti their respectivepumps, two decay heat compone nts) are srecified . However, only one of each is necessary to supply emergen cooling water systems reactor in the even of a loss-of-coolant accident. Both CFTs are required because cy coolant to the insufficient inventory to reflood the core for hot and cold line breaks (Referen a single CFT has ce 1).

For a Decay Heat Remova l/ Low Pressure Injection train to be OPERABLE, the of performing automatic injection from the BWST, recirculation and cooling of thesystem must be capable and post LOCA reactor vessel boron concentration control. Train A of post LOCAreactor building sump, concentration control includes remote operation of the Decay Heat Pressuri reactor vessef boron V-4. Train B of post LOCA reactor vessel boron concentration control includes zer Spray Isolation Valve RC-Decay Heat Drop Line Suction and Containment Isolation Valves DH-V-1, DH-V-2 remote operation of the breaker), DH-V-3, and DH-V-12B. DH-V-12B Is Locked Open, to allow drop line (after closing in its post LOCA boron concentration control even if the valve is inaccessible. flow path to be used for Manage ment of gas voids is important to Emergency Core Cooling System and Spray System OPERABILITY. Reactor Building The operability of the borated water storage tank (BWST) as part of the ECCS ensures supply of borated water is available for injection by the ECCS in the event that a sufficient The limits on BWST minimum volume and boron concentration ensure thatof1)a sufficien LOCA (Reference 2).

available within containment to permit recirculation cooling flow to the core, and t water is remain at least one percent subcritical following a Loss-of-Coolant Accident (LOCA). 2) the reactor will The contained water volume limit of 350,000 gallons includes an allowance for water of tank discharge location and sump recirculation switchover setpoint. Redunda not usable because borated water supply at a temperature greater tha.n 40°F. nt heaters maintain the The BWST can be placed on cleanup path, or redrculation path for weekly surveilla concentratiori., on a temporary basis, until the end of the Fuel Cycle 21 operation nce testing for boron has been perrormed that concluded the cleanup and recirculation seismic Class . A seismic evaluation maintain pressure bol{ndary integrity during a Safe Shutdown Earthquake (SSE).II piping paths would BWST would maintain its safety functions auring an SSE. The limiting condition The seismic Class I BWST cleanup operation is a total duration of not more than 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months (30 da1s)for operation (LCO) for Refueling Outage and is a total duration of not more than 1440 hours0.0167 days 0.4 hours 0.00238 weeks 5.4792e-4 months (60 days prior to Fall 2015 operation. BWST Cleamm can be started and stopped at any time as during Fuel Cycle 21 exceeded. The LCO for BWST recirculation operation is limited to the long as he total durations are not time it takes to adequately recirculate the BWST volume to perform the boron sampling surveillance, which is approximately 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months per week. The temporary LCOs are in effect to allow time for a permane nt solution to the issue of interconnecting seismic Class I and II piping during BWST cleanup and recircula tion operation.

The Reactor Building emergency sump pH control system ensures a sump pH between during the recirculation phase of a postulated LOCA. A minimum pH level of 7.3 7.3 and 8.0 potential for chloride induced stress corrosion cracking of austenit1c stainless steel is required to reduce the retention of elemental iodine in the recirculating fluid. A maximum pH value of 8.0 and assure the minimizes the

. 3-23 Amendm ent No. 449,4a- 7,4ea,47 3,227,2 29,2eJ,2 73,~.m. ECR 14-00208

3.3 EMERGENCY CORE COOLING, REACTOR BUILD ING EMER GENC Y COOLING AND REACTOR BUILDING SPRAY SYSTEMS (Cont d.) .

Base s (Cont'd.}

formation of precipitates that may migrate to the emerg ency sump and minimizes post-LOCA hydrogen generation. Trisodium.phosphate dodecahydrate is used because of the high humid ity that may be prese nt in the Reactor Building during normal operat ion. This form is less likely to absorb large amounts of water from the atmosphere.

All TSP baskets are located outside of the secondary shield wall in the Reactor Building basement (El.

281 '-0"). Therefore, the baskets are protected from the effects of credible internal missiles inside the shield wall. The designated TSP baske t locations ensur e that the baskets are not impacted by the effect of potential LOCA jet impingement forces and pipe whip. *

Maintaining MUT pressure and level within the limits of Fig. 3.3-1 ensures that MUT gas will not be drawn into the pumps for any design basis accident. Preventing gas entrai nmen t of the pumps is not depen dent upon opera tor actions after the event occurs.

The plant operating limits (alarms and procedures) will include margins to accou nt for instrument error.

The post-accident reactor building emergency coolin g may be accomplished by three emerg ency cooling units, by two spray systems, or by a combination of one emerg ency cooling unit and one spray system. The specified requirements assure that the required post-accident components are available.

The iodine removal function of the reactor building spray system requires one spray pump and TSP in baske ts located in the Reactor Building Basement.

The spray system utilities comm on suction lines with the decay heat removal system . If a single train of equipment is removed from either system, the other train must be assured to be opera ble in each system.

When the reactor is critical, maintenance is allowed per Specification 3.3.2 and 3.3.3 provided requirements in Specification 3.3.4 are met which assure operability of the duplic ate components.

Maint enanc e as described here includes preventative and corrective type activities. The specified mainte nance times are a maximum. Operability of the specified components shall be based on the satisfa ctory completion of surveillance and inservice testing and inspection required by the INSERVICE TEST ING PROGRAM and Technical Specification 4.2 and 4.5.

The allowa ble maintenance period of up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months may be utilized if the operability of equipm ent redundant to that removed from service is verified based on the results of surveillance and inservice testing and inspection require d by the INSERVICE TEST ING PROG RAM and Technical Specification 4.2 and 4.5.

In the event that the need for emerg ency core cooling should occur, operation of one makeup pump, one decay heat removal pump , and both core flood tanks will protect the core. In the event of a reacto r coolant system rupture their operat ion will limit the peak clad temperature to less than 2,200 °F and the metal-water reaction to that representing less than 1 perce nt of the clad.

Two nucle ar service river water pump s and two nuclea r service closed cycle coolin g pump s are required for normal operation. The normal operating requirements are greater than the emerg ency requirements following a loss-of-coolant.

REFE RENC ES (1) UFSAR, Section 6.1- "Emer gency Core Cooling System (2) UFSAR, Section 14.2.2.3 - "Large Break LOCA" 3-24 Amen dmen t No. BO, 148, 1a7, 165, 178, 227, 263, ECR TM 09 OQ160, 290

CONTROll ED COPY FIGURE 3.3-1 Mak eup Tan k Pres sure vs Leve l Limi ts (Inst rume nt Error NOT Included) 110 105 _,____,___ _ _ _~-- -"- "--... .i... --'-* ---; -~~ ----

-*-- ..... .._- --'-- ---i 100 i. J*

95 90 __ ____..__ UNR ESTRICT OPl:RATINGED -+- --- --- 1-- --t ---

r-- -::. a REGION ui Ill 85

=u

§. j.

ai

> i Ill 80

..J C

ca 75

~

70 65 I ;

PROHIBITED l REGfON i 60 I

\

55 i. i I '

\

I I l I

I I i I !

I

! iI 50 10 15 20

.I 25 30

i. .l. * *

i 35 40 41$ 50 55 60 65 MU Tank Pressure (psig )

3-24a Amendment No. 227,

CONTR llE D C PY.

3.4 DECAY HEAT REMOVAL <OHR) CAPABILITY Applicability Applies to the operat ing status of systems and components that function to remove decay heat when one or more fuel bundles are located in the reactor vessel.

Objective To define the conditi ons necessary to assure continuous capability of OHR.*

Specification 3.4.1 Reacto r Coolan t System (RCS) temperature greater than 250 degrees F.

3.4.1.1 Three indepe ndent Emergency Feedwater (EFW) Pumps and two redundant flowpaths to each Once Through Steam Generator (OTSG) shall be OPERABLE **

with: *

a. Two EFW Pumps, each capable of being powered from an OPERABLE emerg ency bus, and one EFW Pump capable of being powered from two OPERA BLE main steam supply paths. *

(1) With one main steam supply path inoperable, restore the inoperable steam supply path to OPERABLE status within 7 days or be in COLD

SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(2) With one EFW Pump or any EFW flowpath inoperable, restore the inoperable pump orflowp ath to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be

£.

in COLD SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(3) With one main steam supply path to the turbine-driven EFW Pump and one motor-driven EFW Pump inoperable, restore the steam supply or the

motor-driven EFW Pump to OPERABLE status within 24 hour~ or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTD OWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . *

(4) With more than one EFW Pump or both flowpaths to either OTSG inoperable, initiate action immediately to restore at least two EFW Pumps and one flowpath to each OTSG:

These requirements supplement the requirements of Specifications 3.1.1.1.c, 3.1.1.2, 3.3.1 .

and 3.8.3. -

- HSPS operability is specified in Specification 3.5.1. When HSPS is not required to be OPERABLE, EFW is OPERABLE by manual control of pumps and valves from the Control Room.

3-25 Amend ment No. 4,7B,9B,119,124,162,190,211, 242

CONTROLLED COPY 3.4 DECA Y HEAT REMOVAL <DHR) CAPABILITY (Continued)

Notes:

1. Specification 3.0.1 and all other actions requiring shutdown or changes in REACTOR OPERATING CONDITIONS are suspended until ~t least two EFW Pumps and one EFW flowpa th to each OTSG are restored to OPERABLE status.

2. While performing surveillance testing, more than one EFW Pump or both flowpaths to a single OTSG may be inoperable for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provided that:

(a) At least one motor-driven EFW Pump shall remain OPERABLE, and (b) With the reactor in STARTUP, HOT STANDBY, or POWER OPERATION, a designated qualified individual who is in communication with the control room shall be continuously stationed in the immediate vicinity of the affected EFW local manual valves. On instruction from the Control Room, the individual shall realign the valves from the test mode to their operational alig nmerit.

b. Four of six Turbine Bypass Valves (TBVs) OPERABLE. With more than two TBVs inoperable, restore operability of at least four TBVs within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

c. The Condensate Storage Tanks (CSTs) OPERABLE with a minimum of 150,000 gallons of condensate available in each CST.

'(1) With a CST inoperable, restore the CST to operability within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and COLD SHUT DOW N within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(2) With more than one CST inoperable, restore at least one CST to OPERABLE status or be subcritical within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , in HOT SHUT DOW N within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3.4.1.2.1 With the Reactor between 250 degrees F and HOT SHUT DOWN, and havin g been subcritical for at least one (1) hour, two (2) Main Steam Safety Valves (MSSVs) per OTSG shall be OPERABLE. Wrth less than two (2) MSSV s per OTSG OPERABLE, restore at least two (2) MSSVs to OPER ABLE status for each OTSG within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in COLD SHUTDOWN within the following

~ 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.4.1.2.2 With the Reactor between HOT SHUTDOWN and 5% powe r, and having been subcritical for at least one (1) hour, two (2) MSSVs per OTSG shall be OPERABLE provided the overpower trip setpoint in the RPS is set to less than 5% full power. With less than two (2) MSSVs per OTSG OPERABLE, restore at least two (2) MSSVs to OPERABLE status for each OTSG within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3-26 Amen dmen t No. 4, 78, 119, 125, 133, 242

3.4 DEC AY H~AT R~M OVA L {OHR) CAPABILITY (Con tinued) 3.4.1 .2.3 Ex'cept qS provided in Specification 3.4.1 ;2:2 9bov~. when the Reactor is above HQT. SHUTD0W.N, -seven (7) MS$Vs per QTSG shall .be OPERABLE. If either OTSG has ie~s than seven (7) MSSVs ttiat are OPE:

FV\BLE, then reduce the .

power and reset the maximum overpower trip setp6 int as follows: .

Minim. um.

. .....Num

.. .. bar I.'. of Ma:X,imuni. Overpower MSSVs Qpegtple on Trip $etpo int Each OTSG Wo of Rated Powe r}

7 see Table 2.3-1 6 85;'1*

5 .70.1 4 55.1 With less than four (4) MSSVs OPERABLE per OT~G

, restore to a condition with at reast tow (4) MSS Vs on.each OTSG to OPER ABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be iii HOT SHU TDO WN Within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3.4.2 Res* temperature less than or equal to 250 degrees F.

3.4.2.1 At least two 'of the*fqllowing means for maintaining OHR capability shall be OPERABLE and at least one shall be in operation exce pt as allowed *by Speciffcations 3.4.2.2, 3.4.2 .3 and 3.4.2.4.

a. OHR String '{Loop "A").

b. OHR String (Loop "B").

c. RCS Loop* "A" and .its associated OTSG with an EFW Pump and a flowpath.

d. RCS Loop "B and its associated OT~G with ari.EF W Pump and a fJowpath.

With less* than the abov£:l *requ,ired means for maintainir

,g OHR capability OPERABLE, immediately initiate corrective action to return th.e required loops to OPERABLE *status as soon as possible.

3.4.2.2 Operation of the means for OHR may be suspended provided the core outlet temperature is*maintai.ried below saturation temperatq re.

3.4.2 .3 The number of means for.OHR required to.be OPER ABLE per Specification* 3.4.2.1 may be reduced to .one provided thatf he Reactor is*in condition with the Fuel Transfer Canal water level great a REFUELING SHUTDOWN er than or equal to 23 feet above the Reactor Vess erflal ige.

3.4.2 .4 Specification 3A2. 1 does not apply when either of the following conditions exist:

a. Decay heat generation is less than 188 KW with the RCS full.

b. Decay heat generation is less than 100 KW with the RCS drained down for maintenance.

3-26a Amen dmen t No. 119, 126, 1aa, 220, 242, 277

C NTROllED COPY 3.4 DECAY HEAT REMOVAL <DHR} CAPABILITY (Continued)

Bases A reactor shutdow n following power operation requires removal of core decay heat.

Normal OHR is by the OTSGs with the steam dump to the condenser when RCS tempera ture is above 250 degrees F and by the OHR System _below 250 degrees F. Core decay heat can be continuously dissipat ed up to 15 percent of full power via the steam bypass to the conden ser as feedwater in.the OTSG is converted to steam by heat absorption. Normally, the capability to return feedwa ter flow to the OTSGs is provided by the main feedwater system.

The Emerge ncy Feedwa ter (EFW) System supplies adequate feedwater to the OTSGs at accident pressur es, removing heat from the Reactor Coolant System (RCS) to support safe shutdown of the reactor when the normal feedwater supply is unavailable. EFW is not required for normal plant startup and shutdown.

The turbine-driven EFW Pump and two motor-driven EFW Pumps take suction from the C_ondensate Storage Tanks (CSTs) and deliver flow to a common discharge header.

Flowpat h redundancy is provide d for those portions of the EFW flowpath containing active compon ents between the pumps ~nd each of the OTSGs. Each EFW line to an OTSG includes two redundant flowpaths, each equipped with an automatic control valve (EF-V-30A/B/C/

D) and a

manual isolation valve (EF-V-52A/B/C/D). Each redundant flowpath is capable of providing adequate flow to the associated OTSG. Heat removed from the OTSGs returns to the Main Condenser through the Turbine Bypass Valves (TBVs) or discharges to the atmosph ere through the. Main Steam Safety Valves (MSSVs) and/or the Atmosp heric Dump Valves (ADVs).

An unlimited supply of river water to the EFW Pumps is available using either'of the two Reactor Building Emerge ncy Cooling Water (Reactor River Water) Pumps (RR-P-1A/B).

Redundant main stec:1ni supply paths are provided to the turbine-driven EFW Pump for certain events involving loss of one steam supply (e.g., main steam and feedwater line breaks).

An operable Main Steam supply path delivers steam to the turbine-driven EFW Pump upon HSPS actuation or by operato r action from the control room when HSPS is not required

. During low pressure conditions, additional steam supply paths from Main Steam (MS-V-10A/B) or Auxiliar y Steam can be made available to the turbine-driven EFW Pump as necessary.

During design basis events the EFW System can withstand any single active failure and still perfonn its function. The limiting design basis accident for the EFW System is a loss of feedwater event with off-site power available. In the event of a loss of all AC power, which assumes multiple single failures, the turbine-driven EFW Pump alone delivers the necessa ry EFW flow. Consideration of additional failures in the EFW System or Heat Sink Protection System (HSPS) is not required for this event. Additionally, the EFW System capabili ties are sufficient to deliver the required flow in licensing basis events (e.g., ATWS failure to trip events,

Generic Letter 81-14 seismic events, and the Station Blackou t event).

The most limiting EFW flow requirement is met when at least two EFW Pumps are operabl e and at least one EFW flowpath to each OTSG is operable. When three pumps and two flowpath s to each OTSG are operable, the EFW System can withstand any single active failure.

Exampl es of single acti~e failures include: failure of any one EFW Pump to actuate, failure of one HSPS train to actuate, or failure of one redundant flowpath to either OTSG. Initially after a shutdow n, any two EFW Pumps are required to remove RCS heat with one pump eventually sufficing as the. decay heat production rate ditninishes.

3-26b Amendment No. 119,12 4,125,1 33,157 ,190,24 2

3.4 DECAY HEAT REMOVAL (OHR) CAPABILITY (Con

-,. t,.

timie. d)

Bases (Continued)

If EFW .were requirec:f during surveillance testing, rninor op.~rator action (e.g., opening a local isoiation valve or mariiplll~t,i.ng a control. switch from the operability of, t~e req1.,1ired.JJ.j,irnps *or flowpaths: Anex control rqom) rrtaYb.e nee~ec:1 to restore ce'ptio.n ,to p~rmit more;than qne EFW Pump* oc~oth,.££FW flowpaftfa.:to a:'singte:otSG to,be Jr;ioperable fqr upto 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .during surveillance t~~ting: recf~ire$ 1) aU~a.st -one: n,otif yen -~FW Pump :operab_lE;l; *ari<:L2) ~~

indi.Yicj_Oc!l)tjvolv~cfinJh~-ta~~ 9f t~stfr1g th:e. E:FWr.;qri contra! fp9m arid .sf~t!~heo irfthe'Jmmeoiat~*-Vicihit

'_ pys_t~ m *rnvst be' in cpmrnunj~atioh ~itli :the the. jndiyidtJctl, is,p~rl'.TlittE!dtq, b~)nvblvec:i in th. er,tes y ,Qf th1J{c 1ff~cte g EFVI{ J!9w,path. v~hies, Thus t activifies l;>y ti;ikihgJest aat~.a.nd*:*h1s .

movern'enfi$ festri'ct~d tfrthe area ofthe EFW Pu conducted. * *

nip and valve rooms where ttfa testing* ls being Theiallowed 9ction times c1re reas,on~blei based oh operafirig experien9e, to rea,ch the required ptant*opetating conditions from fl.ill power in* an orde

. system,s. *. Witnol.lt at lea9.t two EFW,.Pumps and rly marihe'r and \yit~ol.it tHat1~*ngi,ng plant

_one.EFW flowp~th fo *each OTSG oper~ble, *the requfred action>:is. to im_n:iediately restore EF.W cor'n pc:ments to. pperqble;statusi *cfn~* c;ill actions r~q4irihp §hUtd9wn qr.~r~hg_~s in Reactor 'Qperatin~

two *EFW pumps or .noflovvpatti' to* eitherOTS

/Q,bhditic,d are susp~nd$d;:* With:less than Gopefable; th.e unit is iit a seriously degraded condition* with ho safety related means for conductin g ,a cooldOVil'J- In. such aJ~ohditipn, the* unit should not b.eperturbed by ariy action, it:1cludi_l1g T~e*:s-~rio~sne*ss 9f this c6_h~iti6n r~quiresthat actio a p.qwer ch~ng~* .wh.iCh mightresylt-in i:{trip.

o,be,$tarted imnjediately.to restore. EFW c6mponents,.tp4operable status: TS 3.0.1 is noLa safe condition. pplic able;.as itcoulq f6rpe the unit irito a less The EFVV system actuates*ori: 1) loss of ~IMout Re~ctor Coolant Punips, 2) IOS$ of bath Main FeedVfat~r:PurtJPSj 3) low OTSG WatetlevEll; or 4):hi activeJalh.ire in the HSPS*wiii:r,either inadvertently.i gtr f{eagtor\\euildiiig pressure. A single nitiate the EF\IV system'.nor isolate*theMain Feedwat~f.'SY$tem. OTSG:V,fat~r level is c~mtroile controllE!d:rfa{nµal!y, if n~ces~ary. . d'aut droatiGally by the HSPS system or can be I

The MS$y~,wfll_bt3 a,.,ble, to)~li:eye to ptn\osphere the total :~team .fl~w rf -necess~ry. .

Specifi¢aydrl3:4.1.2<3 proviqes ,limiting co!Jditions of op_~~a,tion if mo_re than;tw6 MSSVs are inop~rat)te*oh *a sirigle_OTSG. The:power level ana overpower trip s~tpolnt m't.jst be rec:Juceq, as stated in Specff1cation 3.4;1 .2.is uch tharthe rer)'laining MSSVs*can prevent secoJ]cfa,ry,sysfem overpreS~ure oh a turbine trip.,

The turbine trip 13Veh{ is the limiting event 'in tetms.,of pea:k):;ecor,c;:lary sy~tem pre~~ure

. Anaty~es have*snovvn t~*at overpressurewlll not occul"if a turbine tdp occfa s Wit~Mfr'or:iessrv;ssVs out of service on*.each OTSG.and*an,initi91pov.,er leve:1 less-~han brequq.l to 102%.df 2772iMWth ..

Having fy1~:fSVs qut ot:service aa,aitowed*by Sp~C ifiGa~ion .3:4.1.2.'3 d,oes not adversely impact the'.transferit progression of the remaining Safety Analysis events.

Below 5°(o power, only a minJmum ru.imber of MSS Vs nel:3d to .be operable as stat~d in Specification*s 3.4.1 ..2.1 *and 3.4.1.2.2. This is,toJ Jrovide OTSG':overpressure profectior1 during hot functional ~esting arid low power physics testi ng. Additiqnally;, ~ben the. Reactor is between hot shutdown and 5% full power operation, the over power trip setppint in the RPS shall be *set to less than,5% as is sp~cifi~d in Specification 3A.1

.2,2. The.minimum number of MSSVs r~qi.Jiredto be op~rable.atlgws margin for testing Without Jeopardizing plant safety; Plant sp~cific,analysis ~hows that one MSSV,is sufficien lto relieve reactorcOolant pump heat and stored energy when the reactor has been subcritica l by 1% delta K/K for at least one hour.

3-~6c Amendment No. :Z8*, 119, 126,13;3, 167 ,2 0,24 4 2,261,277

3.4 DECAY HEAT REMOVA L (OHR) CAPABILITY (Continued)

Bases (Continued)

Other plant analyses show that two (2) MSSVs on either OTSG are more than sufficient to relieve reactor coolant pump heat and stored energy when the reactor is below 5% full power operation but had been subcritical by 1% delta K/K for at least one hour subsequent to power operation above 5% full power. According to Specification 3.1.1.2a, both OTSGs shall have tube integrity wheneve r the reactor coolant average temperature is above 200 degrees F. This assures that all four (4) MSSVs are available for redundancy. During power operations at 5%

full power or above, if MSSVs are inoperable, the power level must be reduced, as stated in Specification 3.4.1.2.3 such that the remaining MSSVs can prevent overpressure on a turbine trip.

The minimum amount of water in the CSTs required by Specification 3.4.1.1.c, provides at least 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of OHR with steam being discharged to the atmosphere. This provides adequate time to align alternate water sources for RCS cooldown. After cooling to 250 degrees F, the OHR System is used to achieve further cooling.

When the RCS temperature is below 250 degrees F, a single OHR String (Loop), or single OTSG with an EFW Pump and a flowpath capable of supporting natural circulation is sufficient to provide removal of decay heat at all times following the cooldown to 250 degrees F. The OHR String (Loop) redundan cy required by Specification 3.4.2.1 is achieved with independent active components capable of maintaining the RCS subcooled. A single OHR flowpath with redundant active. components is sufficient to meet the requirements of Specifications 3.4.2.1.a and 3.4.2.1.b. The requirement to maintain two operable means of OHR ensures that a single active failure does not result in a complete loss of OHR capability. The requirement to keep a OHR Loop in operation as necessary to maintain the RCS subcooled at the core outlet provides the guidance to ensure that steam conditions which could inhibit core cooling do not occur.

Managem ent of gas voids is important to OHR System OPERAB ILITY.

With the Reactor Vessel head removed and 23 feet of water above the Reactor Vessel flange, a

large heat sink is available for core cooling. In this condition, only one OHR Loop is required to be operable because the volume of water above the Reactor Vessel flange provides a large heat sink which would allow sufficient time to recover active OHR means.

Following extensive outages or major core off-loading, the decay heat generation being removed from the Reactor Vessel is so low that ambient losses are sufficient to maintain core cooling and no other means of heat removal is required. The system is passive and requires no redundant or diverse backup system. Decay heat generation is calculated in accordance with ANSI 5.1-1979 to determine when this situation exists (Reference 4).

REFERENCES (1) UFSAR, Table 6.1 ECCS "Single Failure Analysis" (2) UFSAR, Section 9.5 - "Decay Heat Removal System" (3) UFSAR, Section 10.6 - "Emergency Feedwater System" (4) TMI Unit 1 Calculation C-3320-85-001, "RCS Decay Heat Removal-Ambient Losses,"

Revision 0, February 28, 1985 3-26d Amendm ent No.-242,!2+1, 285

l,**'"'c

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3.5 INSTRUMENTATION SYSTEMS 3.5. 1 OPERATIONAL SAFETY INSTRUMENTATION Appl icab n ity Applies to uni t instrumentation and control systems.

Objective To deli nea te the conditions of necessary to assure reac tor safethe ty.

unit instrumentation and safe ty circ uits Specifications 3.5. 1.1 The reactor shall not be in a star tup mode unless the requirements of Table 3.5-1, Coluor in a crit ica l stat e except as provided in Table 3.5- 1, Column mn "A" and "B" are met, app lies . "C". Spe cifi cati on 3.0.1 3.5 .1.2 The key operated channel bypass switch asso protection channel may be used to lock the ciated with each reac tor

---the ~ntripped stat e as indicated by a ligh reac tor trip module in shal~ be locked in this untripped stat e at t. Only one channel operation at rated power shall be permitted any one time. Unit 3.5 -1, Column "A". Only one channel bypass to continue with Table the control room. key sha ll be kept in 3.5 .1.3 In the event the number of protect the lim it given under Table 3.5~1, ion Colu channels operable fall s below mn "A", operation sha ll be limited as spe cifi ed in Column "C". Spe cifi cati on 3.0.1 app lies .

3.5 .1.4 The key operated shutdown bypass switch asso rea cto r protection channel shall not be used ciated with each operation {except for required maintenance during reac tor power or test ing ).

3.5 .1.5 During START-UP when the intermediate rang sca le, the overlap between the inte e instruments come on range instrumentation sha ll not be less than rme diat e range and the source one decade.

3.5 .1.6 During START-UP, HOT STANDBY or POWER OPER a control rod drive trip breaker is ino p,raATION, in the event tha t place the breaker in trip . Specificatio~ ble; within one hour .

3.0.1 app lies .

3.5 .1.7 During START-UP, HOT STANDBY or one of the control rod driv e tripPOWE brea R OPERATION, in the event tha t ker diverse trip feat ure s (shunt trip or undervoltage trip atta~hm ent) is inoperable:

a. Restore to OPERABLE stat us within 48 hou rs or

b. Within one additional hour place the brea ker in trip .

Spe cifi cati on 3.0 .l app lies .

3-27 Amendment No. ~i, ~-. JiJ, 189

3.5.1.7.1 Power may be restored through the breaker with the failed trip feature for up to two hours for surveillance testing per T.S.

4.1.1.

3.5.1.8 Deleted 3.5.1.9 The reactor shall not be in the Startup mode or in a critica l state unless both HSPS actuation logic trains associated with the Functional units listed in Table 3.5-1 are operable except as provided in Table 3.5-1,D.

3.5.1.9.1 With one HSPS actuation logic train inoperable, restore the train to OPERABLE or place the inoperable device in an actua ted state within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With both HSPS actuation logic trains inoperable, restore one train to OPERABLE within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Bases Every reasonable effort will be made to maintain all satety instrumentation in operation. The reactor trip, on loss of feedwater may be bypassed below 7%

reactor power. The bypass is automatically removed when reactor power is.

raised above 7%. The reactor trip, on turbine trip, may be bypassed below 45%

reactor power (Reference 1). The safety feature actuation system must have two analog channels functioning correctly prior to startup.

The anticipatory reactor trips on loss of feedwater pumps and turbine trip have been added to reduce the number of challenges to the safety valves and power operated relief valve but have not been credited in the safety analyses ...

--- ~ -

Operation at rated powe r is permitted as long as the system s have at least the redundancy requirements of Column "B" (Table 3.5-1). This is in agreement with redundancy and single failure criteria of IEEE 279 as described in FSAR Section 7.

There are four reactor protection channels. Normal trip logic is two out of four. Minimum required trip logic.is one out of two.

3-27a Amendment No. 123, 124, 136, 167, 189, 273

The four reactor protection channels were provided with key operated bypass switches to allow on-line testing or maintenance on only one channel at a time during power operation. Each channel is provided alarm and lights to indicate when that channel is bypassed. There will be one reactor protection system bypass switch key permitted in the contro l room.

Each reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdo wn bypass switch is being used.

Power is norma lly supplied to the control rod drive mecha nisms from two separate parallel 460 volt sources through four AC trip breake rs, designated A, B, C and D. The breaker undervoltage trip coils are powered by RPS chann els A, 8, C and D, respectively. From these circuit breakers, the CAD power travels through voltage regulators and stepdown transformers to complete redundant power buses that feed the CAD Single Rod Power Supplies (SRPSs) A and B.

Two AC breakers (A and C) are arranged in series to feed SAPS power bus A, and the other two AC breakers (Band D) are in series to feed SAPS power bus B. Opening at least one circuit breaker in each of the two parallel paths to the SAPS will cause a reacto r trip, in a one-out-of-two taken twice logic.

Either path can provide sufficient power to operate all CRDs.

Diverse trip features are provided on each breaker. These are the

'-unde rvolta ge relay and shunt trip attachment. Each trip feature is tested separately. Failure of one breaker trip feature does not result in loss of redun dancy and a reasonable time limit is provided for correc tive action.

Failure in the untripped state of a breake r results in loss of redundancy and prompt action is required.

Failure of both trip features on one breaker is considered failure of the breaker.

Power may be restored through the failed breaker for a limited time to perform required testing.

The 4.16kv ES Bus Undervoltage Relays detect a degrad ed voltage or Loss of Voltage on the associated ES Bus. Detection of low voltag e will separate the ES bus from the offsite power, initiate load shedding and* start the associated diesel generator. The relays do not function during design basis events where acceptable offsite voltage is available. If the voltage relays on one train are not operable, the time permitted for repair is

.consistent with other safety related equipment. If both trains are affected then shutdown is initiated in accordance with Specification 3.0.1 since automatic response of the diesel generator is required to assure completion

of the safety function if offsite power is degraded or lost.

Automatic initiation of EFW is provided on loss of all reacto r coolant pumps, loss of both main feedwater pumps, low OTSG level, and high reactor building pressure. High reactor building pressure would be indicative of a loss of coolan t accident, main steam line or feedwater line break inside the reactor building. Operability of these instruments is require d in order to assure that the EFW system will actuate and control at the appropriate OTSG level withou t operator action for those events where timely initiation of EFW is required.

Automatic isolation of main feedwater is provided on low OTSG pressure in order to maintain appropriate RCS cooling (minimize overco oling) following a loss of OTSG integrity and minimize the energy released to the Reactor Building atmosphere.

3-28 Amen dment No. 78, 123, 124,1 93,27 3

WEDOOPY 1

li~ PS instrument op er ili ty sp ec cr ite rio n for the EFWabsys ifi ed meets the sin gle fa ilu fo r automat~c EFW in iti ati ontem . Four instrument channels are re bu ild ing pr es su re, and for on OTSG low 1evel and high rea provided cto r OTSG pr es su re. Normal tri p automatic main feedwater iso lat io n on low the 4 ch an ne ls. in bypass, log ic is two out of four.

a sec With one of se rv ice (placed in *:he tri pp ond channel may be taken out of wi11 prevent ac tua tio n of ed po sit ion ) and no sin gle ac tiv e fa ilu re No sin gl e ac tiv e fa ilu re ofthe as so cia ted HSPS tra in ac tua tio n lo gi c.

HSPS tra in from operating to eit he r HSPS tra in will prevent the oth er supply EFW to both OTSGs.

REFERENCE (1) B&W Report No. BAW-189 An tic ipa tor y Reactor Tr3,ip "Basis fo r Raising Arming Threshold fo r on Turbine Tr ip, " Rev. 0, October 1985 3-28a Amendment No. tZl, it4 , !.8.5, 157

.<?;;/\ .

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3 m

, TAB LE 3.5-1 0.

3 INSTRUMENTS OPERATING CON DITI m ONS

?.

z Functional Unit (A) 9 (B)

Minimum Operable (C)

Mini mum Degree Ope rato r Action if Conditions Channels of Red unda ncy of Column A and B Cannot be Met A. Rea ctor Protection Syst em

1. Manual pushbutton 1 0 (a)

2. Pow er range instr ume nt channel 2 1 (a)

3. Intermediate range instr ume nt 1 0 channels (a) (b)

4. Source range instr ume nt 1

c.:>

I channels 0 (a) (c)

I\)

co

5. Rea ctor coolant temp erat ure 2 1 instrument channels (a)

6. Rea ctor Coolant 2 1 Pres sure -Tem pera ture (a)

Instrument channels

7. Flux / imba lanc e / flow 2 1 (a)

8. Reactor coolant pressure

a. High reac tor cool ant pressure 2 1 instr ume nt channels (a)

b. Low reac tor cool ant pressure 2 1 instr ume nt channel~ (a)

x: ,,

3

'D TABLE 3.5-1 (Cont'd)

i..

3

'D INSTRUMENTS OPERATING CONDITIONS r+

z 0 Funct i ona 1 Un it (A) (B) (C)

Minimum Operable Minimum Degree Operator Action if Conditions ISi Channels of Redundancy . of Column A and B Cannot Be Met ea A.

""" Reactor Protection System (cont'd)

~

' 'l,R

9. Power/number of pumps instrument channels

10. High reactor building 2

2 I

1 (a)

(a) pressure channels 00 I.O (a) Restore the conditions of Column (A) and Column (B) within one hour or place the unit in HOT SHUTDOWN within an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

w w

I (b~ When 2 of 4 power range instrument channels are greater than 10* perc_ent ful 1* power, intermediate range 0

instrumentation is not required.

(c) When 1 of 2 -intermediate range instrument channels is greater than 10*rn_amps, or 2 of 4 power range instrument channels are greater than 10 percent full power, source range instrumentation is not required.

@ ** 1.

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3 ID

~

TABLE 3.5-1 (Cont'd) 0.

3 ID

~

INSTRUMENTS OPERATING CONDITIONS M'

z 0

Functional Unit (A) (8) - ' (i:)

Minimum Operable Minimum Degree Operator Action if .Conditions CD I..O Channels of Redundancy of Column A and B Cannot Be Met B. Other Reactor Trips I. Loss of Feedwater (c) 2 I j< a}

2. Turbine Trip (c) 2 I :(b}

l (a) Restore the conditions of Column (A) and Column (8) within one hour or reduce indicated reactor power to less than 7% within an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(b) Restore the conditions of.Column (A) and Column (B) wi~hin one hour or reduce fndicated reactor power to less

(.J I

than 45% within an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(.J 0

DI (c) Trip may be defeated during low power physics tests.

§"

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11)

TABLE 3.5-1 (Cont'd)

D.

3 11)

INSTRUMENTS OPERATING CONDITIONS c-+

2 0 Functi.onal Unit (A) (B) (C)

Minimum Operable Minimum Degree Operator Action if Conditions Channels of Redundancy of Column A and B Cannot Be Met . I C. Engineered Safety.Features

1. Makeup and Purifica tion System (high pressure injectio n mode) 0

a. Reactor Coolant Pressure 2 l(b) (a) b.

Instrument Channels Reactor Building 4 psig.

Instrument Channels 2

~

l(b) (a) .~

w I

w 2.

c. Manual Pushbutton (also actuates Low Pressure Injectio n)

Decay Heat System (low 2 '\*,. N/A ) (g)

I~

~

~ pressure injectio n mode)

a. Reactor Coolant Pressure 2 l(b) (a)

Instrument Channels

b. Reactor Building 4 psig 2 l(b) (a)

Instrument Channel~

c. Reactor Coolant Pressure I 0 Open circuit breaker at MCC D.H. Valve Interloc k for DH-VI or DH-V2 with the Bistable affected valve in the closed position within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or maintain R.C. pres~ure less than I

350 psig.

~;'~

\

"'**-*'"'~*.

TABLE 3.5-1 (Cont'd)

INSTRUMENTS OPERATING CONDITIONS

~

m Functional Unit (A) (B) (C)

Minimum Operable

~ Channels*

Minimum Degr ee Operator Action if Conditions a::

t11 of Redundancy of Column A and B -Cannot Be Met

~- C. En&ineered Safety Features (cont'd)

~ 3. Reactor Building Isolation and

"-I Cooling System

!5

"-I

a. Reactor Bldg. 4 psig Instrument ~han nel 2 N l(b) (a)

~

00

b. -Manual Pushbuttons N i. 4 psig feature 2 0

s c!.a N

C.

ii. 30 psig feature Deleted 2 ~i /

/

(g)

d. Reactor Building 30 psig pressure switches 2 1 (c)

I

e. RCS Pressure less than 1600 psig 2 l(b) (a)

f. Reactor Building Purge Line Isolation 1 (AH- VIA and AH-V lD) High Radiation 0 (t)

4. Reactor Building Spray System

a. Reactor Building 30 psig pressure switches 2 1 (d)

/--\ ',

b. Spray Pump Manual Switches 2 1

/N/A ) (g)

5. 4.16KV ES Bus Undervoltage Relays /

a. Degraded Grid Voltage Relays 2 l (e)

b. Loss of Voltage Relay 2 1 (e)

q('c-\:

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3 ID CL TABLE 3.5-1 (Cont'd) 3 ID M'

INSTRUMENTS OPERATING CONDITIONS

z

.0 C. Engineered Safety Featu~es (cont'd)

'51 (a) Restore the conditions of Column (A) and Column (8) within one hour or place the reactor in HOT SHUTDOWN within an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1'Q (b) The minimum degree of redundancy ~ay be reduced to O up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance testing .

1-.a (c) The Operability requirement is two out of three pressure switches in each train, with a minimum degree of

.... redundancy of one,in each train.

1'Q

_., I. If the ainiaua conditions are not aet on one train, restore the function to OPERABLE within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or I.Cl w place the reactor in HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2. If the minillUM conditions are not met on either train, then place the reactor in HOT SHUTDOWN in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . *

(d) The Operability requireaent is two out of three pressure switches in each train, with a minimum degree of redundancy of one, in each train.

l. If the mini1111a conditions are not aet on one train, restore the function to OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or place the reactor in HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2. If the 11ini11u111 conditions are not met on either train, then place the reactor in HOT SHUTDOWN .in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(e) The operabi.lity requirement for the undervoltage relay, its associated auxiliary relays, and the timer I. If one 4.16 kv ES Bus does not meet the minimum conditions, restore the function to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in hot shutdown.within,an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . *

2. If *both 4.16 kv Buses do not ~et the minimum conditions, then restore at least one meet the minimum conditions within I hour or be in hot shutdown within *an additional 4.16 kv ES Bus to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(f) Discontinue Reactor Building purging and close AHV-lA, 18, IC, and ID within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

.. *:~.~:~-~~-

f ... '. }

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3 \(~/

ID

, TABLE 3.5-1 (Cont'd) a.

3 ID INSTRUMENTS OPERATING CONDITIONS

":z C. Engineered Safety Features (cont'd )

. 0 (g) The Operability requirement is for the manual actuation switch for the specifi ed OPERABLE. feature on each train to be I.* If the manual actuation switch on one train is inoperable, restore the switch to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . OPERABLE status within

2. If both manual actuation switches for that feature are inoperable, then place the SHUTDOWN in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLO SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . reactor in HOT I

,J

\)

T

§" TABLE 3.5-1 lCont'd).

111

I 0..

3 INSTRUMENTS OPERATING CONDITIONS 111 "z

.0 functional Unit {A)

Minimum Operable

{B)

Minimum Degree (C)

Operator Action if Conditions Channels of Redundancy of Column A and B Cannot Be Met D. Heat Sink Protection System

1. EFW Auto Initiati on

a. Loss of both Feedwater pumps N/A(b) . -N/A(b) (a)

b. Loss of a11 RC Pumps N/A(b) N/A(b) (a)

C. OTSG A Low Level 2-- l (a)

d. OTSG B Low Level 2 I (a)

e. High Reactor Building Pressure 2 I (a)

2. HFW I so 1ati on

a. OTSG A Low Pressure 2 I (a)

b. OTSG B Low Pressure 2 I (a}

J. EFW Level Control w a. OTSG A Level Control N/A(b) N/A(b) (a) w I

b. OTSG B Level Control N/A(b) N/A(b) (a)

N n

(a) Restore the cond.itions of Column {A) and Column (B) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or place the unit in HOT SUUT_DOWN within) the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . *

(b) Operability requirements are specified in Section 3.5Ll.9.

3.5.2 CONTROL ROD GROUP AND POWER DISTRIBUTION LIMITS Applicability This specification applies to power distribution and operation of control rods during power operation.

Objective To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejectio n, and to assure core subcriticality after a reactor trip.

Specification 3.5.2.1 The available shutdown margin shall not be less than one percen t delta K/K with the highest worth control rod fully withdrawn.

3.5.2.2 Operation with inoperable rods:

a. Operation with more than one inoperable rod as defined in Specif ication 4. 7.1 in the safety or regulating rod banks shall not be permitted. Verify SOM ~ 1% delta k/k or initiate boration to restore within limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The reacto r shall be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. If a control rod in the regulating and/or safety rod banks is declar ed inoperable in thewit hdraw n position as defined in Specification Paragraph 4.7.1.1 and 4.7.1.3, an evaluation shall be initiated immediately to verify the existen ce of one percent delta k/k hot shutdown margin. Boration may be initiated to increa se the available rod worth either to compensate for the worth of the inoperable rod or until the regulating banks are fully withdrawn, whichever occurs first. Simult aneously a program of exercising the remaining regulating and safety rods shall be initiated to verify operability.

c. If within one hour of determination of an inoperable rod as define d in Specification 4.7.1, and once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, it is not determined that a one percent delta k/k hot shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the HOT SHUTDOWN condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months until this margin is established.

d. Following the determination of an inoperable rod as defined in Specification 4. 7.1,

all rods shall be exercised within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and exercised weekly until the rod problem is solved.

e. If a control rod in the regulating or safety rod groups is declared inoperable per 4.7.1.2, and cannot be aligned per 3.5.2.2.f, power shall be reduce d to s 60% of the thermal power allowable for the reactor coolant pump combi nation within

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and the overpower trip setpoint shall be reduced to s 70%

of the thermal power allowable within 1O hours. Verify the potential ejected rod worth (ERW) is within the assumptions of the ERW analysis and verify peaking factor (Fa(Z) and F:H ) limits per the COLA have not been exceeded within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3-33 Amendment No. +7, 4-W, 211,2 46 278

f. If a control rod in the regulating group is declared inoperable per Specification 4.7.1.2, operation may continue provided that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the rods in the group are positioned such that the rod that was declared inoperable is maintained within average position limits of Specification 4.7.1.2. allowable group

g. If the inoperable rod in Paragraph "e" above is in groups 5, 6, or 7, the other rods in the group may be trimmed to the same position. Normal operation of 100 percent of the thermal power allowable for the reactor coolant pump combination provided that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the rod that was declared inoperable may then continue is maintained within allowable group average position limits in 3.5.2.5.

3.5.2.3 The worth of single inserted control rods during criticality is limited by the restriction of Specification 3.1.3.5 and the Control Rod Position Limits defined in Specification 3.5.2.5.

3.5.2.4 Quadrant Tilt:

a. Except for physics tests, the quadrant tilt, as determined using the full (FIS), shall not exceed the values in the CORE OPERATING LIMITS incore system REPORT.

The FIS is OPER ABLE for monitoring quadrant tilt provided the number of valid symmetric string individual SPND signals in any one quadrant is not less than the limit in the CORE OPERATING LIMITS REPORT.

b. When the full incore system is not OPERABLE and except for physics tilt as determined using the power range channels for each quadra tests quadrant nt (out of core detector system) (OCD), shall not exceed the values in CORE OPERATING LIMITS REPORT.

c. When neither detector system above is OPERABLE and, except quadrant tilt as determined using the minimum incore system (MIS), for physics tests, the values in the CORE OPERATING LIMITS .REPORT. shaH not exceed

d. Except for physics tests if quadrant tilt exceeds the tilt limit, allowa ble power shall be reduced 2 percent for each 1 percent tilt in excess of the tilt limit.

For less than tour pump operation, thermal power shall be reduced 2 percent below allowable for the reactor coolant pump combination for each 1 the thermal power percent tilt in excess of the tilt limit.

e. If quadrant power tilt exceeds the tilt limit then within a period of 10 quadrant power tilt shall be reduced to less than the tilt limit except hours, the for physics tests, or the following verifications and/or adjustments in setpoints and limits shall be made:

1. Verify Fa (Z) and F :H are within limits of the COLA once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and restore OPT to s steady state limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or perform steps 4 below. 2, 3, &

3-34 Amen dment No. 17, 29, 39, 40, 60, 90, 126, 142, 160, 211, 273

2. The protection system reactor power/imbalance envelope trip setpoin ts shall be reduced 2 percent in power for each 1 percent tilt, in excess of the tilt limit, power is equal to or less than 50% full power with four reactor or when thermal coolant pumps running, set the nuclear overpower trip setpaint equal to or less than 60%

full power.

3. The control rad group withdrawal limits in the CORE OPERA TING LIMITS REPOR T shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.

4. The operational imbalance limits in the CORE OPERATING LIMITS REPO RT shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.

f. Except for physics or diagnostic testing, if quadrant tilt is in excess of the maximum tilt limit defined in the CORE OPERATING LIMITS REPORT and using the applica ble detector system defined in 3.5.2.4.a, b, and c above, reduce therma l power to s15% FP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Diagnostic testing during power operation with a quadrant tilt is permitted provided that the thermal power allowable is restricted as stated in 3.5.2.4.d above.

g. Quadrant tilt shall be monitored on a minimum frequency of once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the OPT alarm is inoperable and at the freque ncy specified in the Surveillance Freque ncy Control Program when the alarm is operable during power operati on above 15 percen t of rated power. When OPT has been restored to s steady state limit, verify hourly for 12 consecutive hours, or until verified acceptable at :::.95% FP.

3-34a Amend ment Na. 29, as, as. 40, 46, 50,120 ,128, 142, 1 eO, 162,211, 274

3.5.2.5 Control Rod Positions

a. Operating rod group overlap shall not exceed 25 perce sequential groups except for physics tests. nt +/-5 percent, between two

b. Position limits are specified for regulating control rods.

exercising control rods, the regulating control rod insert Except for physics tests or ion/withdrawal limits are specified in the CORE OPERATING LIMITS REPORT.

1. If regulating rods are inserted in the restricted opera ting region, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attain FQ(Z) and ~:H ed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and shall be verified within limits once every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or be reduced to s power allowed by insertion limits.

power shall

2. If regulating rods are inserted in the unacceptable opera boration within 15 minutes to restore SOM to ~1% delta ting region, initiate regulating rods to. within restricted region within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months K/K, and restore power allowed by rod insertion limits. or reduce power to s

c. Safety rod limits are given in 3.1.3.5.

3.5.2.6 Deleted 3.5.2.7 Axial Power Imbalance:

a. Except for physics tests the axial power imbalance, as system (FIS), shall not exceed the envelope defined determined using the full incore in the CORE OPERATING LIMITS REPORT.

The FIS is operable for monitoring axial power imbal ance provided the number of valid self powered neutron detector (SPND) signals in any one quadrant is not less than the limit in the CORE OPERATING LIMITS REPORT.

b. When the full incore detector system is not OPERABLE axial power imbalance, as determined using the powe and except for physics tests r

detector system)(OCD), shall not exceed the envelope range channels (out of core defined in the CORE OPERATING LIMITS REPORT.

c. When neither detector system above is OPERABLE and, except for physics tests axial power imbalance, as determined using the minimum incore system (MIS), shall not exceed the envelope defined in the CORE OPERATING LIMITS REPORT.

d. Except for physics tests if axial power imbalance excee measures (reduction of imbalance by control rod move ds the envelope, corrective ments and/or reduction in reactor power) shall be taken to maintain operation within the envelope. Verify FQ(Z) and F:H are within limits of the CQLR once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months when not within imbalance limits.

3-35 Amendment No. 10, 17, 29, as, 39, 60, 120, 128, 142, 150, 179, 211, 219, 273 278

e. If an acceptable axial power imbalance is not achieved within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , reactor power shall be reduced to S40% FP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

f. Axial power imbalance shall be monitored at the frequency specified in the Surveillance Frequency Control Program when axial power imbalance alarm is OPERABLE, and every 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months when imbalance alarm is inoperable during power operation above 40 percent of rated power.

3.5.2.8 A power map shall be taken at intervals not to exceed 31 effective full power days using the incore instrumentation detection system to verify the power distribution is within the limits shown in the CORE OPERATING LIMITS REPORT.

The axial power imbalance, quadrant power tilt, and contro l rod position limits are based on LOCA analyses which have defined the maximum linear heat rate.

These limits are developed in a manne r that ensures the initial condition LOCA maximum linear heat rate will not cause the maximum clad temperature to exceed 10 CFR 50 Appendix K. Operation outside of any one limit alone does not necessarily constitute a situation that would cause the Appen dix K Criteria to be exceeded should a LOCA occur. Each limit represents the boundary of operat ion that will preserve the Acceptance Criteria even if all three limits are at their maximum allowa ble values simultaneously. Additional conservatism included in the limit, development is introdu ced by application of:.

a. Nuclear uncertainty factors

b. Thermal calibration uncertainty

c. Fuel densification effects

d. Hot rod manufacturing tolerance factors e'

e.

f.

Postulated fuel rod bow effects Peaking limits based, on initial condition for Loss of Coola nt Flow transients.

The incore instrumentation system uncertainties used to develop the axial power imbalance and quadrant tilt limits accounted for various combinations of invalid Self Powered Neutron Detector (SPND)'signals. If the number of valid SPND signals falls below that used in the uncertainty analysis, then another system shall be used for monitoring axial power imbalance and/or quadrant tilt.

For axial power imbalance and quadrant power tilt measu rements using the incore detector system, the minimum incore detector system consists of OPERABLE detectors configured as follows:

Axial Power Imbalance

a. Three detectors in each of three strings shall lie in the same axial plane with one plane in each axial core half.

b. The axial planes in each core half shall be symmetrical about the core mid-planes.

c. The detectors shall not have radial symmetry.

Quadrant Power Tilt

a. Two sets of four detectors shall lie in each core half. Each set of four shall lie in the same axial plane. The two sets in the same core half may lie in the same axial plane.

b. Detectors in the same plane shall have quarter core radial symmetry.

3-35a Amendment No. 17, 29, aa, ae, 60,12 0, 12e, 142,1 50,15 7, 1e8, 211,2 73,27 4

CONTROWSO COPY A system of 52 incore flux detector assemblies \1,ith.seven detectors per assembly has been provided primarily for fuel managemen t purposes. The system includes data display and record functions and is also used for out-of-core nuclear instrumentation calibration and for core power distribution vcrifi cation. * *

a. The out-of-core instrumentation calibration includes:

I. Calibrations of the split detectors at initial reactor startup, during the power escalation program, and periodically thereafter.

2. A comparison check with the incore instrumentation in the event one of the four out-of-core power range detector assemblies gives abnormal readings during operation.

3. Confirmatio n that the out-of-core axial power splits arc as expected.

b. Core power distribution verification includes:

1. Measuremen t at low power initial reactor startup to check that power distribution is consistent with calculations.

2. Subsequent checks during operation to ensure that power distribution is consistent with calculations.

3. Indication of power distribution in the event that abnormal situations occur during reactor operatioi1.

c. The minimum requirement for 23 individual incore detectors is based on the follo"'ing:

1. An adequate axial imbalance indication can be obtained with nine individual detectors.

Figure 3.5-1 shows a typical set of three detector strings with three detectors per string that will indicate an axial imbalance. The three detector strings arc the center one, one from the inner ring of symmetrical strings and one from the outer ring of symmetrical strings. *

2. Figure 3.5-2 shows a typical detection scheme ,vhich will indicate the radial power distribution with 16 individual detectors. The readings from two detectors in a radial quadrant at either plane can be. compared with readings from the other quadrants to measure a radial flux tilt.

3. Figure 3 ..5-3 combines Figures 3.5-1 and 3.5-2 to illustrate a typical set of 23 individual detectors that can be specified as a minimum for axial imbalance determinatio n and radial tilt indication, as well as for the determination of gross core power distributions.

3-35b Amendment No. 150,157,21 1

The. 45 +/- 5% overlap betw~en successive control rod grpups is allowed lowerat the upper and lower part of the stroke. Control rods are arrangesince the worth of a rod is as follows: d iil groups or banks defined Group Function 1

2*

. 3*

~;{;~

S~feJy 4 Saf~ty 5 J=legtilating 6 RegliJatinij.

7 Re~uJa.tihg Control r9d groups are* withdrawn in seqQence. begiriniilg'with grc;,up overlapped 25 percent. The hor.mal position a:tpoweris for group 7tq. 1. Groups 5, 6 and 7 are be partially insertel:I.

The rod positior, limits are based cin the mosUirnitin:g J>.f the following peaking, Stiutdq~h.hiargiry, and pote.ntial.eject~d ror,l WQl'fh. As discusthr<<;je crite.ria: ECCS p9wer EGGS powerpeaki.ng qritElrio.n is.e.hsureJJ. ,bythe;tqq,p*osition limits; sed aoove; Compliance with the worth; CollSistent with)M:-rod position . liniit~; proyi~.es for*achieving .Theirniriirh~rn available:~od

any )im e; ,f;lssum irig t.he highest worth *control rod that is wit_hdra hC>t,shutr,lown by rea,Qtor trip at (Reference :1 ). ;The rqd position limits*a1s*o ens11r~ tliat .inserted rodwn remain s irHt,e full out:'.positlon.

wqrt~s greaterthgfl: O.f:!~°(o*delta k/k at.rated. pef.wer.. The.se values gmups: 'Nill not con.t~in single rod safety an.c:tlysis ofJhe nypoth~tic;:al :rciq ejectJ_ort ac;cide_nt (Ref~rence have,.beert~hoWn to be safe by the control,rod yjor:thc.oU .Oo/c,'~eltaJ<lk I~ i;tllpWed .by th~"roq posjtiori !im 2):

A maxin:,ufn single. in*serted in~ei1e.d co11trorrod worth\1.qo/~ delta:k/k at begi111iing qflife; hot, :zero its at-~ot *zero power.. A single transienfp~ak ~herm~I power and, therefore, less severe environ p9we~wd.uld result /n a lower .

k/k ejected rod worth a.trated :power. mental conseq4ences than 0.65% delta The*J1ant compute{wi(I ~can for tilt arid imbalailce and will satisfy the requirefnE:3nts. If.the computer is out of si;!rvice, then manual calcula technical.spe~ification power a.pd 'iriipalance abOVE! 40 percent.power must be performed as tion: for tilt above 15' perc~nt returned to service. specifi(:jd until the computer is 3-36 Amendmerit No. 17, 29, 39,* 40, 50, 126,14 2,150, 167,21 1, EGR;TM 04 01026, 273

fft-,

}

1

\

~;,.

Reduction of the nuclear overpower trip setpoint to 60% full power when thermal pms,,er is equal to or less than 50% full power maintains both core protection and an operability margin at reduced power similar to that at full power.

During the physics testing program, the high flux trip setpoints arc administratively set as follows to assure an additional safety margin is provided:

Test Power Test Setpoint 0 <5%

~80 90%

>80 105.1%

REFERENC ES (I) UFSAR, Section 3.2.2.1.2 - "R~ctivity Control Distribution" (2) UFSAR, Section 14.2.2.2 - IIRod Ejection Accident"

,.. 3-36a

/ Amendmen t No. 39, 126, 142, 150, 1, 211

INFORMATION ON THIS PAGE HAS BEE N DEL ETE D

3-36b Ame ndm ent No. 142, A52, 167, 168 278

3.5.3 ENGINEE~_ED SAFEGUARDS PROTECTION SYSTEM ACTUATION SETPOINTS Applfcability:

This. specification appli~s to the engineered safeguarps protection system actuation setpomts. . .

Objective:

To. provide for*a1;,1ton,atic initic!,t_ion. of the. engineered SqfeglJards pro~(?ction system in*the event of a breach of Reactor co*o1ant System integrity~

Specification:

3.5.3.1 Tt,~ engineered saJeg~ards protection system actuation setpoirits and permissible bypasses shall be as follows:

Initiating Signal Function Setooint High Reactor Building Reactor Building $pray s 3() psig Pressure ..(1)' *

Reactor Building Isolation *s 30psig High~F>ressure lrijE!ction s 4 psig Low-Pressure Injection s 4 psig Start Reactor Building C6pling **& Reactor Building Isolation s 4 psig Low Reactor Coolant High .Pressure lnj$ction ;;:: 11300(2) and System Pressure ;;:: 50Q(3) psig Low Pressure Injection ;;:: 1600(2) .qnd

500(3) *p*sig Reactor Building Isolation ;;

1600 *psig(2) 4.16 kv ES. Buses Undervoltage Relays Degraded Voltage Switch to Onsite Power Source and load shedding 3760 volts (4)

Degraded voltage timer 10 sec (5)

Loss of voltage Switch to Om~ite Pqwer Source and load shedding 2400 Volts (6) /

Loss of voltage timer 1.5 sec (7)

(1) May be bypassed for reactor building *1eak rate test.

(2) May be bypassed below 177_5 psig on decreasing pressure and is automatically reinstated before 1800 psig.* on increasing pressure.

(3) May be bypassed bE!low 925 psig_ on decreasing pressure and-Js automatically reinstated before exceeding. 950 psig on increasing pressure. * *

3-37 Amendment No. 70, 78, 78; 89, 14 Q, 159 l

CONTROlJ ID OOPY (4) Minimum allowed setting is 3740 v. Maximum allow ed setting is 3773 V.

(5) Minimum allowed time is 8 sec. maximum allowed time is 12 sec.

(6) Minimum allowed setting is 2200 volts, maximum allow ed setting is 2860 volts (7) Minimum allowed time is I .0 second, maximum allow ed time is 2.0 seconds.

High Reactor Building Pressure The basis for the 30 psig and 4 psig setpoints for the high pressure signal is to establish a setting which would be reached in adequate time in the event of a LOCA, cover a spectrum of break sizes and yet be far enough above normal operation maximum internal pressures to prevent spurious initiation (Reference I).

Low Reactor Coolant System Pressure The basis for the 1600 and 500 psig low reactor coolant pressu re setpoint lor high and low pressure injection initiation is to establish a value which is high enough such that protection is provided for the entire spectrum of break sizes and is far enough below normal operating pressure to prevent spurious initiation. Bypass of HPI below 1775 psig and LPI below 925 psig, prevents ECCS actuation during normal system cooldown (Reference s I and 2).

4.16 KV ES Bus Undervoltage Relays The basis for the degraded grid voltage relay setpoint is to protec t the safety .

related electrical equipment from loss of function in the event of a sustained degraded voltage condition on the offsite power system.

The timer setting prevents spurious transfer to the onsite source for transient conditions.

The loss of voltage relay and timers detect loss of offsite power condition and initiate transf er to the onsite source with minimal time delay.

The minimum and maximum degraded voltage setpo int are "as found" readings.

References (I) UFSAR, Table 7.1-3 (2) UFSAR, Section 14.1.2.10 - "Steam Generator Tube Failure"

if"

~

3-37a

~

/

. Amen dmen t No. 7Q, 73, 78, 89, 149, 157, 1.§9, 224

COf\ITROl I ED COPY 3.5.4 INCORE INSTRUMENTATION This page intentionally Left Blank (Page 3-39 deleted)

. 3-38 Amendment No. HO, 211

= ' .

LACK RADIAL SY~METRY TOP AXIAL CORE HAl,.F

- ....._*.~

AXIAL PLANE

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BOTTOM AXIAL CORE HALF INCORE INSTR.UMENT,ATION . SPE<:IFICATION AXIAL IMBALANCE:INDICATION .

THREE MILEJSLAND' NUCLEAR' STATIO'N UNIT 1 FIGURE 3.5-1 3-398 Amendment No. .:t.tff, 269

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11111111, ,;RADIAL"SYMMETRY IN THI' PLANE INCORE INSTRUMENTATION SPECIFICATION

. . RADIAL FLUX'iNDICATION *.

THREE MILE ISLAND NUCLEAR STATION UNIT 1 FIGURE 3.5-2 3-3~b .

Ameridmerit N.o. ~, 269

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INCORE INSTRU MENTA TION SPECIFICATION THREE MILE ISLAND NUCLE AR STATION UNIT 1 FIGURE 3.5-3 3-39c Amendment No. ti1, 269

CONTROi ! ED COPY

(("7) f<.-:"'\p

':, ;i.-;!,

\\i This page intentionaily Left B1ank 3-40 Amendment H+-, 211

CONTROi i ED COPY

~CIDENT MONI1DRING INSTRU£NTATI0N Applicabill ty.

Applies to the operabi lity requirements for the instrum ents-ide ntified in Table 3.S-2 and TBD!s 3.5-.3 ouring STARTUP, POWER OPERATION and HOT STANDBY.

Objectives To assure operabi lity of key instrumentation useful in diagnosing

situatio ns which could represen t or lead to inadequate core cooling or evaluate and predict the course of accidents beyooo the desir,i basis.

Specific ation 3.5.5.l The minimum nU1JDer of channels identifi ed for the instruments in Table 3.5-2, shall be OPERABLE. With the nuni:Jer of instrumentation channels less than the minimum required, restore the inoperable channel(s) to OPERABLE status within seven (7) days (48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months fer pressuri zer level) or be in at least HOT SHUlDOWN within the next six (6) hours ahd in COLD SHUlDOWN within an addition al .30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Prior t!I start&.p following a COLD SHUTDOWN, the mini.mun nUl1tler of channels shJwn in Table 3.5-2 shall be operable.

/Iqi~~?i~);..

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3.5.5.2 The channels identifi ed for the instruments specifie d in Table 3.5-3 shall be OPERABLE. With the nunt:Jer of instrumentation channels less than required , restore the inq:>erable channel(s) to CFERABLE in accordan::e with the action specifie d in Taole 3.5-3.

Bases The saturati on Margin M:>nitcr provides a*quick and reliable means for determination of saturati on temperature margins. Hand calculat ion of saturati on pressure and saturati on temperature r.iargins can be easily and quiekly performed as an alternat e indicati on for the Saturati on Margin t-tinitors. *

Discharge now fl'l?Tii the two (2) pressuri zer code safety valves and the PORV is measured by differen tial pressure ttansmi tters connected across elbow taps downstream cf each valve. A delta-pr essure indicati on from each pressure transmitteT is availaol e in the control room to indicate code safety or relief valve line flow. An alarm is also provided in the control room to indicate that discharge from a pressuri zer code safety or relief valve is occurrirg. In addition ,

an acoustic monitor is provided to detect flow in the PORV discharge line. An alarm is provided in the control room for th-e acoustic mcnitar.

3-40a

~)

Amendment No. 7~, 100

CONTROLLED COPY 3.5.5 ACCIDEN T MONITORING INSTRUMENTATION (Continued)

The Emergenc y Feedwater System (EFW) is provided with two channels of flow instrument ation on each of the two discharge lines. Local flow indication is also available for the EFW System.

Although the pressurizer has multiple level indications, the separate indications are selectable via a switch for display on a single display. Pressurizer level, however, can also be determined via the patch panel and the computer log. In addition, a second channel of pressurizer level indication is available independent of the NNI.

Although the instruments identified in Table 3.5-2 are significant in diagnosing situations which could lead to inadequate core cooling, loss of any one of the instrument s in Table 3.5-2 would not prevent continued; safe, reactor operation. Therefore, operation is justified for up to 7 days (48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for pressurizer level). Altemate indications are 0

available for Saturation Margin Monitors using hand calculations, the PORV/Safe ty Valve position monitors using discharge line thermocouple and Reactor Coolant Drain Tank indications , and for EFW flow using Ste~m Generator level and EFW Pump discharge pressure .. Pressurizer level has two channels, one channel from NNI (2 DIP instrument strings through a single indicator) and one channel independent of the NNI. Operation with the above pressurizer level channels out of service is permitted for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Alternate indication would _be available through the plant computer.

r:_he operability of design basis accident monitoring instrumentation as identified in Table 3.5-3, ensures that sufficient information is available on selected plant parameters to monitor and assess the variables following an accident. (This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an

. Accident," Rev. 3, May 1983.) These instruments will be maintained for that purpose.

3-40b Amendme nt No. 78,100, 144,161,2 40,242

)>

3 TABLE 3.5-2 CD

J

a. ACCID ENT MONITORING INSTRUMENTS 3

CD

J zp FUNCTION INSTR UMEN TS NUMBER OF CHANNELS . MINIMUM NUMBER OF CHAN NELS I

N

~

N 1

2 Saturation Margin Monitor Safety Valve Differential Pressure Monito r 2

1 per discharge line 1

1 per discharge line 3 PORV Position Monito r 2 1*

4 Emergency Feedw ater Flow 2 per OTSG 1 perOT SG w

.h 5 *Pressurizer Level 2 0

1 0

6 Backup lncore Thermocouple 4 thermocouples/core 2 thermocouples/core quadrant Display Channel quadrant

With the PORV Block Valve closed in accordance with Specifi cation 3.1.12.4.a, the minimum numbe r of channels is zero.*

POST ACCIDENT MONITORING INSTRUMENTATION REQUIRED NUMBER MINIMUM NUMBER FUNCTION INSTRUMENTS OF CHANNELS OF CHANNELS

)> ACTION 3*

CD

J

1. High Range Noble Gas Effluent ;

0. a. Condenser Vacuum Pump. Exhaust (RM-~5-Hi) 1 1 3 A

b. Condenser Vacuum Pump Exhaust (RM-G25) 1 CD 1

J A C. Auxiliary and FuelHan dling 1 z 1 A Building Exhaust (RM-AB-Hi) .

9

d. Reactor Building Purge Exhaust (RM-A9-Hi)

e. Reactor Building Purge.Ex haust (RM-G24) 1 1 A 0

f. Main Steam Lines Radiation 1 1 A 0 (RM-G26/RM-G27) 1 each OTSG 1

each OTSG A

z

2. Containment High Range Radiation 2 2 A =I 3.

(RM-G22/G-23)

Containment Pressure

,a N

.j::o, 4. Containment Water Level 2 1 8 0 Ow

~

a. Containment Flood (LT-806/807) 2 1 B r

0 b. Coritainment Sump (LT-804/805) 1 0 F C.

5. DELETED

C m

6.

7.

Wide R~nge_ Neutron Flux Reactor Coolant System Cold Leg Water 2 1 A 'c Temperature (TE-959, 961; Tl-959A, 961 A) 2 1* A n

8.

9.

Reactor Coolant System Hot Leg Water Temperature (TE-958, 960; Tl-958A, 960A)

Reactor Coolant System Pressure (PT-949, 963; Pl-949A, 963) 2 2

1 1

A A

,,0

~

10. Steam Generator Pressure (PT-950, 951, 1180, 2/0TSG 1/0TSG 1184; Pl-950A, 951 A 1180, 1184) A

11. Condensate Storage Tank Water Level (LT-1060, 2/Tank 1/Tank 1061, 1062, 1063; Ll-1060, 1061, 1062, 1063) A

CONTROLUED COPY

\t:::':;'::_) TABLE 3.5-3 (Cont*inued)

J --

J, ACTIONS A. With the*number of OPERABLE channels less than required by the Minimum Channels OPERABLE requfrements:

1. either restore the inoperable channel(s) to OPERABLE status within 7 days of the event, or

2. prepare and submit a Special Report within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

B. 1. With the number of OPERABLE accident monitoring instrumen-tation channels less than the Required Channels OPERABLE requirements, restore the inoperable channel(s) to OPERABLE*

status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .*

2. With the number of OPERABLE accident monitoring instrumen-tation channels less than the Minimum Channels OPERABLE requirements, restore the inoperable channel.Cs) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

C. Restore the inoperable sump level instrument to OPERABLE status prior to startup following the COLD SHUTDOWN subsequent to its inoperab1lity declaration.

3-40e Amendment No . .l.80',166

W: DC OP Y 3.5.6 DELETED 3-40f Amendment No. J1J, JJ7, 182

CONTROUID COPY 3.5.7 REMOTE SHUTDOWN SYSTEM Applicability Applies to the operability requirements for the Remote Shutdown System Panel "B" Functions in Table 3.5-4 during STARTUP, POWER OPERATION AND HOT STANDBY.

Objectives To assure operability of the instrumentation and controls necessary to place and maintain the unit in HOT SHUTDOWN from a location other than the control room.

Specification The minimum number of functions identified in Table 3.5-4 shall be OPERABLE. With the number of functions less than the minimum required, restore the required function to OPERABLE status within 30 days or be in at least HOT STAND BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within an additional 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The Remote Shutdown System provides the control room operator with sufficient instrumentation and controls to place and maintain the unit in a safe shutdown condition from locations other than the control room. This capability is necessary to protect against the possibility that the control room becomes inaccessible. A safe shutdown condition is defined as HOT SHUTDOWN.

In the event that the control room becomes inaccessible, the operators can establish control at the remote shutdown panel and place and maintain the unit in HOT SHUTDOWN. Not all controls and necessary transfer switches are located at the remote shutdown panel. Some controls and transfer switches will have to be operated locally at the switchgear, motor control panels, or other local stations. The unit automatically reaches HOT SHUTDOWN following a unit shutdown and can be maintained safely in HOT SHUTDOWN for an exiended period of time.

3-40g Amendment No. 216

COf\ITROWED COPY The OPERABILITY of the Remote Shutdown System control and instrumentation Functions ensures that there is sufficient infonnation available on selected unit parameters to place and maintain the unit in HOT SHUTDOWN should the control room become inaccessible.

The Remote Shutdown System is required to provide equipment at appropriate locations _

outside the control room with a capability to promptly shut down and maintain the unit in a safe condition in HOT SHUTDOWN.

The criteria governing the design and the specific system requirements of the Remote Shutdown System are located iri IO CFR 50, Appendix A, GDC 19.

The controls, instrumentation, and transfer switches are those required for: Reactor Coolant Inventory Control, Reactor Coolant System Pressure and Temperature Control, Decay Heat Removal, Reactivity Monitoring, OTSG Level and Pressure Control, Reactor Coolant Flow Control, and Electrical Power.

The Remote Shutdown System instruments and control circuits covered by this specification do not need to be energized to be considered OPERABLE. This specification is intended to ensure the Remote Shutdown System instruments and control circuits will be OPERABLE if unit conditions require that the Remote Shutdown System be placed in operation. The operability of components and equipment are detem1ined by their respective Technical Specification requirements. If a component required for safe shutdown is placed in its fail-safe condition, as permitted by Technical Specifications, then the safety function has been assured and the remote shutdown panel function is considered operable.

Entry into an applicable REACTOR OPERATING CONDITION while relying on the specification actions is allowed even though the specification actions may eventually require a unit shutdown. This is acceptable due to the low probability of an event requiring these instruments.

The conditions of the specificatio_n may be entered independently for each Function listed on Table 3.5-4 and completion times of inoperable Functions v.ill be tracked separately for each Function.

3-40h Amendment No. 216

CONTROL! ED COPY TABLE 3.5-4 (Sheet I of2)

REMOTE SHUTDOWN SYSTEv! INSTRUMEN TATION AND CONTROLS Function/Instrument Required Number or Control Parameter of Functions I. Reactor Coolant Coolant Temperature Inlet Temperature Coolant Pressure Pressurizer Level RC-V-2 RC-V-3

2. Emergency Feedwat~r Controls EFW A Flow Indicator EFW B Flow Indicator OTSGA Level OTSGB Level EF-V-30B EF-V-300

.,.... . OTSG "B" Pressure Control Outlet Pressure MS-V-4B MS-V-8B MS-V-8A 3-40i Amendment No. 216

CONTROU .ED COPY TABLE 3.5-4 (Sheet 2 of2)

Function/Instrument Required Number or Control Parameter of Functions

4. Decay Heat Removal Cooler Outlet Temperatur e Pump Inlet Temperatur e Flow

5. Reactor Neutron Power Source Range Flux

6. Makeup Control and Status MU-P-IB MU-P-IC MU-P-3B c~~*~,

MU-P-3C MU-V.,2A MU-V-2B MU-V-8 MU-V-148 MU-V-16C MU-V-16D MU-V-18 MU-V-20 MU-V-32 Indicator MU-V-37 DH-T-1 BWST Level Makeup Tank Level

7. Decay Heat Closed Cycle Cooling Water DC-P-1 B (Auxiliary "B" Panel)

8. Diesel Generator EG-Y-IB 3-40j Amendment No. 216

3.6 REACTOR BUILDING Applicability Applies to the CONTAINMENT INTEGRITY of the reactor buildin g as specified below.

Objective To assure CONTAINMENT INTEGRITY.

Specification 3.6.1 Except as provided in Specifications 3.6.6, 3.6.8, and 3.6.12, CONT AINMENT INTEGRITY (Section 1.7) shall be maintained whenever all three of the following conditions exist:

a. Reactor coolant pressure is 300 psig or greater.

b. Reactor coolant temperature is 200 degrees F or greater. *

c. Nuclear fuel is in the core.

3.6.2 Except as provided in Specifications 3.6.6, 3.6.8, and 3.6.12, CONT AINMENT INTEGRITY shall be maintained when both the reactor coolant system is open to the containment atmosphere and a shutdown margin exists that is less than that for a refueling shutdown.

3.6.3 Positive reactivity insertions which would result in a reduction in shutdown margin to less than 1% delta k/k shall not be made by control rod motion or boron dilution unless CONTAINMENT INTEGRITY is being maintained.

3.6.4 The reactor shall not be critical when the reactor building interna l pressure exceeds 2.0 psig or 1.0 psi vacuum. .

3.6.5 Prior to criticality following refueling shutdown, a check shall be made to confirm that all manual Containment Isolation Valves (CIVs) which should be closed are closed and are conspicuously marked.

3.6.6 When CONTAINMENT INTEGRITY is required, if a CIV (other than a purge valve) is determined to be inoperable:

a. For lines isolable by two or more CIVs, the CIV(s)* required to isolate the penetration shall be verified to be OPERABLE. If the inoperable valve is not restored within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , at least one CIV* in the line will be closed or the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to the COLD SHUTDOWN condition within an additional 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. For lines isolable by one CIV, where the other barrier is a closed system, the line shall be isolated by at least one closed and de-activated automatic valve, closed .

manual valve, or blind flange within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to the COLD SHUT DOWN condition within an additional 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

e.

Ai'I CIVs required to isolate the penetration.

3-41 Amen dmen t No. 87, +Ga, 4-08, .:t-QQ, +98, 240, 246 278

CONTROLLED C PY 3.6 REACTOR BUILDING {Continued) 3.6. 7 DELETED 3.6.8 While CONTAINMENT INTEGRITY is required (see Specification 3.6.1), if a 48" reactor building purge valve is found to be inoperable perform either 3.6.8.1 or 3.6.8.2 below.

3.6.8.1 If inoperability is due to reasons other than excessive combined lea~age, close the associated valve and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verify that the associated valve is OPERABLE.

M~intain the associated valve closed until the faulty valve can be declared OPERABLE. If neither purge valve in the penetration can be declar ed OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.6.8.2 If inoperq.bility is due to excessive combined leakag~ (see Specif ication 6.8.5), within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months restore the leaking valve to OPERABILITY or perform either a orb below.

a. Manually close both associated reactor building isolation valves and meet the leakage criteria of Specification 6.8.5 and perform either (1) or (2) below:

(1) Restore the leaking valve to OPERABILITY within the following,72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

\

(2) Maintain both valves .closed by administr~tive controls, verify both valves are closed at leas.t once per 31 days and perform the interspace pressu rization test in accordance with the Reactor Building Leakage Rate Testin g .

Program. In order to accomplish repairs, one .containment purge valve may be opened for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following successful completion of an interspace pressurization test.

b. Se in HOT SHUT DOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUT DOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.6.9 Except as specified in 3.6.11 below, the Reactor Building purge isolation valves (AH-V-1 A&D) shall be limited to less than 31 degrees and (AH-V

-1 S&C) shall be limited to less than 33 degrees open, by positive means, while purgin g is conducted.

3.6.10 During STARTUP, HOT STANDBY and POWER OPERATION

a. Containment purging shall not be performed for temperature or humidity control.

b.

Containment purging is permitted to reduce airborne activit y in order to facilitate containment entry for the following reasons.

(1) Non-routine safety-related corrective maintenance.

(2) Non-routine safety-related surveillance.

3-41a Amendment No. 87, 4GB, .:t-e-7, +98, 2{}:t., 246

CO NT RO llB ) COPY

3. 6 REAC TOR BUIL DING (Continued)

(3) Perfo nnanc e of Technical Specification required surveillance s.

(4) Radiation Surveys.

(5) Engin eering support of safety-related modifications for pre-outage planning.

(6) Purgin g prior to shutdown to prevent delaying of outag e commencement (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to shutdown).

c. Containment purgin g is permitted for React or Building pressu re control.

d. To the extent practicable the above containment entries shall be scheduled to coincide, in order to minimize instances of purging.

3.6.11 When the reacto r is in COLD .SHU IDOW N or REFU ELIN G SHUT DOW N, continuous purging is permitted with the React or Building purge isolation valves opened fully.

3. 6.12 Person11el or emergency air locks:

a. At least one door in each of the personnel or emergency air locks shall be closed and sealed during personnel passage through these air locks.

b. One door of the personnel or emergency air lock may be open for maintenance, repair or modification provided the other door of the air lock is verifie d closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , locked within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and verified to be locked closed monthly.

Air lock doors in high radiation areas may be verified locked closed by administrative means.

c. Entry and exit is permissible to perform repairs on the affect ed personnel or emerg ency air lock components. With both air locks inoper able due to inoperability of only one door in ~ch airlock, entry and exit days under administrative contro.s. With the personnel or emerg is permissible for 7 ency air lock door interlock mechanism inoperable, entry and exit is penni ssible under the control of a dedicated individual.

d. With one or more air locks inoperable for reasons_ other than "b" or "c" above, initiate action immediately to evaluate the overall containment leakage rate with respect to the requirements of Specification 6.8.5, verify a door is closed in the

affected air lock within I hour, and restore the affected air lock(s) to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the reactor shall be brought to HOT SHUT DOW N within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUT DOW N within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3-4lb Amen dmen t 87, 108, 167, 198 201

CONTROLLED COP Y

}_.6 __ REACT OR BUILDING (Continued)

{'/ "\

},,;,.,\., /\BASE S

. 11 1 .*,........ -.. __/,/

i The Reactor Coolant System conditions of COLD SHUTDOWN assure that no steam will be formed and hence no pressure will build up in the containment if the Reactor Coolant System ruptures. The selected shutdown condjtions are based on*the type of activities that are being carried out and wili preclude criticality in any occurrence.

A condition requiring integrity of containment exists whenever the Reactor* Coolant System is open to the atmosph ere and there is insuffic_ient soluble poison in the reactor coolant to maintain the core one percent subcritical in the event all -control rods are withdraw

n. The R~actor Building is designed for an internal pressure of 55 psig, and.an external pressure 2.5 psi greater than the internal pressure. *

The primary Containment Isolation Valves (CIVs) are identified in UFSAH Table 5.3-2.

Additional Vent, drain, test and other manually operated valves which complete the containm ent bounda ry are identified in the containm~nt integrity checklist. For the purpose of this specification, check valves and relief valves identified in the containment integrity checklist are defined to be active valves.

The loss of redundant capability for containment isolation is limited for all penetra tions after which the containm ent penetration must be isolated. Isolation of certain penetra tions may require the closure of multiple CIVs due to piping branches.

1. When one of two CIVs in a line is inoperable, the capability to isolate the penetra tion using the other CIV in the line is promptly verified and at least ohe valve in the line must be closed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or the plant rnust commence shut down. 1 i .

2. For those CIVs where the second barrier is a closed system within the Reactor Building there is no other CIV to isolate the penetration. If operability cannot be regained

, the valve must be closed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant must commence shut down. An action time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable considering the relative stability of the closed system (hence, reliability} to act as a containment isolation bounda ry and the relative importance of suppb~!ng containment integrity.

The definition of Containment Integrity permits normally closed CIVs, except for the 48 inch purge valves, to be unisolated intermittently or manual control to be substituted for automa tic control under administrative control. Administrative control includes the following considerations: (1} stationing an operator, who is in constant communication with the control room, at the valve controls,. (2} instructing this operator to close these valves in an accident situation,* and (3) assuring that environmental conditions will not preclude access to close the valves and that this actioi') will prevent the release of radioactivity outside the containm ent (Reference 1). The dedicated individual can be responsible for closing more than one valve provided that the valves are in close vicinity and can be closed in a timely manner

. Due to the size of the containment purge line penetration and the fact that those penetrations exhaust directly from the containment atmosphere to the environment, the containment penetrations containing these valves may not be opened under administrative control.

An analysis of the impact of purging on ECCS performance and an evaluation of the radiological consequences of a design basis a_ccident while purging have been completed and accepte d by the NRG staff. Analysis has demonstrated that a purge isolation valve is capable 3-41c Amendm ent No. 87, 400, 4-e7, 498, 20+, 24-0, 246

CONTROLLED COPY 3.6 REACTO R BUILDING {Continued)

(s~~~d} . . . .

\ '

~"fclos ing-a~ the dynamic forces associated with a LOCA when the valve is limited to a

nominal 30 degree open position.

,I Allowing purge operations during STARTUP, HOT STANDBY and POWER OPERATION (T.S. 3.6.1 O) is more beneficial than requiring a cooldown to COLD SHUTDOWN from the standpoint of (a) avoiding unnecessary thermal stress cycles on the reactor coolant system and its components and (b) reducing the potential for causing unnecessary challeng*es to the reactor trip and ~afeguards systems.

The hydrogen mixing is provided by the reactor building ventilation system to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.

Maintaining containment air locks OPERABLE requires compliance with the leakage rate test requirements of 1O CFA 50, Appendix J (Reference 2), and the Reactor Building Leakage Rate Testing Program. Each air lock door has been designed and is tested to certify its ability to withstand a pressure in excess of the maximum expected pressure following a Design Basis Accident (OBA) in containment. Closure of a single door in each air lock is sufficient to provide a leak tight barrier following postulated events. \

Entry and exit is allowed to perform repairs on the affected air lock component.* If the outer door is inoperable, then it may be easily accessed to repair. If the inner door is the one that is inoperable, however, then a short time exists when the containment boundary is not inta~t (during access through outer door). The ability to open the OPERABLE door, even if it means the containment boundary is temporarily not intact, is acceptable due to the low probabiHty of an event that could pressurize the containment during the short time in which the OPERABLE door is expected to be open. After each entry and exit the OPERABLE door must be immediately closed. If ALARA conditions permit, entry 1:;1.nd exit should be via an OPERABLE air lock. With both air locks inoperable due to inoperability of one door in each of the two air locks, entry and exit is allowed for use of the air locks for 7 days under administrative controls. Containment

entry may be required to perform Technical Specifications (TS) Surveillance and Required Actions, as well as other activities on equipment inside containment that are required by TS or activities on equipment that support TS-required equipment. This is not intended to preclude performing other activities (i.e., non-TS-required activities) if the.containment was entered, using the inoperable air lock, to perform an allowed activity listed above. This allowance is acceptable due to the low probability of an event that could pressurize the containment during the short time that the OPERABLE door is expected to be open.

With one or more air locks inoperable for reasons other than those described in 3.6.12 "b" or "c," Section 3.6.12.d requires action to be immediately initiated to evaluate previous combined leakage rates using current air lock test results. An evaluation is acceptable since it is overly conservative to immediately declare the containment inoperable if both doors in an air lock have failed a seal test or the overall air lock leakage is not within limits. In many instances (e.g., only one seal per door has failed), containment remains OPERABLE, yet only 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months would otherwise be provided to restore the air lock to OPERABLE status prior to requiring a plant shutdown. In addition, even with both doors failing the seal test, the overall containment leakage rate can still be within limits.

3-41d Amendm ent No. 498, 20+, 246

CONTROLLED COPY 3.6 REACTOR BUILDING (Continued)

~)con tinued )

"----~*

Section 3.p.12.d requires that one door in the affected containment air locks(s) must be verified to be closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Additionally, the affected air lock(s) must be restored to OPERABLE status within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is considered reasonable for restoring an inoperable air lock to OPERABLE status assuming that at least one door is maintained closed in each affected air lock.

R~s

______../ I (1) NRG Generic Letter 91-08 (2) 10 CFR 50, Appendix J.

3-41e Amendm ent No. 246

COPY 3.7 UNIT ELECTRIC POWER SYSTEM Applicability Applies to the availability of electrical power for operation of the unit auxiliaries.

Objective To define those conditions of electrical power availability necessary to ensure:

a. Safe unit operation

b. Continuous availability of engineered safeguards Specification

3. 7 .1 The reactor shall not be made critical unless all of the following requirements are satisfied:

a. All engineered safeguards buses, engineered safeguards switchgear, and engineered safeguards load shedding systems are operable.

b. One 7200 volt bus is energized.

c. Two 230 kY lines are in s.ervice.

d. One 230 kV bus is in service.

e. Engineered safeguards diesel generators are operable and at least 25,000 gallons of fuel oil are available in the *storage tank.

f. Station batteries are charged and in service. Two battery chargers per battery are in service.

3.7.2 The reactor shall not remain critical unless all of the following requirements are satisfied:

a. Offsite Sources:

(i.) Two 230 kV lines are in service to provide auxiliary power to Unit 1, except as specified in Specification 3. 7.2e below.

(ii.) The voltage on the 230 kV grid is* sufficient to power the safety related ES loads, except as specified in Specification 3.7.2.h below.

b. Both 230/4.16 kV unit auxiliary transformers shall be in operation except that within a period not to exceed eight hours in dur~tion from and after the time one Unit 1 auxiliary transformer is made or found inoperable, two diesel generators shall be operable, and one of the operable diesel generator will be started and run continuously until both unit auxiliary transformers are in operation. This mode of operation may continue for a period not exceeding 30 days.

3-42 Amendment No. 188,212, 224

c. Both diesel generators shall be operable except that from the date that one of the diesel. generators is made or found to be inoperable for any reason, reactor operation *is permissible.for the succ eeding seven days provided that the redundant diesel generator is:

1. verified to be operable immediately;

2. within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , either.:

a. determine the redundant diesel generator is not inoperable

due to a common mode failure; or,

b. -test redundant d_iesel generator in accordan ce with

surveillance requirement 4.6.*1.a. * .

In the.. event two diesel generators are inop erable, the u.ni.t s.hall .be placed in HOT SH_UTDOWN in 12.hours. If one dies el is not operable witnin an acfditionai 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period the pl.ant $hall be plcJ,ced in COLD *sHU TDO WN within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

~- * *

With one qiesel generator inqperabl~, in additi9n to the -abov~. verify that:

All required SY$tems, subsystems, trains, cortjponents and devices that depend on the remaining OPERABLE dies emergency power are also OPE_RABLE or a el gem~rator as source of

follow specifications.3.0. L

d. If one Unit Auxiliary Transformer is inoperab le and a diesel generator becomes inoperable, th~ unit will be placed in HOT SHUTDOWN within 12 h¢u rs.. If one of the *above sources of power is* not made opera.pie within an additional 24 .

h*oors the unit :shall be placed iri COLD SHU TDOWN withirJ an additidnai 24 hoLJrs thereafter.

e. If Unit J is separated from the system whil e carrying its QWr) auxiliaries, or if only _one 230 kV li'rie is in service, continue d reactor*operatiqn is permissible provided one emergency diesel generato r shaH be started and run

continuously untirtwo trarisniission liries are restored. *

f. The engineered safeguards electrical bus, s\J\litqhgear, load shedding, and automatic diesel start systems shall be oper able i9xcept as provided iri Specification 3.7.2c above and as required for testing.

g. One station battery may be removed from

service for not more than eight hours ..

h. If it is determined that a trip of the Unit 1 generator, Jn conjunction with LOC A loading, wi!I result in a loss of offsi te pow er to Engineered. Safeguards buses, the plant shall begin a power redu ction within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and be in HOT SHUTDOWN in an additional 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s, except as provided in Specification 3. 7.2.e above.

3-43 I.

Ame ndm ent No. 98,*188, 212 ,224 ,268 278 I

I

/

C COPY II

1

~\~ The _unit Electric Power System is designed to provide a reliable source of power for balance of plant auxiliaries and a continuously available. power supply for the engineered I

i safeguards equipment. The availability of the various components of the Unit Electric I Power System dictates the operating mode for the station.

Verification of emergency diesel generator and station battery operability normally consists of verifying that the surveillance is* current, and that other available information does not indicate inoperability.

It is recognized that while testing the redundant emergency diesel generator (EOG) in accordance with surveillance requirement 4.6.1.a, the EOG will not respond to an automatic initiation signal. In this situation, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time clock will not be entered per the provisions of section 3.7.2.f. due to the low probability of an event occurring while the EOG is being tested.

Trip of TMl-1 could result in a change in the 230 kV system (Grid) voltage at the TMI substation. The predicted voltage following a loss of the unit is referred to as the Post-Contingency voltage for trip of TMl-1. The transmission system operator monitors 230 kV system conditions for Post Contingency voltages. If the Post-Contingency voltage is less than the value required to support safety related ES loads, the transmission system

operator will notify the TMI Unit 1 control room. The required voltage setpoint values for dual or single auxiliary transformer operation are specified by degraded grid calculations.

The appropriate setpoint for the current plant condition(s) is provided to the Grid operator.

The required voltage setpoint is based on the Large Break LOCA loading which results in the greatest ES loads.

Upon receipt of a valid Post-Contingency voltage Alarm for Loss of TMl-1, TMI will implement the Low System (Grid) Voltage Procedure. An allowed.action time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides the transmission system operator time to take actions to reconfigure the 230 kV system for improved voltage support. The time allowed has been evaluated for the level of risk associated with the increased reliance on use of the onsite sources.

3-43a Amendment No. 4-88 , 224

GEN ERAL CONTROLLED COPY 3.8 FUEL LOADING AND REFUELING Applicability: Applies to fuel loading and refueling operations.

Objective:

To assure that fuel loading and refueling operations are performed in a responsible manner.*

Specification 3.8.1 DELETED 3.8.2 Core subcritical neutron flux shall be continuously monitored by at least two neutron flux monitors; each with continuous indication available, whenever core geometry is being changed. When core geometry is not being changed, at least one neutron flux monitor shall be in service:

3.8.3 At least one decay heat removal pump and cooler shall be operable.

3.8.4 During reactor vessel head removal and while loading and unloading fuel from the reactor, the boron concentration shall be maintained at not less than that required for refueling shutdown.

3.8.5 Direct communications between the control room and the refueling personnel in the Reactor Building shall exist whenever changes in core geometry are taking place.

3.8.6 During the handling of irradiated fuel in the Reactor Building at least one door in each of the personnel and emergency air locks shall be capable of being closed.* The equipment hatch cover shall be in place with a minimum of four bolts securing the cover to the sealing surfaces~

. --------- --------- --------- --NOTE --------- --------- --------- -

The equipment hatch may be open if all of the following conditions are met:

1) The Reactor Building Equipment Hatch Missile Shield Barrier is capable of being closed within 45 minutes,

2) A designated crew is available to close the Reactor Building Equipment Hatch Missile Shield Barrier, and *

3) Reactor Building Purge Exhaust System is in service.

3.8.7 During the handling of irradiated fuel in the Reactor Building, each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. Closed by an isolation valve, blind flange, manual valve, or equivalent, or capable of being closed,* or

2. Be capable of being closed by an operable automatic containment purge and exhaust isolation valve.

Administrative controls shall ensure that the Reactor Building Purge Exhaust System is in service, appropriate personnel are aware that air lock doors and/or other penetrations are open, a specific individual(s) is designated and available to close the air lock doors and other penetrations as part of a required evacuation of containment. Any obstruction(s) {e.g., cable and hoses) that could prevent closure of an air lock door or other penetration will be capable of being quickly removed.

3-44 Amendment No. 27,198,236 ,267, 260

3.8.8 GENERAL CONTR llED If-any of the above co*py specified limiting conditions for.fuel loading and refueling are not met, movement of fuel into the reactor core s~all cease; action shall be initiated to correct the conditions so that the specified limits are met, ahd no operations which may increase the

.~!""'"*\,... reactivity of the core shall be made. * *

. 3.8.9 The reactor building purgE! isolation valves, and associated radiation monitors which initiate purge (solation, shall be tested arid verified to be operable no more than 7 days prior to initial

fuel movement in the reactor building.

3.8.1 O Irradiated fuel shall not be. removed from the reactor until the uni~ has. been subcritical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3.8.11 During the handling c;,f irradiated fuel in the Reactor Building at least 23 feet of water shall be maintained above the _level. of the reactor pressure vessel flange~ as determined by a shiftly check and a daily verification. If the water level is less.than:23 feet above the reactor pressure vessel flange, place the fuel assembly(s) *being handled into a safe position, then cease fuel handling until the water level has been restored to 23 feet or greater above the reactor pressure vessel flange.

Detailep written procedures will be available for use by refueling personnel. These procedures, the above specifications, and the design of the fuel handling equipment as described .in Section 9. 7 of the UFSAR incorpora~ing built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety. If no charige Js being made in core geometry, one flux monitor is suffici11mt. This permits maintenance on the instrumentation. Continuous monitoring of neutron flux provides immediate indication of an unsafe condition. The decay heat removal pump is used to maintain a uniform boron concentration. The shutdown margin indicated in Specification 3.8.4 will keep the core subcritical, even with all control rods withdrawn from the core (Reference 1). The boron concentration will be sufficient to maintain the core.k 811 s 0.99 if all the control rods were removed from the core, however only a few control rods will be removed at any one time during fuel shuffling and replacement. The ke11 with all rods in the core and with refueling boron concentration is approximately 0.9. Specification 3.8:5 allows the control room operator to inform the reactor building personnel of any impending urisafe condition detected from the main control board indicators during fuel movement. * *

Per Specification 3.8.6 and 3.8.7, the personnel and emergency air lock doors, and penetraticms may be open during movement of irradiated fuel in the containment provided a minimum of one door in each of the air locks, and penetrations are capable of being closed in the event of a fuel handling accident, and the plant is in REFUELING SHUTDOWN orREFUE:LING OPERATION with.at least 23*feet of water above the fuel seated within the reactor pressure vessel, The minimum water level specified is the basis for the accident analysis assumption of a decontamination factor of 200 for the release to the containment atmosphere from the postuiated damaged fuel rods located on top of the fuel core seated in the reactor vessel. Should a fuel handling acq\dent occur inside containment, a minimum of one door in each personnel and emergency air lock, and the open penetrations will be closed following an evacuation of containment. Administrative controls will be in place to assure closure of at least one door in each air lock, as well as other open containment penetrations, following a containment evacuation.

Specification 3.8.6 is modified by a NOTE:

- *----* --------- * --------- ---- NOTE --- * ------- - * ------- ------- --

The equipment hatch may be open if all of the following conditions are met:

1) The Reactor Building Equipme~t Hatch Missile Shield Barrier is capable of being closed within 45 minutes,

2) A designated crew is available to close the Reactor Building Equipment Hatch Missile

,.-,--. Shield Barrier, and

'.i;@i 3) Reactor Building Purge* Exhaust System is in service.

3-45 Amendment No. 167, 178, 239, 246, 2SQ, 267,260

CONTROLLED COPY

These restrictions in9lude administrative controls to allow the opening of the reactor building equipment hatch during the handling of irradiated fuel in the Reactor Building provided that 1)

The Reactor Building Equipment Hatch Missile Shield Barrier is capable of being closed within 45 minutes, 2) A designated crew is available to close the Reactor Building Equipment Hatch Missile Shield Barrier, and 3) Reactor Building Purge Exhaust System is in service. The Reactor Building Equipment Hatch Missile Shield Barrier includes steel plating on the bottom-of the shield structure, which acts to restrict a release of post-accident fission products. The capability to close the reactor building missile shield barrier includes requirements that the barrier is capable of being closed and that any cables or hoses across the opening have quick disconnects to ensure the barrier is capable

of being closed within 45 minutes. The 45-minute closure time for the reactor building missile shield barrier starts when the control room communicates the need to shut the Reactor Building Equipmen t*

Hatch Missile Shield Barrier. This 45-minute requirement is significantly less thari the fuel handling accident analysis assumption that the reactor ouilding remains open to the outside environme nt for a two-hour period subsequent to the accident. Placing reactor building purge exhaust in service will ensure any release from the reactor building will be monitored, and ensure continued air flow into the Reactor Building in the event of a fuel handling accident. The Reactor Building purge valve high radiation interlock will be bypassed to ensure continued air flow into the Reactor Building in the event of a Fuel Handling Accident.

The administrative controls will also include the responsibility to be able to communicate with the control room, and the responsibility to ensure that the reactor building missile shield barrier is capable of being closed in the event of a fuel handling accident. These administrative controls will ensure reactor building closure would be established in the event of a fuel handling accident inside.

containment. * *

Provisions for equivalent isolation methods in Technical Specification 3.8.7 include use of a material (e.g. temporary sealant) that can provide a temporary, atmospheric pressure ventilation barrier for other containment penetrations during fuel movements.

Specification 3.8.9 requires testing of the reactor building purge isolation system. This system consists of the four reactor building* purge valves and the associated reactor building purge radiation monitor(s). The test verifies that the purge valves will automatically close when they receive initiation signals from the radiation detectors that monitor reactor building purge exhaust, and the valves remain open when the isolation system is bypassed. The test is performed no more than 7 days prior to the start of fuel movement in the reactor building to ensure that the monitors, purge valves, and associated interlocks are functioning prior to operations that could result in a fuel handling accident within the reactor building. The Fuel

Handling Accident analysis assumes that the four purge valves remain open.

Specification 3.8.1 O is required as the safety analysis for the fuel handling accident was based on the assumption that the reactor had been shutdown for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (Reference 2).

-REFERE NCES (1) UFSAR, Section 14.2.2.1 - "Fuel Handling Accident

(2) UFSAR, Section 14.2.2.1 (2) - "FHA Inside Containment

3-45a Amendme nt No. 236, 245, 25 7

3.9 DELETED 3.10 MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES Applicability Applies to byproduct, source, and special nuclear radioactive material sources.

Objective To assure that leakage from byproduct, source, and special nuclear radioactive material sources does not exceed allowable limits.

Specification 3.10.1.1 The source leakage test performed pursuant to Specification 4.13 shall be capable of detecting the presence of 0.005 µCi of radioactive material on the test sample. If the test reveals the presence of 0.005 µCi or more of removable contamination, it shall immediately be withdrawn from use, decontaminated, and repaired, or be disposed of in accordance with Commission regulations.

Sealed sources are exempt from such leak tests wheri the source contains 100 µCi or less of beta and/or gamma emitting material or 5 µCi or less of alpha emitting material.

3.10.1.2 A complete inventory of licensed radioactive materials in possession shall be maintained current at all times.

Bases The limitations on removable contamination for sources requiring leak testing, including alpha emitters, are based on 10 CFR 70.39(c) limits for plutonium.

This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.

3-46 (Pages 3-47 to 3-54 deleted)

Amendment No. e4, +29, 284 (3-31-81)

CONTROL! FD COPY 3.11 Handling of Irra diat ed Fuel

App lica bili ty Applies to the operation of the fuel handlin

. confines of U~it 1 and there is any g building crane when within the handling building. spent fuel in stor age in the Unit 1 fue1 Objective To define the lift conditions and allowable lift ed and *transported with the fuel handlin areas of trav el when loads to be 15 tons or between 1.5 tons and 15 tons or g building crane are in excess of con sist of irra dia ted fuel elements.

Spe cifi cati on 3.11.1 Spent fuel elements having less than 120 days irra dia ted fuel sha ll not be loaded into a for decay of the ir in the shipping cask area. spe nt fuel tran sfer cask 3.11.2 The key operated trav el inte rloc ~ system for the trav el area of the fuel handling building auto mat ical ly lim itin g whenevet:, loads in excess of 15 tons are to crane sha ll be imposed with the exception of fuel handling bridge be 1ifte d and tran spo rted maintenance.

3.11.3 The lowest surface of all loads in excess administratively limited to an elev of 15 tons sha ll be atio n one concrete surface at the nominal 348 ft-0 in. foot or less above the handling building. elev atio n in the fuel 3.11.4 Loads in excess of hook capacity sha ll not test ing . be lift ed, except for load 3.11.5 Following modifications or rep airs to any members, the crane sha ll be subjected of the load bearing of its rated load. to a tes t lif t of 125 percent 3.11.6

Administrative con trol s sha ll requ ire procedure with an ide ntif ied safe load the path use of an approved 3,000 lbs. handled above the Spent Fuel Pool for loads in excess of elev atio n). Operating Floor (348' 3.11.7 During tran sfer of the cask to and cask will be res tric ted to the tranfromsfe r the cask loading pit , the path Administrative con trol s wil l be usea to ensu shown in Figure 3. 11-1 .

movements of the cask are perfonned at slow re tha t all late ral speeds. During this tran sfe r the cask lift bridge and trol ley in the East-West dire ctio n *. ing yoke sha ll be oriented

, (t2;}" Amendments No

1-, JJ, 103 3-55

CONTROi J ED OOPY

o.'~ *~\ Bases 1.' -** )

©I( This sp ec ifi ca tio n will lim res ult ing from damage to speit ac tiv ity rel ea se s to un res tri cte d areas pools in the po stu lat ed evennt fuel sto red in the spent fuel sto rag e the fuel handling building cra t of the dropping of a heavy load from was performed assuming tha t ne. A Fuel Handling accident an aly sis fuel assemblies are su ffi cie the cask and its en tir e contents of ten cask, to ~llow the escape of ntl y damaged as a re su lt of dropping the (Reference 1) . This rel ea se all noble gases and iodine in the gap atmosphere and to occur instanwas assumed to be dir ec tly to the res ult ing from th is acciden taneously. The sit e boundary doses thy ro id, and are within the t are 5;25 R whole body and 1.02 R to lim its sp ec ifi ed in 10 CFR 10

0.

Sp ec ifi ca tio n 3.11.1 req uir es 120 days decay po st- irr ad iat th at spent fu el, having les s than tra ns fer cask in order. to enion , not be loaded in a spent fuel highly improbable spent fuel sure th at the doses res ult ing from a tal cu lat ed abo~e. tra ns fer cask drop would be within those Sp ec ifi ca tio n 3.1 1.2 req uir es which au tom ati ca lly lim its the key operated int erl oc k sys the tem, while it is lif tin g and tra ns tra ve l area of the fuel handling crane be imposed whenever loads in po rti ng the spent fuel shipping cask, to tra ns po rte d while the re is anyexcess of 15 tons are to be lif te d and fuel sto rag e pools in Unit 1. spent fuel in sto rag e in the spent heavy loads tra ve l in areas This au tom ati ca lly ensures th

(~":) drop ac cid en t, the re would be wh ere , in the un lik ely event of aat the se load

. in any damage to the spent fue no po ss ib ili ty of th is event res ult ing

.. structural*daniage to the sp l sto red in the pools, any unacceptable redundant tra in s of sa fet y rel ent fuel pool str uc tu re , or damage to are a is designed to w1thstan ate d components. The shipping cask cask from the 349 ft- 0 in. ele d the drop of the spent fuel shipping the spent fuel pool str uc tur va tio n without unacceptable damage to e {Reference 2) .

Sp ec ifi ca tio n 3.1 1.3 ensures load never ge ts higher tha tha t the lowest surface of any the 348 ft- 0 in. ele va tio n n one foo t above the concrete surfaheavy in the fuel handling building ce of ele va tio n 349 ft- 0 in. ). thereb {no mi nal un lik ely load drop ac cid en t y keeping any impact force from an within acceptable lim its .

Sp ec ifi ca tio n 3.1 1.4 en res used for lif tin g and trasu tha t the proper capacity cra of a load drop ac cid en t. ns po rti ng loads thus reducing thene hook is

pr ob ab ili ty Following modification load rat ing of the craneor. rep air s, sp ec ifi ca tio n 3.11.5 confirms the (l(~ere~~,~'S

'-1:------- *-**-***-****.

,r'"~** ... . . , .\ (1) UFSAR, Section 14 .2. 2.1 11

- Fuel Handling Accident" (2) UFSAR, Section 14.2 .. 2.8 11

~* . Fuel Cask Drop Accident" 3-56 Amendment No. 2i, 48, i09, 157

3p ec ifi ca tio n 3.1 1.6 imposes weighing in excess of 300 ad mi nis tra tiv e lim its on handli 0 lbs ng loads loads~ if dropped, to impact . to minimize the po ten tia l for heavy or to impact redundant saf e shu irr ad iat ed fuel in the spent fuel pool, sh all follow, to the ex ten t pr tdown equipment. The saf e load path beams, etc ., such tha t if the ac tic al, str uc tur al flo or members, more lik ely to withstand the load is dropped, the str uc tur e is 3000 lbs . without these res impact. Handling loads of les s than consequences of dropping loatri ds cti on s is acceptable because the those produced by the fuel han in thi s weight range are comparable to and found acc ep tab le. dling accident considered in the FSAR Sp ec ifi ca tio n 3.11.7 in combin fuel cask is handled in a manneation with 3.11.3 ensures the ~pent an aly sis (Reference 3). r co ns is; en t with the load dro p

Reference (3) GPU Evaluation of Heavy Loa February 21, 1984, as tra ns mi d Handling Operations at TMl-1 No. 5211-84-2013. tte d to the NRC in GPUN Le tte r

3-S 6a Amendment Nos. l<<, <<B, !CS, 157

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l'Ja Ua Z J.l l-1 Amendment No. 109 3-56b

CONTROU ED COPY 3 .12

REACTOR BUILDING POL.AR CEA.NE ApPl icabi l.i ty Appli es to the use ot the react or build ing polar crane gene rator comp artme nts and the tu.el trans fer cana l. hoist s over the steam Obje ctive To iden tify those cond itions tor vhich the opera tion of -the react or build ing polar crane hoist s are restr icted .

S'Dec ificat ion 3.12. 1 The react or build ing polar crane hoist s ~hal i not be opera ted over the fuel trans fer canal vhen e;ny fuel assem bly is being moved.

3.12. 2 .Duri ng the perio d vhen the react or vesse l head is removed and irrad iated fuel is in the react or build ing and fuel is not being moved, the react or build ing polar crane hoist shall be opera ted over the fuel trans fer canal only vhere neces sarr, and in accor dance with appro ved opera ting proce dures stati ng the purpo se of such use.

3.12. 3 Durin g the perio d when the react or coola nt syste m is press urize d above 300 psig, and is above 200 F, and fuel is in the react o~ build ing polar crane hoist s shall not be opera core, the the steam gene rator comp artme nts. ted over Bases Rest rictio n of use of the react or build ing polar crane hoist s over the fuel trans fer canal vhen the react or vesse l head is remov ed to perm it those opera tions neces sary for the fuel hand ling and core inter nals opera tions is to precl ude the dropp ing of mate rials or equip ment into the react or vesse l and possi bly damaging the fuel to the exten t that any escap e of resu lt. fissio n produ cts would

.Rest rictio n of use of the react or build ing polar crane hoist s over the steam gene rator comp artme nts durin g the time when steam could be formed from dropp ing a load on the steam gener ator or react or coola nt pipin g resul ting in ruptu re of the syste m is requi red to prote ct again st a loss-of-co olant accid ent.

Amendment No. 115 3-57

3.13 . SECONDARY COOLANT SYSTEM ACTIVITY Applfcab111t,¥ Applies to the 11m1t1ng conditions pressure 1.s greater than 300 psi g orforTavg operation when reactor coolant system f s grea ter than 200° F.

Objective To lim it the inventory of acti vity in the seco ndary system.

Speciffcatfon 3.13 .l The spe cifi c acti vity of the secondar

.! 0.10 ~ Cf/gram DOSE EQUIVALENT I-13y1. coolant system shal l be 3.13 .2 With the spe cifi c acti vity of the Cf/gram DOSE EQUIVALENT I-131, be secoin ndary coolant system> O~lOJJ at hours and fn* COLD SHUTDOWN within the folleas t HOT STANDBY within 6 lowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Bases The lim itat ions on secondary system spe resu ltan t off- site radiatio cifi c acti vity ensure that the n dose 10 CFR Par t 100 lim its in the evenwill be limited to a small *fra ctio n of t of a steam line rupture. This dose includes the effe cts of a coincident 1.0 GPM prim the steam generator of the affected steam ary-to-secondary tube leak in line

Amendment No. 115 3-58

3.14 FLOO D 3.14.1 PERI ODIC INSP ECTI ON OF THE DIKES ARO UND TMI Applicability Applies to inspection of the dikes surrounding the site.

Objec tive To specify the minimum frequency for inspection of the dikes and to define the flciod stage after which the dikes will be inspected.

Specification 3.14.1.1 *The dikes shall be ,inspected at the frequency speci fied in the Surveillance FreqLJency Control*Program and atter the river has 0

returned to normal, following the condition defined below:

a. The level of the Susquehanna River exceeds flood stag~; flood Gage at Harrisburg.

at stage is defin ed as elevation 307 feet the Susqueha nna River Bases The earth dikes are comp acted to provide a stable impervious embankment that protects the site from inu*ndation* during the de Sign flood -of 1,100,000 cfs.

The rip-rap, prov1ded*to prote ct the dikes .from *wave action and the Jlow of the river, _continues dqwn ward into natural ground, for a minimum depth of two feet to preve nt undermining of the dike (Reference s 1 and 2).

P~riodic inspecti9n, and inspection of the dikes* and rip-rap after the river has returned to norm al f~om flood stage, will assur e prope r maint~nance of the dikes, thus assuring protection of the site .durin g the desig n flood.

References (1) UFSAR, Section 2.'6.5 - "Design of Hydraulic Facilities" (2) UFSA R, Figure 2.6-1 7 - 'Typi cal Dike Section" 3-59 Amen dmen t No. 167, 182, 274

C

3. 14 .2 FLOOD CONDITION FOR PL ACING THE UNIT IN HOT Applicability STANDBY Applies to the ri ve r stage for placing the un it in hot standby.

Objective To define the action taken in fe et at the inta ke st ru ct ur e. the event ri ve r elevation reaches 302 Specification 3.14.2.1 If the ri ve r stage reaches el ev at Water Intake Struct io n 302 fe et at th ri ve r flow, the un iture, corresponding to 1,000,000 ecfsRiver condition. will be brought to th e hot standby Bases The dikes provided pr 1,100,000 cf s. The deot ec t the plant si te during the desi approximately 303 fe et sign flood corresponds to an elev gn flood of (Reference l} . The at the River Water Intake St ru ct uration of fe et . The minimum freedike elevation at the intake st ru ct e si te where the dike elboard is at the downstream end of ur e is 305 approximately one fo otevation is 304 fe et providing a free the pl an t the plant si te from flo . Ad equate freeboard is pr board of od in g due to wave action du ovided to pr ot ec flood (Reference 2) . ring the design t Placing the un it elevation providesinanhot standby when the rive~ stag~ reac th at adequate freeboar additional margin of conservatism hes 302 fe et d ex is ts during oper by ation of the un it .assuring References

{1) UFSAR, Fi~ure 2.6-15 - "Dike Freeboard

- Design Flood" (2) UFSAR, Section 2.

6 .. 4 "Flood Studies 11 f -';\ "

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3-60 Amendment No. 157

3.15 AIR TR EA TM EN T SYSTEMS

.i.

3.15.1 . EM ER GE NC Y CO NT RO L RO OM AIR TR EA TM EN T SYS TEM Applicability Applies to the em erg enc y con trol room air treatment sys tem Control Ro om Env elo pe Bounda and its associated filte rs and ry. to the The Control Room Env elo pe Not e (CRE) bou nda ry ma y be ope control. ned intermittently under adm inis trat ive Objective To specify min imu m ava ilab ility and efficiency for the em erg enc system and its ass oci ate d filte y control room air trea tme nt rs.

Spe cific atio ns 3.15.1.1 Exc ept as spe cifi ed in Specification 3.15.1.3 bel ow, systems, AH -E1 8A fan and ass bot h emergency trea tme nt ociated filte r AH -F3 A and AH filte r AH -F3 B sha ll be operab -E18B fan and ass oci ate d le at all times, per the require

3. 15 .1.2 bel ow; whe n containme ments of Spe cific atio n nt integrity is req uire d and whe han dlin g ope rati ons are in pro n irradiated fue l gress. '

3.15.1.2 a. The res ults of the in-place DO P and hal oge nat ed hydrocarb flows on HE PA filters and cha on tests at des ign rcoal abs orb er ban ks shall sho pen etra tion and < 0.05% hal oge w < 0.0 5% DO P nat ed hyd roc arb on penetration DOP tes t will be conducted with , exc ept tha t the prefilters installed.

b. The res ults of laboratory car bon sample ana lysi s shall sho w~

methyl iod ide decontamination 95% rad ioa ctiv e efficiency wh en tes ted in acc 038 03- 198 9 at 30°C, 95% R.H ord ance with AS TM

c. The fan s AH-E1 BA and B sha ll each be sho wn to operate with design flow (40 ,00 0 CFM). in +/- 400 0 CF M of

d. The Con trol Roo m Envelope bou nda ry shall be maintained occ upa nt dos e from a large rad such that the CR E ioactiye rele ase doe s not exc dos e in the lice nsin g basis con eed the cal cul ate d seq uen ces ana lyse s for DBA's occ upa nts are protected from and tha t CR E haz ard ous che mic als and smo ke.

3.15.1.3 Fro m and afte r the date tha t one control roo m air be ino per abl e for reason oth trea tme nt system is ma de or er tha n 3.15.1.2d, rea cto r ope fou nd to han dlin g ope rati ons are permis ration or irradiated fue l sible only dur ing the succeedin red und ant sys tem is verified g 7 days pro vid ed the to be OP ER AB LE.

3.1 5.1 .4 Fro m the dat e tha t bot h control room air trea tme ino per abl e for a rea son other nt sys tem s are made or fou nd tha n 3.15.1.2d, or if the inoper to be can not be ma de operable in able system of 3.1 5.1 .3 7 days, irradiated fuel han dlin g term ina ted in 2 hou rs and rea operations sha ll be cto r shutdown shall be initiate in CO LD SHUTDOWN within d and the reactor sha ll be 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

3-61 Am end me nt No . 56, 67, 76, 149, 190, 226, 264

3.15.1.5 From the date that one or both control room air treatment systems are made or found to be inoperable due to an inoperable Control Room Envelope boundary, actions to implement mitigating actions shall be initiated immediately, verification that the mitigating actions ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and the CRE boundary shall be restored to OPERABLE status within 90 days. Irradiated fuEJI handling operations shall be terminated immediateiy. If the CRE boundary cannot be

\ made OPERABLE in 90 days, reactor shutdown shall be initiated and the reactor shall be in HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The emergency control room air treatment systems AH-E18A and 188 and their associated filters are two independent systems designed to filter the control room atmosphere for intake air and/or for recirculation during control room isolation conditions. Air is recirculated and filtered in the Control Room Envelope (CRE) and within a CRE boundary that limits the inleakage of unfiltered air. Ductwork, valves or dampers, doors, barriers, and instrumentation also form part of the systems. The control building is designed to be automatically placed in the recirculation mode upon an RM-A1 high radiation alarm, air tunnel device actuation, ESAS actuation or station blackout condition. The emergency control room air treatment fan and filter AH-E1 BA or B and AH-F3A or B is designed to be manually started by the operator if a high radiation alarm from RM-A 1 is indicated.

Prefilters and high efficiency particulate absolute (HEPA) filters are installed before the charcoal absorbers to prevent clogging of the iodine adsorbers and remove particulate activity. The charcoal adsorbers are installed to reduce the potential intake of radioiodine to the control room.

If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR Part 50. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.

If one system is found to be inoperable, for reasons other than an inoperable control room envelope boundary, there is no immediate threat to the control room and reactor operation or refueling may continue for a limited period of time while repairs are being made. If the system cannot be repaired within 7 days, the reactor is shut down and brought to cold shutdown within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months and irradiated fuel handling operations are terminated within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

If both systems are found to be inoperable, for reasons other than an inoperable control room envelope boundary, reactor shutdown shall be initiated and the reactor will be brought to cold shutdown in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months and irradiated fuel handling operations will be stopped within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

In-place testing for penetration and system bypass shall be performed in accordance with ANSI N510-1980. Charcoal samples shall be obtained in accordance with ANSI N509-1980. Any HEPA filters found defective shall be replaced with filters qualified according to Regulatory Guide 1.52, Revision 2. Any lot of charcoal_ adsorber which fails the laboratory test criteria shall be replaced with new adsorbent qualified according to ASTM D3803-1989.

3-62 Amendment ~Jo. se June 3, i 980, 226, 264

Laboratory testing of charcoal samples will be performed in accordance with the methods prescribed by ASTM 03803-1989. Design basis accident analyses assume the carbon adsorber is 90% efficient in its total radioiodine removal. Therefore, using a Safety Factor of 2 (Ref. 3), the acceptance criteria for the laboratory test of carbon adsorber is set at greater than or equal to 95% [(100 - 90) / 2 = 5% penetration].

The CRE is the area within the confines of the CRE boundary*that contains the spaces that control room occupants inhabit to control the unit during normal and accident conditions. This area encompasses the control room, and may encompass other non-critical areas to which frequent personnel access or continuous occupancy is not necessary in the event of an accident. The CRE is protected during normal operation, natural events, and accident 1 conditions. The CRE boundary is the combination of walls, floor, roof, ducting, doors, penetrations and equipment that physically form the CRE. The OPERABILITY of the CRE boundary must be maintained to ensure that the inleakage of unfiltered air into the CRE will not exceed the inleakage assumed in the licensing basis analysis of design basis accident (OBA) consequences to CRE occupants. The CRE and its boundary are defined in the Control Room Envelope Habitability Program.

In order for the Emergency Control Room Air Treatment trains to be considered OPERABLE, the CRE boundary must be maintained such that the CRE occupant dose from a large radioactive rele.ase does not exceed the calculated dose in the licensing basis consequence analyses for DBAs, and that CHE occupants are protected from hazardous chemicals and smoke.

The analysis of hazardous chemical releases demonstrates that the toxicity limits are not exceeded in the CRE following a hazardous chemical release. This is because there are no credible hazardous chemical releases that exceed toxicity limits in the CRE (Ref. 4). The evaluation of a smoke challenge demonstrates that it will not result in the inability of the CRE occupants to control the reactor either from the control room or from the remote shutdown panels (Ref. 1).

The control room envelope (CRE) boundary may be opened intermittently under administrative control. This only applies to openings in the CRE boundary that can be rapidly restored to the design condition, such as doors, hatches, floor plugs, and access panels. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls should be proceduralized and consist of stationing a dedicated individual at the opening who is in continuous communication with the operators in the control room. This individual will have a method to rapidly close the opening and to restore the CRE boundary to a condition equivalent to the design condition when a need for CRE isolation is indicated.

If the unfiltered inleakage of potentially contaminated air past the CRE boundary and into the CRE can result in CRE occupant radiological dose greater than the calculated dose of the licensing basis analyses of OBA consequences (allowed to be up to 5 rem TEDE), or inadequate protection of CRE occupants from hazardous chemicals or smoke, the CRE boundary is inoperable. Actions must be taken to restore an OPERABLE CRE boundary within 90 days.

3-62a Amendment No. 66, 67, 76, 108, 149, 167, 226, 246, 264

During the period that the CRE boundary is considered inoperable, action must be initiated to implement mitigating actions to lessen the effect on CRE occupants from the potential hazard$

of a radiological or chemical event or a challenge from smoke. Actions must be taken within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to verify that in the event of a OBA, the mitigating actions will ensure that CRE occupant radiological exposures will not exceed the calculated dose of the licensing basis analyses of OBA consequences, and that CRE occupants are protected from hazardous chemicals and smoke. These mitigating actions (i.e., actions that are taken to offset the consequences of the inoperable CRE boundary) should be preplanned for implementation upon entry into the condition, regardless of whether entry is intentional or unintentional. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of mitigating actions. The 90 day Completion Time is reasonable based on the determination that the mitigating actions will ensure protection of CRE occupants within analyzed limits while limiting the probability that CRE occupants will have to implement protective measures that may adversely affect their ability to control the reactor and maintain it in a safe shutdown condition in the event of a OBA. In addition, the 90 day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most problems with the CRE boundary.

In the event that irradiated fuel handling operations shall be terminated immediately, this does not preclude the movement of fuel to a safe position.

References (1) FSAR Section 9.8 (2) DELETED (3) NRC Generic Letter 99-02, dated June 3, 1999.

. (4) FSAR Section 7.4.5 3-62b Amendment No. 66 June 3, 1980, 226, 264

3.15.2 REACTOR BUILDING PURGE AIR TREATMENT SYSTEM Deleted

' ' 3-62c Amendment No. 66, 67, 76,108,149,1 67,226,246,2 64

3.15.3 AUXILIARY AND FUEL HANDLING BUILDING Al R TREATMENT SYSTEM Deleted 3-62d Amendment No. 66, 76, 122, 167,177,216,248, 264

3.15.4 Fuel Handling Building ESF Air Treatment System Applicability Applies to the Fuel Handling Building (FHB) ESF Air Treat ment System and its associated filters.

Objective To specify minimum availability and efficiency for the FHB ESF Air Treatment System and its associated filters for irradiated fuel handling operations.

Specifications 3.15.4.1 Prior to fuel movement each refueling outage, two trains shall be operable. One train shall be operating continuously whenever TMl-1 irradiated fuel handling operations in the FHB are in progress. *

a. With one train inoperable, irradiated fuel handling opera tions in the Fuel Handling Building may continue provided the redundant train is opera ting.

b. With both trains inoperable, handling of irradiated fuel in the Fuel Handling Building shall be suspended until such time that at least one train is operable and operating.

Any fuel assembly movement in progress may be completed.

3.15.4.2 A FHB ESF Air Treatment System train is opera ble when *its surveillance requirements are met and:

a. The results of the in-place DOP and halogenated hydro carbon tests at design flows on HEPA filters and carbon absorber banks shall show <

0.05% DOP penetration and < 0.05% halogenated hydrocarbon penetration.

b. The results of laboratory carbon sample analysis shall show ~ 95% radioactive methyl iodide decontamination efficiency when tested in accordance with ASTM D3803-1989 at 30°C, 95% R.H.

c. The fans AH-E-137A and B shall each be shown to opera te within +/- 10% of design flow (6,000 SCFM).

Bases Compliance with these specifications satisfies the condition of operation imposed by the Licensing Board as described in NRC's letter dated Octob er 2, 1985, item 1.c.

The FHB ESF Air Treatment System contains, controls, mitigates, monitors and records radiation release resulting from a TMl-1 postulated spent fuel accident in the Fuel Handling Building as described in the FSAR. Offsite doses will be less than the 10 CFR 100 guidelines for accidents analyzed in Chapter 14 (Reference 1).

~-

~ 3-62e Amen dmen t No. 122, 157, 226 278

CONTROU .ED COPY Bases (Continued)

Normal operation of the FHB ESF Air Treatmen t System will be during TMl-1 irradiated fuel movements in the Fuel Handling Building. The system includes air filtration and exhaust capacity to ensure that any radioactive relea se to atmosphere will be filtered and monitore Effluent radiation monitoring and sampling capa d.

bility are provided.

The in-pl ant testi ng for pene trati on and syst em bypa ss shal l be perf orme d in acco rdan with ANSI N510-1980. Cha rcoa l sam ples ce shall be obta ined in acco rdan ce with ANS N509-1980. Any HEPA filter s foun d defe ctive I shal l be repla ced with filter s qual ified acco rding to Reg ulato ry Guid e 1.52, Revi sion 2. Any lot of char coal adso rber whic the labo rato ry test crite ria shal l be replaced h fails with new adso rben t qual ified in acco rdan with ASTM D3803-1989. ce Labo rato ry testi ng of char coal sam ples will be perf orme d in acco rdan ce with the test meth ods pres cribe d by AST M D3803-1989; Test ing of char coal at 95% relat ive hum idity will be requ ired until SL;Jch time that a surv eilla nce to dem onst rate oper abili ty of the heat ers is inco rpor ated by ame ndm ent into the spec ifica tion. The acci dent anal ysis FSAR Cha pter 14 (Ref eren ce 1) assu mes in the char coal adso rber is 90% effic ient in tota l radio iodin e remo val. Ther efor e, usin its g a Safe ty Fact or of 2 (Ref. 2), the acce ptan crite ria for the labo rato ry test of char coal ce adso rber is set at grea ter than or equa l to

((100 - 90) / 2 = 5% pene trati on]. 95%

References (1) UFSAR, Section 14.2.2.1 - "Fuel Handling Accid ent" (2) NRC Gen eric Lett er 99-02, d.ated June 3, 1999.

r,~.

~ ...;* --

t7

-~*

Ame ndm ent No. 122, 157 , 226 3-62f

CONTROWED COPY 3.16 SHOCK SUPPRESS~S (SNUBBERS)

LIMITING CONDITION FOR OPERATION 3.16. l Each saf~ty relate d snubber shall be OPERABLE.

APPLICABILITY:

Whenever the system protected*by the snubber is required to be OOERA BLE.

ACTION:

With one or more snubbers inoperable, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or resto re the inoperable snubber(s) to OPERABLE status and perform an engineering evalu aticn per Speci ficati on 4.17 .l.g.2 on the attach ed componi:nt or decla re the attach ed syste~ inope rable and follow the appro priate action statement for that system.

BASES Snubbers are designed to prevent unres traine d pipe as might occur during an earthquake or severe transmotion *under dynamic loads ient, while allowing normal thermal motion during startu p and shutdown. The conseq snubber due to failur e to activa te (lockup) is an increauence se in of an inope rable the proba bility of struc tural damage to piping as a resul t of a seismic or other

,{:1;,--'r.'.*~:'!. initia ting dynamic loads. The consequence of snubber inope rabili event

,\~r.*,) failur e to extend or retrac t is an increa se in the proba bility of ty due to damage to piping as a resul t of thermal motion. It is theref ore struc tural all snubbers requi red*to protec t the primary coola nt system or anyrequi red that safety system or component which is required to be operable must other ooerable. During plant conditions _other than opera ting, snubbers also be systems that are required to be operable during that plant condi on.tho se required to be operable

tion are also

~,"

3-63 A~endment No. ~~. 106

tf'n'-a-nl

'VVB'll ~01"*'~J ED oo,~v . r Y INTENTIONALLY BLANK PAGES 3-64 through 3-79 Ar:iendment No. 10, CL, 11, 3-64 105

3 17 REACTOR BtJILDING AIR TEMP ERATURE App1icabi1ity This sp ec ifi ca tio n app1ies piimary containment durin to the average ai r temperature of the g power operations.

Objective To assure th at the tempe the Reactor Bui1ding are ratures assumed in the structura1 .ana1ys flot exceeded. is of Specification 3.17.1 Primary containmen aver sh al l not exceed 13t0°F age ai r temperature abov an e Elev. 320 Elev. 320 sh al l not exce d average ai r temperature below ed 120°F.

3.17.2 If~ while the reac r is temperature lim its toare ex cr iti ca l, th~ above st at ed shall be reduced to the ce eded, the average tempe in at le as t HOT STANDBY. above lim its within 8 ho rature*

urs, or be in COLD SHUTDOWN within within the next si x (6) hours and the following th ir ty (30)

3. 17 .3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

The primary containment calculated as follows: average ai r temperature sb al l be a) The av~rage temperat ca lc ul at ed by taking urthe e above elevati6n 320 wi ll be temperatures from at le asar ith m et ic average of the elevation 320. A li st of t 13 lo ca tio ns above lo ia tio ns is given below b) Thi average temperatu ca lc ul at ed by taking th res below elevation 320 will be temperatures from at leeasar ith m et ic average of -th e A li st of locations is givt 4 lo ca tio ns below Elev. 320 en below.

Location Location SE Wall Elev. 352 1 NW Sec Sh1eld Elev 352 1 NE Wall Elev 314'*

NE Sec Shield Elev 35 S Wail Elev 3147"*

E Wall Elev 382 1 ~2 1

NW Wall Elev 314 1

NE Sec Shield Elev 352 1 E Sec Shield Elev 352 1 NW Sec Shield Elev 352 1 S Rx Wall Elev 32 1 NE Sec Shield Elev 35 NE ija ll Elev 287 1 1*

NW Sec Shield Elev 3522 1 1

S Wall Elev 287 1*

NW Wall Elev 352 1 NW Wa11 Elev 287'*

E Wall Elev 400 1 ~ E Sec Shield E1ev 352' S Sec Sh ie ld ~e v 352 1 NW Sec Shield Elev 287 1

NW Sec Shield Elev 352 NE Sec Shield Elev N Sec Shield Elev 3636441 1 1 NOTE: (1)

Detectors lo cated below elev 320 1

3-80

Amendment No. 41, 7!, 15 7

C Bases The spec ified temp eratu re 1imi ts asgur e that the conta inme nt desig n temp eratu re and press ure will not be excee ded in the even t of a desig n basis loss of ,cool ant accid ent. The limit s also assur e the main tenan ce of accep table ambi ent envir onme ntal cond ition s.for safet y-rel ated comp onent s locat ed insid e the conta inmen t~

Amendment No. 41 3-81 (5-24 -78)

y PAGES 3~82 THROUGH 3-85 INTENTIONALLY :BLANK

.e7*** ) .Amendment No. 111 (5-2) 1-78 )

'e\'*"**

COPY THIS PAGE LEFT BLANK INTENTIONALLY 3-86 (Pages 3-87 through 3-94 dele ted)

Anendrient No. 32, 101 11.6

CONTROLLED COPY 3 .19 CONTAINMENT SYSTEMS 3.19 .l CONTAINMENT STRUCTURAL INTEGRITY Appl kab il ity:

Applies to the stru ctur al inte grit y of the reac tor building.

OBJECTIVE:

To verify containment struct.ura l inte grit y in accordan tendon surveillance program for the reac ce with the i nservi ce tor building pres tres sing system.

Specification 3.19 .1.l With the stru ctur al inte grit y of the cont the inservice tendon surveillance programainm ent not conforming to requ I

for the tendon lift off forces, perform an engiirem neer ents of 4;4.2.1 of the stru ctur al inte grit y of the containment to ing evaluation SHUTDOWN is required. The margins avai labl e in thedetermine if COLD design may be considered during the inve stig atio n. containment acce ptab ility of the containment tendons cannot be If the within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , rest ore the stru ctur al inte grit y established limi ts within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least.HOT STANDBY to within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the followingwithin the 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.19 .1.2 DELETED 1

J;(:;?J;,~~j

\;( )':,;f

,;,,,.;J 3-95 Amendment No. ,,,. ii,, 187

CONTROL.! ED

<f:('(,":~)

\' J.- 3.20 (DELETED) i}t"(i17" Anendnen t No. ,81', 13 9 3-95a

UEDOOPY 3.21 RADIOACTIVE EFFLUENT INSTRUMENTATION

{'.:*:).

3.21.1 Deleted Radioactive Liquid Effluent Instrumentation Deleted 3.21.2 Radioactive Gaseous Process and Effluent Monitoring Instr umentation Deleted 3.22 RADIOACTIVE EFFLUENTS Deleted 3.22.1 Liquid Effluents Deleted 3.22.2 Gaseous Effluents Deleted 3.22.3 Solid Radioactive Waste Deleted

.3.22.4 Total Dose Deleted 3.23 RADIOLOGICAL ENVIRONMENTAL MONITORING Deleted 3.23 .l Monitoring Program Deleted 3.23.Z

Land Use Census Deleted 3.23.3 Inter labo rator y Comparison Program Deleted 3-96 (3-97 .thru 3-127 deleted)

Amendment No. 72,. 7J, JJ, JJ, J77,-JJ~, JJZ, 197.

i,r, JP-, nz, 1z,. ]17, u,, JJ7, UJ, J71,

CONTROLLED COPY

3. 24 Rea cto r Ves sel Wa ter Lev el Ind ica tion Ap pli cab ilit y Ap plie s to the ope rab ilit req Lev el Ind ica tio n whe n theyrea uire me nts for the Rea cto r Ve sse l Wa ter cto r is cri tic al.

Ob jec tive s To ass ure ope rab ilit y of the whi ch may be use ful in dia gnoRea cto r Ves sel Wa ter Lev el ins tru me nta tion lea d to ina deq uat e cor e coo ling sin g situ ati ons whi ch cou ld rep

. res ent or Spe cif ica tio n Two cha nne ls of the Rea cto r Ves sha ll be OPERABLE. sel Wa ter Lev el Ins tru me nta tion Sys tem If one cha nne l bec ome s INO OPERABLE wit hin 30 day s. PER If ABLE tha t cha nne l sha ll be ret urn ed to day s, wit hin 14 day s, sub mit the cha nne l is not res tor ed wit hin 30 the det ail s of the ino per abi a spe cia l rep ort to the NRC pro vid lity , to inc lud e cau se, act ion ing tak en and pro jec ted dat e for bei ng ret urn to OPERABLE sta tus .

Wit h no cha nne ls OPERABLE, sta tus wit hin 7 day s. If atone cha nne l sha ll be res tor ed to OPERABLE 7 day s, wit hin 14 day s, sub mitlea st one cha nne l is not res tor ed wit hin the det ail s of the ino per a spe cia l rep ort to the NRC pro vid ing and pro jec ted dat e for retabi lity , to inc lud e cau se, act ion urn to OPERABLE sta tus . bei ng tak en The Rea cto r Ves sel Wa ter Lev ind ica tio n of the tre nd in wat el Ind ica tio n (Re fere nce 1) pro vid es ves sel dur ing the app roa ch to er inv ent ory in the hot leg s and *rea cto r man ner add itio nal inf orm atio n ina deq uat e cor e coo ling (IC C). In thi s dia gno se the app roa ch of ICC may be ava ilab le to the ope rat or to tak en to res tor e cor e coo lingand to ass ess the ade qua cy of res pon ses Eac h Rea cto r Ves sel Wa ter Lev ind ica tio n and a rea cto r ves el cha nne l is com pris ed of a hot leg l.ev el sel lev el ind ica tio n.

The sys tem is req uir ed to be pla nt is cri tic al. ope rab le (as def ine d pre vio usl y) whe n the The sys tem is an inf orm atio app roa ch to ina deq uat e cor e ncoosys tem to aid the ope rato r dur thi s sys tem . ling . The re is not reg ula tor ying the lim it for Ino per abi lity of the sys tem rem sys tem . Oth er use ful ins tru me ove s the ava ila bil ity of an inf nta orm atio n tion for ina deq uat e cor e coo ling be ava ilab le. The Sub coo ling wil l det erm ine sub coo ling mar gin whe Ma rgin Ind ica tion Sys tem is rel ied upo n to or when nat ura l cir cul ati on can n the rea cto r coo lan t pum ps are ope rati ng cir cul atio n can not be ver ifie be ver ifie d. When nat ura l or for ced by man ual cal cul ati on, bas ed d, the mar gin to sat ura tio n is det erm ine d the rmo cou ple s) and pre ssu re on rea cto r coo lan t tem per atu re {in cor e/

and stea m tab les . See Tec h. ind ica tio ns ava ilab le in the con tro l roo m Spe c. 3.5 .5.

3-1 28 Amendment No. 147 , 157 , 1-9-+/--,

254

CONTROLLED. COPY The system is not a required system to mitigate evaluated accidents. It may be useful to have the system operable but there will be no adverse impac t if it is not operable.

The LCO action statement provides the level of emphasis required for an information

.* system.

The Reactor Vessel Wate r Level is a Regulatory Guide 1.97 Category 1 variable.

Reference (1) UFSAR, Update Section 7.3.2.2(c)1 O(d) - "Reactor Coola nt Inventory Trending System".

(2) USNRC Regulatory Guide 1.97.

tt~~').

9 .

,e.} Amen dmen t No. 147, 167, 191, 251 3-129

4. SURVEILLANCE STANDARDS 4.0.1 During Reactor Operational Conditions for which a Limiting Condition for Operatio n

(LCO} does not require a system/component to be operable, the associated surveillance requirements do not have to be performed. Prior to declaring a system/

component operable, the associated surveillance requirement must be current.

Failure to perform. a surveillance within the specified Frequency shall be failure to meet the LCO except as provided in 4.0.2. .

4.0.2 If it is discovered th~t a surveillance was not performed within its specified frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. The delay period is only applicable when there is a reasonable expectation the surveillance will be met when performed. A risk evaluation shall be performed for any surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

If the surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable condition(s} must be entered.

When the surveillance is performed within the delay period and the surveillance is not met, the LCO must immediately be declared not met, and the applicable condition (s}

must be entered.

SR 4.0.1 establishes the requirement that SRs must be met during the REACTOR OPERATING CONDITIONS or other specified conditions in the SRs for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This specifica tion is to ensure that surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits.. Failure to meet a surveilla nce within the specified frequency, in accordance with definition 1.25, constitutes a failure to meet an LCO; Surveillances may be performed by means of any series of sequential, overlapping, or total

steps provided the entire Surveillance is performed within the specified frequency.

Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:

a. The system or components are known to be inoperable, although still meeting the SRs or '

b. The requirements of the Surveillance(s) are known to be not met between required Surveillance performances.

Surveillances do not have to be performed when the unit is in a REACTOR OPERAT ING CONDITION or other specified condition for which the requirements of the associat ed LCO are not applicable, unless otherwise specified. Unplanned events may satisfy the requirem ents (including applicable acceptance criteria) for a given SR. In this case, the unplanne d event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given REACTOR OPERATING CONDITION or other specified condition.

4-1 Amendm ent No. 46, QQ, 100,124,138, 181, 2W, 292

Surveillances, including surveillances invoked by LCO required actions, do not have to be performed on inoperable equipment becauf:ie the actions define the remedial measures that apply. Surveillances have to be met and performed in accordance with the specified frequency, prior to returning equipment to OPERABLE status.

Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE:. This includes ensuring applicable surveillances are not failed and their most recent performance is in accordance with the specified frequency. Post maintenance testing may not be possible in the current REACTOR OPERATING CONDITION or other specified conditions in the SRs due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not othen,_vise believed to be incapable of performing its function. This will allow operation to proceed to a REACTOR OPERATING CONDITION or other specified condition where other necessary post maintenance tests can be completed.

Some examples of this process are:

a. Emergency feedwater (EFW) pump maintenance during refueling that requires testing at steam pressures greater than 750 psi. However, if other appropriate testing is satisfactorily completed, the EFW System can be considered OPERABLE. This allows startup and other necessary testing to proc~ed until the plant reaches the steam pressure required to perform the EFW pump testing.

b. High pressure injection (HPI) maintenance during shutdown that requires system

~ functional tests at a specified pressure. Provided other appropriate testing is V@J1 satisfactorily completed, startup can proceed with HPI considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing.

SR 4.0.2 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a surveillance has not been-performed within the specified frequency. A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified frequency, whichever is greater, applies from the point in time that it is discovered that the required surveillance has not been performed in accordance with Surveillance Standard 4.0.2 and not at the time that the specified frequency was not met.

The delay period provides an adequate time to perform surveillances that have been missed. This delay period permits the performance of a surveillance before complying with required actions or other remedial measures that might preclude performance of the surveillance.

The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the surveillance, the safety significance of the delay in completing the required surveillance, and the recognition that the most probable result of any particular surveillance being performed is the verification of

. conformance with the requirements.

4-1a Amendment No. 46, QQ, 100, 124, 1as, 181,256,292

~ Bases (Contd.}

~

When a surveillance with a frequency based not on time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering power operation after each fuel loading, or in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, etc.) is discovered to not have been performed when specified, ;:-_::**

Surveillance Standard 4.0.2 allows for the full .delay period of up to the specified frequency to perform the surveillance. However, since there.is not a time* interval specified, the missed surveillance should be performed at the first reasonable opportunity. When a Section 6.8, "Procedures and Programs," specification states that the provisions of TS 4.02 are applicable, a 25% extension of the testing interval, whether stated in the specification or incorporated by reference, is permitted.

SurVeillance Standard 4.0.2 provides a time limit for, and allowances for the performance of, surveillances that become applicable as a consequence of operating condition changes imposed by required LCO actions.

SR 4.0.2.is only applicable if there is a reasonable expectation the associated equipment is OPERABLE or that variables are within limits, and it is expected that the Surveillance will be met when performed. Many factors should be considered, such as the period of time since the Surveillance was last performed, or whether the Surveillance, or a portion thereof, has ever been performed; and any other indications, tests, or activities that might support the expectation that the Surveillance *will be met when performed. An example of the use of SR 4.0.2 would be a relay contact that was not tested as required in accordance with a particular SR, but previous successful performances of the SR included the relay contact; the adjacent, physically connectec:i relay con~acts were tested during the SR performance; the subject relay contact has been tested by another SR; or historical operation of the subject relay contact has been successful. It is not suffiqientto infer the behavior of the associated equipment from the performance of similar equipment. The rigor of determining whether there is a reasonable expectation a Surveillance will be met when performed should increase based on the length of time since the last performance of the Surveillance. If the Surveillance has been performed recently, a review of the Surveillance history and equipment performance may be sufficient to support a reasonable expectation that the Surveillance will _be met when performed. For Surveillances that have not been performed for a long period or that have never been performed, a rigorous evaluation based on objective evidence should provide a high degree of confidence that the equipment is OPERABLE. The evaluation should be documented in sufficient detail to allow a knowledgeable individual to understand the basis for the determination.

Failure to comply with specified surveillance frequencies is expected to be an infrequent occurrence. Use of the delay period established by Surveillance Standard 4.0.2 is a flexibility which is not intended to be used repeatedly to extend surveillance intervals. While up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit of the specified frequency is provided to perform the missed surveillance, it is expected that the missed surveillance will be performed at the first reasonable opportunity.

The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the surveillance as well as any plant configuration changes required or shutting the plant down to perform the surveillance) and impact on any analysis assumptions, in addition to unit conditions, planning, availability of personnel, and the time required to perform the surveillance. This risk impact should be managed through the program 4-1b Amendment No. 181, 256, 2QO, 292

Bases (Contd.)

in place to implement 10CFR 50.65 (a)(4) and its implementation guidance, NRG Regulatory Guide 1.182, 'Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants'. This Regulatory Guide addresses consideration of temporary and aggreg ate risk impacts, determination of risk management action thresholds, and risk manag ement action up to and including plant shutcfown. The missed surveillance should be treated as an emergent condition as.discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluat ion should be commensurate with the importance of the component. Missed surveillances for important components should be analyzed quantitatively. If the results of the risk evaluat ion determine the risk increase is significant, this evaluation should be used to determine the safest course of action. All missed surveillances will be placed in the licensee's Corrective Action Program.

If a surveillance is not completed within the allowed delay period, then the equipm ent is considered inoperable or the variable is considered outside the specified limits and the completion times of the required actions for the applicable LCO conditions begin immediately upon expiration of the delay period. If a surveillance is failed within the delay period, then the equipment is inoperable; or the variable is outside the specifie d limits and the completion times of the required actions for the applicable_ LCO conditions begin immediately upon failure of the surveillance.

Completion of the surveillance within the delay period allowed by this specific ation, or within the completion time of the actions, restores compliance.

4-1c Amendment No. 181, 256, 290, 292

4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation.

Objective To specify t~e minim'um frequency and type of surveillance to be applied to unit equipment and conditions.

Specification 4.1.1 The type of surveillance required for reac tor protection system, engineered safety feature protection system, and heat sink prote ction system instrumentation when the reactor is critical shall be as stated in Table 4.1-1

. The frequency of surveillance required for the instrumentation shown in Tabl e 4.1-1 is specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.1-1.

4.1.2 Equipment and sampling test shall be perfo rmed as detailed in Tables 4. 1-2, 4.1-3, and 4.1-5 at the frequencies specified in the Surveillan ce Frequency Control Program unless otherwise noted in Tables 4.1-2, 4. 1-3, and 4. 1-5.

4.1.3 Each post-accident monitoring instrume ntation channel shall be demonstrated OPERABLE by the performance of the check, test and calibration at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.1-4.

4.1.4 Each remote shutdown system function shown in Table 3.5-4 shall be demonstrated.

OPERABLE by the performance of the following check, test, and calibration at the

frequencies specified in the Surveillance Frequenc y Control Program:

a) Perform CHANNEL CHECK for each required instrumentation channel that is normally energized.

b) Verify each required control circuit and transfer sWitch is capable of performing the intended function.

c) Perform CHANNEL CALIBRATION for each requ ired instrumentation channel (excludes source range flux).

Bases Check Failures such as blown instrument fuses, defective indicators, or faulted amplifiers which result in "upscale" or "downscale" indication can be easil y recognized by simple observation of the functioning of an instrument or system. Furthermo re, such failures are, in many cases, revealed by alarm or annunciator action. Comparison of output and/or state of independent channels measuring the same variable supplements this type of built-in surveillance. The acceptance criteria for the daily check of the Makeup Tank pressure instrument will be maintained within the error used to develop the plant oper ating limit. Based on experience in operation 0 .

, of both conventional and nuclear systems, when frequency stated in the Surveillance Frequenc reactor system instrumentation.

the unit is in operation, the minimum checking y Control Program is deemed adequate for 4-2 Amendment No. 78, 123, 138, 166, 1a7, 158, 181, 21e, 22a, 227,274

Bases (Cont'd)

The 600 ppmb limit in Item 4, Table 4.1-3 is used to meet the requirements of Section 5.4. Under other circumstances the minimum acceptable boron concentration would have been zero ppmb.

Calibration .

Calibration shall be performed to assure the presentation and acquisition of accurate information.

The nuclear flux (power range) channels amplifiers shall be checked at the frequenc y specified in the Surveillance Frequency Control Program against a heat balance standard and calibrated if necessary. The frequency of heat balance checks will assure that the difference between the out-of-core instrumentation and the heat balance remains less than 4%.

Channels subject only to "drift" errors induced within the instrumentation itself can tolerate longer intervals between calibrations. Process system instrumentation errors induced by drift can be expected to remain within acceptance tolerances if recalibration is performed at the frequency specified in the Surveillance Frequen cy Control Program.

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.

Thus, minimum calibration frequencies set forth in the Surveillance Frequen cy Control Program are considered acceptable.

Testing On-line testing of reactor protection channels is required at the frequenc y specified in the Surveillance Frequen cy Control Program on a rotational basis. The rotation scheme is designed to reduce the probability of an undetected failure existing within the system and to minimize the likelihood of the same systema tic test errors being introduced into each redundant channel (Referen ce 1).

Surveillance Frequencies are controlled under the Surveillance FrequeAcy Control Program.

Upon detection of a failure that prevents trip action in a channel, the instrumentation associated with the protection parameter failure will be tested in the remaining channels. If actuation of a safety channel occurs, assurance will be required that actuation was within the limiting safety system setting.

The protection channels coincidence logic, the control rod drive trip breakers and the regulating control rod power SCRs electronic trips, are trip tested at the frequency specified in the Surveillance Frequen cy Control Program. The trip test checks all logic combinations and is to be performed on a rotational basis.

Discover y of a failure that prevents trip action requires the testing of the instrume ntation associated with the protection parameter failure in the remaining channels.

For purposes of surveillance, reactor trip on loss of feedwater and reactor trip on turbine trip are considered reactor protection system channels.

4-2a Amendment No. 78,167 ,181,20 0,216,2 66,274

,.. , Bases (Cont'd}

The equipment testing and system sampling frequencies specif ied in the Surveillance Frequency Control Program are considered adequate to mainta in the equipment and systems in a safe operational status.

The primary to secondary leakage surveillance in TS Table 4.1-2, Item 12, verifies that primary to secondary leakage is less than or equal to 150 gallons per day through any one (1) SG.

Satisfying the primary to secondary leakage limit ensures that the operational leakage performance criterion in the Steam Generator Program is met.

If this surveillance is not met, compliance with TS 3.1.1.2, "Steam Generator (SG) Tube Integri ty," and TS 3.1.6.3, should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational leakage rate limit applies to leakag e through any one SG. If it is not practical to assign the leakage to an individual SG, all the primar y to secondary leakage should be conservatively assumed to be from one SG:

The TS Table 4.1-2 primary to secondary leakage surveillance is modified by a Note, which states that the initial surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

The TS Table 4.1-2 primary to secondary leakage surveillance frequency specified in the Surveillance Frequency Control Program is a reasonable interva l to trend primary to secondary leakage and recognizes the importance of early leakage detect ion in the prevention of accidents. The primary to secondary leakage is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 5) .

The surveillance test procedures for the Variable Low Pressure Trip Setpoint do not compare the as-found Trip Setpoint (TSP) to the previous surveillance test as-left TSP. Basing operability determinations for the as-found TSP on the Nominal Setpoint (NSP) is acceptable because:

1. The NSP as-left tolerance specified in the surveillance test procedures is less than or equal to the calculated NSP as-left tolerance.

2. The NSP as-left tolerance is not included in the Total Loop Uncertainty (TLU) calculation. This is acceptable because the NSP as-left toleran ce specified in the surveillance test procedures is less than half of the calculated NSP as-left tolerance.

This prevents masking of excessive drift from one side of the toleran ce band to the other.

3. The pre-defined NSP as-found tolerance is based on the square root of the sum of the square of the instrument accuracy, M&TE accuracy and drift.

The NSP as-left tolerance is not included in this calculation.

Credible uncertainties for the Variable Low Pressure Trip Setpo int include instrument uncertainties during normal operation including drift and measu rement and test equipm ent uncertainties. In no case shall the pre-defined as-found accep tance criteria band overlap the Allowable Value. If one end of the pre-defined as-found accep tance criteria band is truncated due to its proximity to the Allowable Value, this does not affect the other end of the pre-defined as-found acceptance criteria band. If equipment is replaced, such that the previous as-left value is not applicable to the current configuration, the as-found acceptance criteria band is not applicable to calibration activities performed immediately followi ng the equipment replacement.

4-2b Amen dment No. 181,2 26, 2a6, 261,2 62,27 1,274

.,,.. Bases (Cont'd)

The TSP is stored in wire mesh baskets placed inside the containm ent at the 281 ft elevatio n.

Any quantity of TSP between 18,815 lb and 28,840 lb. will result in a pH in the desired range.

If it is discovered that the TSP in the containment building is not within limits, action must be taken to restore the TSP to within limits. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed for restoring the TSP within limits, where possible, because 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is the same time allowed for restorat ion of other ECCS components.

Surveill ance Testing Periodic determination of the mass of TSP in containm ent must be performed due to the possibil ity of leaking valves and components in the containm ent building that could cause dissolut ion of the TSP during normal operation. The surveilla nce is required to determi ne that ~

18,815 lbs and s 28,840 lbs are contained in the TSP baskets. This requirement ensures that there is an adequate mass of TSP to adjust the pH of the post LOCA sump solution to a value ~

7.3 ands 8.0. The periodic verification is required at the frequen cy specified in the Surveill ance Frequen cy Control Program.

Periodic testing is performed to ensure the solubility and bufferin g ability of the TSP after exposur e to the containm ent environment. Satisfactory complet ion of this test assures that the TSP in the baskets is "active." Adequate solubility is verified by submerging a represe ntative sample, taken via a sample thief or similar instrument, of TSP from one of the baskets in containm ent in un-agitated borated water heated to a tempera ture representing post-LO CA conditio ns; the TSP must complet ely dissolve within a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period. The test

. time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is to allow time for the dissolve d TSP to naturally diffuse through the un-agitated test solution .

Agitatio n of the test solution during the solubility verification is prohibited, since an adequa te standar d for the agitation intensity (other than no agitation) cannot be specified

. The agitation due to flow and turbulen ce in the containment sump during recircµlation would significantly decreas e the time required for the TSP to dissolve. Adequa te bufferin g capabili ty is verified by a measur ed pH of the sample solution, following the solubilit y verification, between

.7.3 and 8.0.

The sample is cooled and thoroug hly mixed prior to measuri ng pH. The quantity of the TSP sample, and quantity and boron concentration of the water are chosen to be represe ntative of post-LO CA conditions.

REFER ENCE (1) UFSAR, Section 7.1.2.3( d) - "Periodic Testing and Reliabilityu (2) NRC SER for BAW-10 167A, Supplem ent 1, Decemb er 5, 1988.

(3) BAW-10 167, May 1986. .

(4) BAW-10 167A, Supplem ent 3, February 1998.

(5) EPRl,."P ressuriz ed Water Reactor Primary-to-Secondary Leak Guidelines."

4-2d Amendm ent No. 261,26 3,274

T1-\DLE 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS

)>

3 (I) 0.. CHANNEL DESCRIPTION CHECK{c) TEST(c) CALIBRATE(c) REMARKS 3

-z (I)

J 1. Protection Channel NA NA Coincidence Logic

!?

2. Control Rod Drive Trip NA NA (1) Includes independent testing of shunt Breaker trip and undervoltage trip features.

3. Power Range Amplifier (1) NA (2) (1) When reactor power is greater than 15%.

(2) When above 15% reactor power run a heat balance check at the freque11.cy specified in the Surveillance Frequency Control Program. Heat balance calibration shall be performed whenever heat balance exceeds indicated neutron power by more than two percent.

4. Power Range Channel (1 )(2) (1) When reactor power is greater than 60% verify imbalance using incor.e instrumentation .

.i:,..

I I\)

w (2) When above 15% reactor power calculate axial offset upper

-..,I

.i:,.

and lower chambers after each startup if not done within the previous seven days.

5. Intermediate Range Channel (1) NA (1) When in service. I

6. Sourqe Range Channel (1) NA (1) When in service.

7. Reactor Coolant Temperature Channel

TABLE 4.1-1 {Continued)

)>

3 CD CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c) REMARKS a.

3

8. High Reactor Coolant CD Pressure Channel zp

9. Low Reactor Coolant Pressure Channel

10. Flux-Reactor Coolant Flow Compa rator

11. Reactor Coolant Pressure-Temperature See Notes (a) and (b}.

Comparator

12. Pump Flux Comparator

13. High Reactor Building Pressure Channel

~

I

~ 14. High Pressure Injection NA NA Logic Channels

15. High Pressure Injection Analog Channels

a. Reactor Coolant (1) (1) When reactor coolant system is pressurized Pressure Channel above 300 psig or Tave is greater than 200°F

16. Low Pressure Injection NA NA Logic Channel

17. Low Pressure Injection Analog Channels

a. Reactor Coolant (1) (1) When reactor coolant system is pressurized Pressure Channel above 300 psig or Tave is greater than 200°F

18. Reactor Building Emergency NA NA Cooling and Isolation System Logic Channel

TABLE 4.1-1 (Continued)

CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c} REMARKS

)> 19. Reactor Building Emergency 3 Cooling and Isolation CD

I System Analog Channels a.

3 CD

-pz

I a.

b.

Reactor Building 4 psig Channels RCS Pressure 1600 psig (1) (1) (1) When CONTAINMENT INTEGRITY is required.

(1) (1) NA (1) When RCS Pressure> 1800 psig.

c. Deleted

d. Reactor Bldg. 30 psi (1) (1) (1) When CONTAINMENT INTEGRITY is pressure switches required.

e. Reactor Bldg. Purge (1) (1) (1) When CONTAINMENT INTEGRITY is Line High Radiation required.

(AH-V-1A/D)

f. Line Break Isolation (1) (1) (1) When CONTAINMENT INTEGRITY is Signal (ICCW & NSCCW) required.

20. Reactor Building Spray NA NA System Logic Channel

.i:,..

- 01 I

21. Reactor Building Spray NA 30 psig pressure switches

22. Pressurizer Temperature NA Channels I\) 23. Control Rod Absolute Position (1) NA (1) Check with Relative Position Indicator CX>

24. Control Rod Relative Position (1) NA (1) Check with Absolute Position Indicator

25. Core Flooding Tanks

a. Pressure Channels

b. Level Channels NA NA NA NA I.

26. Pressurizer Level Channels NA

TABLE 4.1-1 (Continued) i~ CD

J NO. CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c) REMARKS

""-13

~CD

J 2

0

27. Makeup Tank Instrument Channels:

a. Level ( 1) NA

~

(1) When Makeup and Punfication System is in operation.

b. Pressure ( 1) NA

28. Radiation Monitoring Systems*

a. DELETED (1) Using the installed check source when background is less .than twice the expected

b. DELETED increase in cpm which would result from the check source alone. Background readings

c. DELETED greater than this value are sufficient in themselves to show that the monitor is

d. RM-A2P (AB Atmosphere particulate) (1)(4) (4) (4) functioning.

e. RM-A21 (AB Atmosphere iodine) (1 )(4) (4) (4) (2) DELETED

f. RM-A2G (RB Atmosphere gas) (1 )(4) (4) (4) (3) DELETED (4) RM-A2 operability requirements are given in T.S. 3.1.6.8

29. High and Low Pressure NIA NIA Injection Systems:

Flow Channels

Includes only monitors indicated under this item. Other T.S. required radiation monitors are included in specifications 3.5.5.2, 4.1.3, Table3.5-1 item C.3.f, andTable4.1-1 item 19e.

~

TABLE 4.1-1 (Continued)

)> CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c) REMARKS 3*

ti

, 30. Borated Water Storage NA 1 Tank Level Indicator 3

0 a 31. DELETED z

~ 32. DELETED

33. Containment Temperature NA NA

34. lncore Neutron Detectors (1) NA NA (1) Check functioning; including functioning of computer readout or recorder readout when reactor power is greater than 15%.

35. Emergency Plant Radiation (1) NA (1) Battery Check.

Instruments

36. (DELETED)

37. Reactor Building Sump NA NA Level

TABLE 4.1-1 (Continued)

)>

3 CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c) REMARKS

I>

l.

3 38. OTSG Full Range Level .NA

-z p

I>

39. Turbine Overspeed Trip NA NA

40. Deleted

41. Deleted

42. Diesel Generator NA NA Protective Relaying

43. 4 KV ES Bus Undervoltage Relays (Diesel Start)

a. Degraded Grid NA (1) (1) Relay operation will be checked by local lest pushbuttons.

11 11

b. Loss of Voltage NA (1) (1) Relay operation will be checked by

.,u D ./:a.

~ local test pushbuttons.

-.J

,, 44. Reactor Coolant Pressure (1) (1) When reactor coolant system is DH Valve Interlock Bistable pressurized above 300 psig or T a,e is greater than 200°F.

45. Loss of Feedwater Reactor Trip (1) (1) (1) When reactor power exceeds 7%

power. *

46. Turbine Trip/Reactor Trip (1) (1) (1) When reactor power exceeds 45%

power.

47. a. Pressurizer Code Safely Val\{e (1) NA (1) When Tave is greater than 525°F.

and PORV Tailpipe Flow Monitors

b. PORV - Acoustic/Flow NA (1) (1) When Ta,~ is greater than 525°F.

48. PORV Setpoints NA (1) ( 1) Per Specification 3. 1.12 excluding valve operation.

)>

3 CHANNEL DESCRIPTION CHECK(c)

TABLE 4. 1-1 \....Jntinued)

TEST(c) CALIBAATE(c) REMARKS CD

I 49. Saturation Margin Monitor (1) (1)

a. (1) When Ta,e is greater than 525°F.

3 z-CD

, 50. Emergency Feedwater Flow NA (1) (1) When Tave is greater than 250°F.

Instrumentation p

51. Heat Sink Protection System

a. EFW Auto Initiation (1) Includes logic test only.

Instrument Channels

1. Loss of both Feedwater NA (1)

Pumps

2. Loss of All RC Pumps NA (1)

3. Reactor Building NA Pressure

4. OTSG Low Level

~

I

-...,I b. MFW Isolation OTSG Low NA Ill Pressure C. EFW Control Valve Control System

1. OTSG Level Loops

2. Controllers NA

d. HSPS Train Actuation Logic NA (1)

-N

-..J 52. Backup lncore Thermocouple ( 1) NA (1) When Tave is greater than 250°F.

.j::,,

Display

53. Deleted

54. Reactor Vessel Water Level NA

NA Notes (a) If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance then the channel shall be evaiuated to verify that it is functioning as required before returning the channel to service. Enter condition into Corrective Action Program.

(b) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conversative than the NSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures to confirm channel performance. The NSP and the methodologies used to determine the as-found and the as-left tolerances are specified in a document incorporated by reference into the UFSAR. ,

(c) Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table. I.

TABLE 4.1-2

if(

MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency

1. Control Rods Rod drop times of all Note 1 full length rods

2. Control Rod Movement of each rod Note 1, when reactor is Movement critical *

3. Pressurizer Setpoint In accordance with the Safety Valves INSERVICE TESTING PROGRAM

4. Main Steam Setpoint In accordance with the Safety Valves INSERVICE TESTING PROGRAM

5. Refueling System Functional Start of each Interlocks refueling period

6. (Deleted}

7. Reactor Coolant Evaluate Note 1, when reactor*

System Leakage coolant system temperature is greater than 525 degrees F (Not applicable to primary-to-secondary lec;1.kage.)

B. (Deleted}

9. Spent Fuel . Functional Each refueling period Cooling System prior to fuel handling 1o. Intake Pump {a) Silt Accumulation - Note 1 House Floor Visual inspection (Elevation of Intake Pump 262 ft. 6 in.) House Floor (b) Silt Accumulation Note 1 Measurement of Pump House Flow

11. Pressurizer Block Functional* Note 1 Valve (RC-V2)

12. Primary to Secondary Evaluate Note 1 (Note: Not required Leakage to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.)

Function shall be demonstrated by operating the valve through one complete cycle of full travel.

Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

4-8

~ Amendment No. ea, M, -78, 49, .i+a, 4-Qi, 2=14, 2~@. 26"1, 2+4, 290

~

TABLE4.l-3 MINIMUM SAMPLING FREQUENCY Frt:gul!ncy

l. Rt:iJClor Coolant a. Verify reactor coolant DOSE EQUIV Al.ENT Xe-133 i} Note I (during all plant conditions except REFUELING specific activity is less than or equal tu 797 SHUTDOWN and COLD SHUTDO~N).

microcuries/gram.

ii) One Sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15% uf the RATED THERMAL POWER within a one hour period during all plant conditions except REFUELING SHUTDOWN and COLD SHUTDOWN.

b. isotopic Analysis fur DOSE EQUIVALENT i) Nute I (during power operations).

1-131 Concentration ii) Om: Sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months lulluwiug a THERMAL POWER change exceeding 15'-k, uf the RATED THERMAL POWER within a one hour period during all plant conditions except REFUELING SHUTDOWN and COLD SHUTDOWN.

iii) # Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the specifo: activity exceeds 0.35 µCi/gram DOSE EQUIVALENT I-131 during all plane condilions except REFUEL.ING SHUTDOWN and COLD SHUTDOWN.

c. Deleted

d. Chemistry (Cl, F and 02) Note I (when Tavg is greater than 20o~FJ.

e. Boron concemration Nute I

f. Tritium Radioactivity Note I

2. Borated Water Boron concentration Note I and after each makeup \Vhen reaclur cuolam Storage Tank system pressure is greater than 300 psig ur Tavg is grcalcr Water Sample than 200°F.

.3. Corl! Flooding Tank Boron concentration Nott: I and aiter each makeup when RCS pressure is Water Sample greater than 700 psig.

TABLE 4.1-3 Cont'd

)> Check Frequency 3

CD

l 4. Spent Fuel Pool Boron Concentration greater than Note 1 a.

3 Water Sample or equal to 600 ppmb CD

~

z 5. Secondary Coolant Isotopic analysis for DOSE Note 1 (when reactor coolant system

!=) EQUIVALENT 1-131 concentration pressure is greater than 300 psig or Tav is greater than 200°F.

6. Deleted

7. Deleted

a. Deleted

9. Deleted

10. Deleted i>'

0 11. Deleted

12. Deleted

Until the specific activity of the primary coolant system is restored within its limits.

Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

.. Deleted

- - Deleted Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

~

'fi;i}

TABLE 4.1-4 POST ACCIDENT MONITORING INSTRUMENTATION

)>

3 FUNCTION INSTRUMENTS CHECK(a) TEST(a) CALIBRATE(a} REMARKS (l)

~

C.

3 1 Noble Gas Effluent (l) 3.

z a. Condenser Vacuum Pump p Exhaust (RM-A5-Hi) (1) Using the installed check source when background is less than twice the expected increase in cpm which would result from the check source-alone.

Background readings greater than this value are sufficient in themselves to show that this monitor is functioning.

b. Condenser Vacuum Pump (1)

Exhaust (AM-G25)

c. Auxiliary and Fuel Handling Building Exhaust (RM-AB-Hi)

~

I d. Reactor Building Purge 0 Exhaust (AM-A9-Hi)

Ill

e. Reactor Building Purge ( 1)

Exhaust (AM-G24)

f. Main Steam Lines (1)

Radiation (RM-G26IAM-G27)

2. Containment High Range Radiation (AM-G22/G23)

3. Containment Pressure NIA

4. Containment Water Level NIA

5. DELETED

6. Wide Range Neutron Flux N/A

TABLE 4.1-4 (Continued)

~ POST ACCIDENT MONITORING INSTRUMENTATION 3

(I)

] FUNCTION INSTRUMENTS CHECK(a) TEST(a) CALIBRATE(a) REMARKS a.

3 (I)

3. 7. Reactor Coolant System Cold Leg NIA zp Water Temperature (TE-959, 961; Tl-959A, 961A)

a. Reactor Coolant System Hot Leg NIA (TE-958, 960; Tl-958A, 960A)

9. Reactor Coolant System Pressure NIA (PT-949, 963; Pl-949A, 963)

N

10. Steam Generator Pressure NIA

....... (PT-950, 951, 1180, 1184;

~

Pl-950A, 951A, 1180, 1184)

~

0 I

11. Condensate Storage Tank Water NIA CJ" Level (LT-1060, 1061, 1062, 1063; Ll-1060, 1061, 1062, 1063)

(a) Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

TABLE 4.1-5 SYSTEM SURVEILLANCE REQUIREMENTS Item Test Frequency

1. Core Flood Tank a. Verify two core flood tanks Note 1 each contain* 940 +/- 30 ft3 borated water.

b. Verify that two core flood Note 1 tanks each contain 600 +/- 25 psig.

I c. Verify CF-V-1A&B are fully open. - Note 1

d. Verify power is removed from Note 1 CF-V-1A&B and CF-V-3A&B valve operators

2. Reactor Building a. Verify the TSP baskets Note 1 Emergency Sump contain 2: 18,815 lbs and pH Control s 28,840 lbs of TSP.

System

b. Verify that a sample from Note 1 the TSP baskets provides adequate pH adjustment of borated water.

Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

4:.1oc Amend ment No. 22a, 2eJ, 274

4.2 REACTOR COOLAN T SYSTEM INSERVICE AND TESTING Applicability This technical specification. applies to the inservice inspection (ISi) of the reactor coolant system pressure boundary and portions of other safety _oriented system pressure boundar ies.

Objective The objective of the ISi program is to provide assurance of the continuing integrity of the reactor coolant system while at the same time minimizing radiation exposure to personne l in the performance of inservice inspections.

Specification 4.2.1 ISi of ASME Code Class 1, Class 2, and Class 3 components shall be performe d in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicab le Addenda as required by 10 CFR 50, Section 50.55a, except where specific written relief has been granted by the NRG.

4.2.2 DELETED.

4.2.3 (Deleted) 4.2.4 The accessible portions of one reactor coolant pump motor flywheel assembly will be ultrasonically inspected within the first ISi period, two reactor coolant pump motor flywheel assemblies within the first two ISi periods and all four by the end of the 1O year inspection interval. However, the U.T. procedure is developmental and will be used only to the extend that it is shown to be meaningful. The extent of coverage will be limited to those areas of the flywheel which are accessible without motor disassem bly, i.e., can be reached through the access ports. Also, if radiation levels at the lower access ports are prohibitive, only the upper access ports will be used.

4-11 Amendment No. i 5, 29, 54, eO, 7"1, i 18, 172, 266, 290

4.2.5 (Deleted) 4.2.6 (Deleted}

4.2.7 A surve illanc e program for the press ure isola tion valves betwe en the primary coolant system .and the 1Dw press ure injec tion system shall as follows: be

1. Periodic leakage testin g(a) at test diffe renti al pressure great er than 150 psid shall be accomplished .for the valves listed in Table 3.1.6 .1 for the following condi tions:

(a) prior to achieving hot shutdown after return ing the valve to servic e following maintenance repai r or replacement work, and (b) prior to achieving hot shutdown following a cold shutdown of great er than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months duration u~less testin g has been

perfonned within the previous 9 months.

2. Whenever integ .rity of a pressure isola tion valve listed 3.1.6 .1 cannot be demonstrated, the integ rity of the otherin Table remaining valve in each high press ure-li ne having a leaking valve shall be determined and recorded daily . In addit ion,

of one other valve located in the high pressure piping the positi on shall t-e recorded daily . *

(a)

To satis fy ALARA requirernents, leakage may be measured indir ectly 0

from the perfonr.ance of pressure indic ators ) if accomplished in (as accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve complianc£ with the leakage crite ria.

Amendment No. !9, !~, ~~. 01"de, dtd. 4/Z07 Sl, X, C'1FF. Lt1. dtd. 1172781., 118 4-12

Specifications 4.2.1 and 2 ensure that inser vice inspection of ASME Code Class 1, 2 and components and inservica testing of ASME 3 Code Class 1, 2 and 3 pumps and valve s will performed in accordance with a periodically be updated version of the ASME Code and Adde as required by 10 CFR 50.55a. Relief from nda any of the above requirements has been provi in writing by the NRC and is not a part of these ded techn 4.0.2 are only applicable to those SRs that reference ical specifications. The provisions of SR usage of the INSERVICE TESTING PROGRAM.

4.3 DELETED

~

"(ffjl 4-13 (Pages 4-14 through 4-28 deleted)

Ame ndm ent No. 29, 54, 60, Order dtd, 4/2Q f81, 71, Gorr. Ur. dtd. 11 /2/81, ReiS6YOd 3/2Q/8a,110,1a7,172,198, 266, 290

4.4 REACTOR BUILDING 4.4.1 CONTAINMENT LEAKAGE TESTS Applicability Applies to containment leakage.

Objective To verify that leakage from the Reactor Building is maintained within allowable limits.

Specification 4.4.1.1 Integrated Leakage Rate Testing (ILRT) shall be conducted in accordance with the Reactor Building Leakage Rate Testing Program at test frequencies established in accordance with the Reactor Building Leakage Rate Testing Program.

4.4.1.2 Local Leakage Rate Testing (LLRT) shall be conducted in accordance with the Reactor Building Leakage RateTesting Program. LLRT shall be performed at a pressure not less than peak accident pressure Pac with the exception that the airlock door seal tests shall normally be performed at 10.psig and the periodic containment airlock tests shall be performed at a pressure not less than Pac. LLRT frequencies shall be in accordance with the Reactor Building Leakage Rate Testing Program.

4.4.1.3 Operability of the personnel and emergency air lock door interlocks and the associat ed control room annunciator circuits shall be determined at the frequency specified in the Surveillance Frequency Control Program. If the interlock permits both doors to be open at the same time or does not provide accurate status indication in the control room, the interlock shall be declared inoperable, except as provided in Technical Specification Section 3.8.6.

Bases c1>

The Reactor Building is designed to limit the leakage rate to 0.1 percent by weight of contained atmosphere in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the design internal pressure of 55 psig with a coincident tempera ture of 281°F at accident conditions. The peak calculated Reactor Building pressure for the design basis loss of coolant accident, Pac, is 50.6 psig. 50.6 psig is a historical value.* The current design basis loss of coolant accident peak reactor building pressure is less than 50.6 psig (Reference 5). The maximum allowable Reactor Building leakage rate, La, shall be 0.1 weight percent of containment atmosphere per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at Pac*

Containment Isolation Valves are addressed in the UFSAR (Reference 2).

4-29 Amendment No. 63, 167, 2Q1, 236, eCR TM QQ Q0703, 274

/

4.4 REACTOR BUILDING {Continued}

The Reactor Building will be periodically leakage tested in accordance with the Reactor Building Leakag e Rate Testing Program (See Section 6.8.5). This program is contained in the surveillance procedures for Reactor Building inspection, Integrated Leak Rate Testing, and Local Leak Rate Testing. These periodic testing requirements verify that Reactor Building leakage rate does not exceed the assumptions used in the safety analysis. At s1 .0 La the offsite dose consequences are bounded by the assumptions of the safety analysis.

During the first unit startup following testing in accordance with this program

, the leakage rate acceptance criteria are s 0,60 La for the combined Type B and Type C leakage, and s O. 75 La for overall Type A leakage. At all other times between required leakage tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 La.

Periodic surveillance of the airlock interlock systems (Reference 4) assures continued operability and precludes instances where one or both doors are inadver tently left open.

When an airlock is inoperable and containment integrity is required, local supervision of airlock operation is specified.

References (1) UFSAR, Chapte r 5.7.4- "Post Operational Leakage Rate Tests" (2) UFSAR, Tables 5.7-1 and 5.7-3 (3) DELET ED (4) UFSAR, Table 5.7-2 .

(5) C-1101-823-5450-001 TMl-1 LBLOCA EQ Temperature Profile Using Gothic Computer Code."

~

'(!:lD 4-30 (Pages 4-31 through 4-34, 4-34a, and 4-34b deleted)

@ Amendm ent No. 27,167 ,201, ECRTM 09 00703, 274

CONTROLLED COPY 4.4.2 Structural Integrity Specification 4.4.2.1 lnservice Tendon Surveillance Requirements The surveillance program for structural integrity and corrosion protection conforms to the requirements of Subsection IWL of Section XI of the ASME Boiler and Pressu re Vessel Code, as incorporated by reference into 10 CFR 50.55a. The detailed surveill ance program for the prestressing system tendons shall be based on periodic inspection and mechanical tests to be performed on selected tendons.

4.4.2.1.1 DELETED

£31~)1',,,,,.i 4-35

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Amendment No. 59, 95, 129, 157, 187, 251

4.4.2 .1.2 DELETED 4.4.2 .1.3 DELETED 4.4.2 .1.4 Tendon Surveillance Previous Inspections The tendon surveillance shall include the reexamination of all abnormalities (i.e. , concrete scali ng, cracking, grease leaka etc.} discovered in the previous inspection to ge, determine wheth conditions have stabi lized . The inspection program shall be er modified accordingly if obvious deter iorat ing conditions are

observed.

4.4.2 .1.5 Inspection for Crack Growth at Dome Tendons in the Anchorage Areas Ring Girder Concrete around the dome tendon anchorage areas shall be inspe for crack growth during ten and 15 year inspections by moni cted cracks grea ter than 0.005 inch in width. Selec t as a minimumtoring dome tendon anchoring areas having concrete cracks with crack nine widths 0.005 inch. In the selec tion of dome tendon anchoring .

to be monitored, preference shall be given to those areas havin areas

cracks grea ter than 0.005 inch in width. g depths can be measured with simple exist ingTheplant width, depth (if instruments, (i.e. ~ feele r gauges, wires) and length of the selec shall be measured and mapped by char ting. This inspeted cracks ction discontinued, if the concrete cracks show no sign of growth.may be however, these inspections indicate crack growth, an inves tigatIf, of the causes and safety impact should be performed. ion

'@*.,,. )

,,,.,v 4-36 Amendment No. ,,, 1i,, 187

CONTROLLED COPY t 11: ' ) 4.4.2.1.6 Reports*

a. Within 3 months after the completion of each tendon surveillance a special report shall be submitted to the NRG Region I Administrator. This Report will include a section dealing with trends for the rate of prestress loss as compared to the predicted rate for the duration of the plant life (after an adequate number of surveillances have been completed).

b. Reports submitted in accordance with :10 CFR 50.73 shall include a descriptiqn of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures , the tolerances on cracking, and any corrective actions taken.

4.4.3 DELETED BASES For ungrouted, post-tensioned tendons, this surveillance requirement ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the TMl-1 Reactor Building Structural Integrity Tendon Surveillance Program. Testing and frequency are consistent with the requirements of Subsection IWL of Section XI of the ASME Boiler and Pressure Vessel Code, as incorporated by reference into 1o CFR 50.55a, and as described in

~eFSAR. -

The modified visual inspection requirements pertaining to the dome tendons in the ring girder were implemented as a result of: 1) discovery of ring girder voids in 1977 and the potential that more undetected voids in the ring girder could exist, and 2) the number of dome tendon bearing areas having cracks appeared to be growing with time (Reference Amendme nt No. 59). -

REFERENCES (1) UFSAR, Section 5.7.5 - Tendon Stress Surveillances 4-37 Amendmen t No. 108, 129, 158, 187, 251

CONTROLLED COPY 4-38 (Page 4-38a deleted)

Amendment No. 87, +§8, 4-7§, 98, ~ . 24-0, 246

4.5 EMERGE NCY LOADING SEQUEN CE AND POWER TRANSF ER, EMERGE NCY CORE COOLING SYSTEM & REACTO R BUILDIN G COOLING SYSTEM PERIODI C TESTING 4.5.1 Emergen cy Loading Sequenc e Applicability: Applies to periodic testing requirements for safety actuation systems.

Objective : To verify that the emergen cy loading sequence and automatic power transfer is operable.

Specifica tions:

4.5.1.1 Sequenc e and Power Transfer Test

a. At the frequenc y specified in the Surveillance Frequenc y Control Program, a test shall be conducte d to demonstr ate that the emergen cy loading sequence and power transfer is operable.

b. The test will be considere d satisfactory if the permane ntly conn~cte d loads and auto-con nected emergen cy loads have been successf ully energized on preferred power using the load sequence r and transferred to emergen cy power.

C. Following successful transfer to the emergency diesel, the diesel generator breaker will be opened to simulate trip of the generato r then re-closed to verify block load on the reclosure.

4;5.1.2 Sequenc e Test

a. At the frequenc y specified in the Surveillance Frequenc y Control Program, a test shall be conducte d to demonstr ate that the emergen cy loading sequence is operable, this test shall be performe d on either preferred power or emergen cy power.

b. The test will be considere d satisfactory if the auto-con nected emergen cy loads have been successf ully energized using the load sequence r.

4-39 Amendm ent No. 70, 78, 149,167 ,212,274 ,276

Bases The Emergency loading sequence and automatic power transfer controls the operation of the pumps associated with the emergency core cooling system and Reactor Bui!ding cooling system.

The requirement to verify the connection and power supply of permanent and auto connected loads (Reference 1) is intended to satisfactorily show the relationship of these loads to the Emergency Diesel Generator (EOG) loading logic. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation. For instance, Emergency Core Cooling Systems (ECCS) injection valves are not desired to be stroked open, high pressure injection systems are not capable of being operated at full flow, or decay heat removal (OHR) systems performing a OHR function are not desired to be realigned to the ECCS mode of operation. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the EOG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.

Automatic start and loading of the emergency diesel generator to meet the requirements of 4.5.1.1 b/c above is described in Technical Specification 4.6.1.b.

Reference (1) UFSAR, Table 8.2-11, "Engineered Safe~uards Loading Sequence" 4-40 Amendment No. 70, 149, 167, ECR TM 13-00306

4.5.2 EMERGENCY CORE COOLING SYSTEM Applicability: Applies to periodic testing requirement for emergency core cooling systems.

Objective: To verify that the emergency core cooling systems are operable.

Specification 4.5.2.1 High Pressure Injection

a. At the frequency specified in the Surveillance Frequency Control Program and following maintenance or modification that affects system flow characteristics, system pumps and system high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable.

b. The test will be considered satisfactory if the valves (MU-V-14A/B

& 16A/B/C/D) have completed their travel and the make-up pumps are running as evidenced by system flow. Minimum acceptable injection flow must be greater than or equal to 431 gpm per HPI pump when pump discharge pressure is 600 psig or greater (the pressure between the pump and flow limiting device) and when the RCS pressure is equal to or less than 600 psig.

c. Testing which requires HPI flow thru MU-V16A/B/C/D shall be conducted only under either of the following conditions:

~ 1) Indicated RCS temperature shall be greater than 313°F.

vfi1I 2) Head of the Reactor Vessel shall be removed.

d. At the frequency specified in the Surveillance Frequency Control Program, verify High Pressure Injection locations susceptible to gas accumulation are sufficiently filled with water.

4.5.2.2 Low Pressure Injection

a. At the frequency specified in the Surveillance Frequency Control Program and following maintenance or modification that affects system flow characteristics, system pumps and high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable. The auxiliaries required for low pressure injection are all included in the emergency loading sequence test specified in 4.5.1.

b. The test will be considered satisfactory if the decay heat pumps have been successfully started and the decay heat injection valves and the decay heat supply valves have completed their travel as evidenced by the control board component operating lights. Flow shall be verified to be equal to or greater than the flow assumed in the Safety Analysis for the single corresponding RCS pressure used in the test.

0 .

4-41 Amendm ent No. 49,57,a 8,449,2 Q3,225 ,234,27 4,~.~. 285

c. When the Decay Heat System is required to be operable, the correct position of DH-V-19A/B shall be verified by observation within four hours of each valve stroking operation or valve maintenance which affects the position indicator.

d. At the frequency specified in the Surveillance Frequency Control Program, verify Low Pressure Injection locations susceptible to gas accumulation are sufficiently filled with water.

4.5.2.3 Core Flooding

a. At the frequency specified in the Surveillance Frequency Control Program, a system test shall be conducted to demonstrate proper operation of the system. Verification shall be made that the check and isolation valves in the core cooling flooding tank discharge lines operate properly.

b. The test will be considered satisfactory if control board indication of core flooding tank level verifies that all valves have opened.

c. At the frequency specified in the Surveillance Frequency Control Program, verify Core Flooding locations susceptible to gas accumulation are sufficiently filled with water.

4.5.2.4 Component Tests

a. At the frequency specified in the Surveillance Frequency Control Program, the components required for emergency core cooling will be tested.

b. The test will be considered satisfactory if the pumps and fans have been successfully started and the valves have completed their travel as evidenced by the control board component operating lights, and a second means of verification, such as: the station computer, verification of pressure/flow, or control board indicating lights initiated by separate limit switch contacts.

Bases The emergency core cooling systems (Reference 1) are the principal reactor safety features in the event of a loss of coolant accident. The removal of heat from the core provided by these systems is designed to limit core damage.

The low pressure injection pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the bypass valves in the borated water storage tank fill line. This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank.

The minimum acceptable HPI/LPI flow assures proper flow and flow split between injection legs.

With the reactor shutdown, the valves in each core flooding line are checked for operability by reducing the reactor coolant system pressure until. the indicated level in the core flood tanks verify that the check and isolation valves have opened.

ECCS piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the ECCS and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.

4-42 Amendment No . .§+,.a8,449,4a-7,4a7,~.274. 285

Bases (Continued)

Selection of ECCS locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

With regard to 4.5.2.1.d, 4.5.2.2.d, and 4.5.2.3.c, the ECCS is OPERABLE when it is sufficiently filled with water. Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump), the Surveillance is not met. If it is determined by subsequent evaluation that the ECCS is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.

ECCS locations susceptible to gas accumulation are monitored and, if gas is found, the gas volu.me is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety.

For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

Reference (1) UFSAR, Section 6.1 - "Emergency Core Cooling System" i

i

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l 4-42a Amendm ent No. 57,08,4 49,4&7, 437,~.~ . 285

4.5.3 REACT OR BUILD ING COOLING AND ISOLATION SYSTEM Applicability Applies to testig of the reactor building cooling and isolation systems ..

Objective To verify that the reactor building cooling system s are operable.

Specification 4.5.3.1 System Tests

a. Reacto r Building Spray System

1. At the frequency specified in the Surveillance Freque ncy Control Progra m

and simultaneously with the test of the emerge ncy loading sequence, a

Reacto r Building 30 psi high pressure test signal will start the spray pump. Except for the spray pump suction valves, all engineered safeguards spray valves will be closed.

Water will be circulated from the borated water storage tank through the reactor building spray pumps and returned through the test line to the borated water storage tank.

The operation of the spray valves will be verified during the compo nent test of the R. 8. Cooling and Isolation System.

The test will be considered satisfactory if the spray pumps have been successfully started.

2. Compressed air will be introduced into the spray header s to verify each spray nozzle is unobstructed at the frequency specified in the Surveillance Frequency Control Program.

3. At the frequency specified in the Surveillance Freque ncy Control Program, verify Reactor Building Spray locations susceptible to gas accumulation are sufficiently filled with water.

b. Reacto r Building Cooling and Isolation Systems*

1. At the frequency specified in the Surveillance Freque ncy Control Program, a system test shall be conducted to demon strate proper operation of the system.

2. The test will be considered satisfactory if measured system flow is greater than accident design flow rate.

4-43 Amend ment No. 4&7,49 8,-24-2 ,~,-274 ,~, 285

4.5.3.2 Component Tests

a. At the frequency specified in the Surveillance Frequency Control Program, the components required for Reactor Building Cooling and Isolation will be tested.

b. The test will be considered satisfactory if the valves have completed their expected travel as evidenced by the control board component operating lights, and a second means of verification, such as: the station computer, local verification, verification of pressure/flow, or control board component operating lights initiated by separate limit switch contacts.

Bases The Reactor Building Cooling and Isolation Systems and Reactor Building Spray System are designed to remove the heat in the containment atmosphere to prevent the building pressure from exceeding the design pressure (References 1 and 2).

The delivery capability of one Reactor Building Spray Pump at a time can be tested by opening the valve in the line from the borated water storage tank, opening the corresponding valve in the test line, and starting the corresponding pump.

With the pumps shut down and the Borated Water Storage Tank outlet valve closed, the Reactor Building spray injection valves can each be opened and closed by the operator action. With the Reactor Building spray inlet valves closed, low pressure air can be blown through the test connections of the Reactor Building spray nozzles to demonstrate that the flow paths are open.

Reactor Building Spray System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the required Reactor Building Spray trains and may also prevent water hammer and pump cavitation.

Selection of Reactor Building Spray System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

With regard to 4.5.3.1.a.3 the Reactor Building Spray System is OPERABLE when it is sufficiently filled with water. Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of 4-44 Amendmen t No . .gg,449,4-e+,4&7,-2-74, 285

Bases

. accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump), the Surveillance is not met. If it is determined by subsequent evaluation that the Reactor Building Spray System is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated ga$ should be eliminated or brought within the acceptance criteria limits.

Reactor Building Spray System locations susceptible to gas* accumul~tion are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are

subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

The equipment, piping, valves and instrumentation of the Reactor Building Cooling System are arranged so that they can be visually inspected. The cooling units and associated piping are loc~ted outside the secondary concrete shield.

Personnel can enter the Reactor Building during power operations to inspect and maintain this equipment.

The Reactor Building fans are normally operating periodically, constituting the test that these fans are operable.

Reference (1) UFSAR, Section 6.2 - "Reactor Building Spray System" (2) UFSAR, Section 6.3 - "Reactor Building Emergency Cooling System" 4-44a Amendment No. ~.449,45 7,4e-7,27 4, 285

4.5.4 ENGINEERED SAFEGUARDS FEATURE (ESF) SYSTEMS LEAKAGE Applicability Applies to those portions of the Decay Heat, Building Spray, and Make-U p Systems, which are required to contain post accident sump recirculation fluid, when these system operable in accordance with Technical Specification 3.3. s are required to be Objective

To maintain a low leakage rate from the ESF systems in order to preven t significant off-site exposures and dose.consequences.

Specification 4.5.4.1 The total maximum allowable leakage into the Auxiliary Building from the applicable portions of the Decay Heat, Building Spray and Make-Up System components as measured during tests in Specification 4.5.4.2 shall not exceed 15 gallons per hour.

4.5.4.2 At the frequency specified in the Surveillance Frequency Control Program the following tests of the applicable portions of the Decay Heat Removal, Buildin g

Spray and Make-Up Systems shall be conducted to determine leakage

a. The applicable portion of the Decay Heat Removal System that is outside containment shall be leak tested with the Decay Heat pump operating, except as specified in "b".

b. Piping from the Reactor *Building Sump to the Building Spray pump and Decay Heat Removal System pump suction isolation valves shall be pressu re tested at no less than 55 psig.

C. The applicable portion of the Building Spray system that is outside containment shall be leak tested with the Building Spray pumps operati ng l,, and BS-V-1 NB closed, except as specified in "b" above.

d. The applicable portion of the Make-Up system on the suction side of the Make-Up pumps shall be leak tested with a Decay Heat pump operating and DH-V-7 NB open.

e. The applicable portion of the Make-Up system from the Make-Up pumps to the containment boundary valves (MU-V -16ND , .. 18, and 20) shall be leak tested with a Make-Up pump operating.

f. Visual inspection shall be made for leakage from components of these systems. Leakage shall be measured by collection and weighing or by another equivalent method.

Bases The leakage rate limit of 15 gph (measured in standard room temperature gallons) for the accide nt recirculation portions of the Decay Heat Removal (OHR), Buildin g Spray (BS), and Make-U p (MU) systems is based on ensuring that potential leakage after a loss-of-coolant accide nt will not result in off-site dose consequences in excess of those calculated to comply with the 10 CFR 50.67 limits (Reference 1 and 2). The test methods prescri bed in 4.5.4.2 above for the applicable portions of the DH, BS and MU systems ensure that the testing results accoun t for the highest pressure within that system during the sump recirculation phase of a design basis accident.

References (1) UFSAR, Section 6.4.4 - "Design Basis Leakage" (2) UFSAR, Section 14.2.2.5(d) - "Effects of Engineered Safeguards Leakag e During Maximum Hypothetical Accident" 4-45 Amend ment No. 1137, 20a, 216, Corrected by lotter dated: 9124/QQ, 23&,

274

4.6 EMERGENCY POWE R SYSTEM PERIODIC TESTS Applicability: Applies to periodic testing and surveillance requirement of _the emerge ncy power

~ system.

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Objective: 'Ta verify that the emergency power system will respond promptly and proper ly when required.

Specification:

The following tests and surveillance shall be performed as stated:

4.6.1 Diesel Generators

a. Manually-initiate start of the diesel generator, followed _by manual synchronization with other power source s and assumption of load by the diesel genera tor up to the name-plate rating (3000 kw). This test will be conduc ted at the frequen cy specified in the Surveillance Frequency Control Program on each diesel generator. Normal plant operation will not be effected.

b. Automatically start and loading the emergency diesel generator in accord ance with Specification 4.5.1.1.b/c including the following. This test will be conduc ted at the frequen cy specified in the Surveillance Frequency Control Progra m on each diesel generator.

(1) Verify that the diesel generator starts from ambien t condition upon receipt of the ES signal and is ready to load in s 1O seconds.

(2) Verify that the diesel block loads upon simulated loss of offsite power 30 seconds. in s (3) The diesel operates with the permanently connected and auto connec ted load for ~ 5 minutes.

(4) The diesel engin~ does not trip when the generator breaker is opened while carrying emergency loads.

(5) The diesel genera tor block loads and operates for ~ 5 minutes upon reclosure of the diesel generator breaker.

c. Deleted.

4.6.2 Station Batteries

a. The voltage, specific gravity, and liquid level of each cell will be measu red and recorded:

(1) at the frequen cy specified in the Surveillance Frequency Control Progra (2) m once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery discharge < 105 V (3) once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery overcharge > 150 V (4) If any cell parameters are not met, measure and record the parame ters on each connected cell every 7 days thereafter until all battery parameters are met.

b. The voltage and specific gravity of a pilot cell will be measured and recorde d at the frequency specified in the Surveillance Frequency Control Program.

If any pilot cell parameters are not met, perform surveillance 4.6.2.a on each connec ted cell within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and every 7 days thereafter until all battery parame ters are met.

c. Each time data is recorded, new data shall be compared with old to detect signs of abuse or deterioration.

4-46 Amend ment No.70, 149, 167, 200, 232, 243, 274

d. The battery will be subjected to a load test at the frequency specified in the Surveillance Frequency Control Program.

(1) Verify battery capacity exceeds that required to meet design loads.

(2) Any battery which is demonstr ated to have less than 85% of manufacturers ratings during a capacity discharge test shall be replaced during the subseque nt refueling outage.

4.6.3 Pressuriz er Heaters

a. The following tests shall be conducted at the frequenc y specified in the Surveillance Frequency Control Program:

(1) Pressurizer heater groups 8 and 9 shall be transferred from the normal power bus to the emergen cy power bus and energized. Upon completion of this test, the heaters shall be returned to their normal power bus.

(2) Demonstrate that the pressuriz er heaters breaker on the emergen cy bus cannot be closed until the safeguards signal is bypassed and can be closed following bypass.

(3) Verify that following input of the Engineered Safeguar ds Signal, the circuit breakers, supplying power to the manually transferred a

loads for pressuriz er heater groups and 9, have been tripped.

Bases The tests specified are designed to demonstr ate that one diesel generato r will provide power for operation of safeguar ds equipment. They also assure that the emergen cy generato r control system and the control systems for the safeguards equipme nt will function automatically in the event of a loss of normal a-c station service power or upon the receipt of an engineered safeguards Actuation

Signal. The intent of the periodic tests is to demonstr ate the diesel capability to carry design basis accident (LOOP/LOCA) load. The test should not exceed the diesel 2000-hr. rating of 3000 kW. The automatic tripping of manually transferre d loads, on an Engineer ed Safeguards Actuation Signal, protects the diesel generato rs from a potential overload condition. The testing frequenc y specified is intended to identify and permit correction of any mechanic al or electrical deficiency before it can result in a system failure. The fuel oil supply, starting circuits, and controls are continuou sly monitored and any faults are alarmed and indicated. An abnorma l condition in these systems would be signaled without having to place the diesel generato rs on test.

Precipito us failure of the station battery is extremely unlikely. The Surveillance Frequenc ies are controlled under the Surveillance Frequency Control Program.

The PORV has a remotely operated block valve to provide a positive shutoff capability should the relief valve become inoperable. The electrical power for both the relief valve and the block valve is supplied from an ESF power source to ensure the ability to seal this possible RCS leakage path.

The requirem ent that a minimum of 107 kw of pressurizer heaters and their associate d controls be capable of being supplied electrical power from an emergen cy bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation.

4-47 Amendm ent No. 78, 157, 187, 176, ECR TM 07 00119, 274

4.7 REACTOR CONTROL ROD SYSTEM TESTS 4.7.1 CONTRO L ROD DRIVE SYSTEM FUNCTIONAL TESTS Applicability Applies to the surveillance of the control rod system.

Objective To assure operability of the control rod system.

Specification 4.7.1.1 The control rod trip insertion time shall be measured for each control rod at either full flow or no flow conditions following each refueling outage prior to return to power. The maximum control rod trip insertion time for an operable control rod drive mechanism from the fully withdraw n position to % insertion (104 inches travel) shall not exceed 1.66 seconds at hot reactor coolant full flow conditions or 1.40 seconds for the hot no flow conditions (Reference 1). If the trip insertion time above is not met, the rod shall be declared inoperable.

4.7.1.2 If a control rod is misaligned with its group average by more than an indicated nine inches, the rod shall be declared inoperable and the limits of Specification 3.5.2.2 shall apply. The rod with the greatest misalignment shall be evaluated first. The position of a rod declared inoperab le due to misalignment shall not be included in computing the average position of the group for determin ing the operability of rods with lesser misalignments.

4.7.1.3 If a control rod cannot be exercised, or if it cannot be located with absolute or relative position

~ indications, in or out limit indication, or zone reference switch indication, the rod

~a declared to be inoperable.

shall be The control rod trip insertion time is the total elapsed time from power interruption at the control rod drive breakers until the control rod has actuated the 25% withdrawn reference switch during insertion from the fully withdrawn position. The specified trip time is based upon the safety analysis in UFSAR, Chapter 14 and the Accident Parameters as specified therein.

Each control rod drive mechanism shall be exercised by a moveme nt of a minimum of 3%

o~ travel at the frequenc y specified in the Surveillance Frequency Control Program

. This requirement shall apply to either a partial or fully withdrawn control rod at reactor operating conditions. Exercising the drive mechanisms in this manner provides assurance of reliability of the mechanisms.

4-48 Amendm ent No. 157, 211, 273, 274

CONTROLLID a:JPY

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. A ,*od is considered inoperable if ins ert ion time is gre ate r than the it cannot be exe rcis ed, if the tri p the rod dev iate s from its group ave spe cif ied allowable time, or if inc hes . Conditions for operation rage pos itio n by more than nine spe cif ied in Technical Sp eci fic atiowith an inoperable rod are n 3.5 .2.

~EFERENCE

{1) UFSAR, Section 3.1 .2. 4.3 - "Co ntrol Rod.Drive Mechanism" J

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4-49 Amendment No. 157

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4. 7.2 CONTROL ROD PROGRAM VERIFICATION (Group vs. Core Positions)

The page intentionally left blank 4-50 Amendment No. ~ . 21 l

CONTROllrED COPY r

4-51 Amendment No. 9a, .:J.e7, 246

4.9 DECAY HEAT REMOVAL (OHR) CAPABILITY- PERIODIC TESTING Applicability Applies to the periodic testing of systems or components which function to remove decay heat.

Objective To verify that systems/components required for DHR are function. capable of performing their design Specification 4.9.1 Reactor Coolant System (RCS) Temperature greater than 250 degrees F.

4.9.1.1 Verify each Emergency Feedwater (EFW) Pump is tested in accordance with the

requirements and acceptance criteria of the INSERVICE TESTING PROGRAM.

Note: This surveillance is not required to be performed for the turbine-driven EFW Pump (EF-P-1) until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding 750 psig.

4.9.1.2 DELETED 4.9.1.3 At the frequency specified in the Surveillance Frequency Contro System fl,owpath yalve from both Condensate Storage Tanks l Pegram, each EFW via the motor-driven pumps and the turbine-driven pump (CSTs) to the OTSGs shall be verified to be in the required status.

4.9. 1.4 At the frequency specified in the Surveillance Frequency Control ~rogram:

~

W!H a) Verify that each EFW Pump starts automatically upon receipt of an EFW test signal.

b) Verify that each EFW control valve responds upon receip t of an EFW test signal.

c) Verify that each EFW control valve responds in manua l control from the control room and remote shutdown panel.

4.9.1.5 Prior to STARTUP, following a REFUELING SHUTDOWN or a COLD SHUTDOWN greater than 30 days, conduct a test to demonstrate that the motor driven EFW Pumps can pump water from the CSTs to the Steam Gener ators.

4-52 Amen dmen t No. 70,11 9,124 ,172, 242,2 66, 2-74,290

4.9 DECAY HEAT REMOVAL (OHR} CAPABILITY-PERIODIC TESTING (Continued}

4.9.1.6 Acceptance Criteria These tests shall be considered satisfactory if controf board indication and visual observation of the equipment demonstrates that all components have operated properly except for the tests required by Specification 4.9.1.1.

4.9.2 RCS Temperature less than or equal to 250 degrees F.*

4.9.2.1 At the frequency specified in the Surveillance Frequency Control Program, verify operability of the means for OHR required by Specification 3.4.2 by observation of console status indication.

~----------------------------1\JOTE-----------------------------~---------------

Entry into 4.9.2.2. below is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature is less than or equal to 250 degrees F.

4.9.2.2 At the frequency specified in the Surveillance Frequency Control Program, verify required OHR loop locations susceptible to gas accumulation are sufficiently filled with water. *

These requirements supplement the requirements of Specifications 4.5.2.2 and 4.5.4.

Bases The ASME Code specifies requirements and acceptance standards for the testing of nuclear

~ safety related pumps. The EFW Pump test frequency specified by the ASME Code will be

~ sufficient to verify that the turbine-driven and both motor-driven EFW Pumps are operable.

Compliance with the normal acceptance criteria assures that the EFW Pumps are operating as expected. The surveillance requirements ensure that the overall EFW System functipn al capability is maintained.

Deferral of the requirement to perform 1ST on the turbine-driven EFW Pump is necessa ry to assure sufficient OTSG pressure to perform the test using Main Steam.

Periodic verification of the operability of the required means for OHR ensures that sufficient OHR capability will be maintained.

OHR System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessa ry for proper operation of the required OHR loop(s) and may also prevent water hammer

, pump cavitation, and pumping of noncondensible gas into the reactor vessel.

Selection of OHR System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometri c

drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the*

4-52a Amendm ent I\Jo . .:ra,4-=t-9,424,4-72,242,200,274, 285

4.9 DECAY HEAT REMOVAL {OHR} CAPABILITY-PERIODIC TESTING (Continued) 9

. . Bases (Continued}

location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

With regard to 4.9.2.2 the OHR System is OPERABLE when it is sufficiently filled with water.

Acceptance criteria are established for the volume of accumulated gas at suscept ible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump),

the Surveillance is not met. If it is determined by subsequent evaluation that the OHR System is not rendered inoperable by the accumulatec;f gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminat ed or brought within the acceptance criteria limits.

OHR System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that are inaccessible due to radiolog ical or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determin ed to not challenge system OPERABILITY. The accuracy of the method used for monitori ng the

susceptible locations. and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

SR 4.9.2.2 is modified by a Note that states the SR is not required to be performed until *12 hours after RCS temperature is less than or equal to 250 degrees F. In a rapid shutdow n,

there may be insufficient time to verify all susceptible locations prior to RCS tempera ture reaching less than or equal to 250 degrees F.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

4-52b Amendment No. 285

,f~';,c;~\,

, ~~~:,. ,.'

4.10 REACTIVITY ANOMALIES App licab ility

App lies to pote ntia l reac ti'Vi ~y anom alies Obje ctive To* requ ire the eval uatio n of reac tivi ty anomaJ.ies o:f a spec i:fie d mag nitud e occu rring durin g the oper ation of the unit Spec ifica tion 4.10 .l -Foll owin g a norm aliza tion o:f the comp uted boro n conc entr atio n as a func tion o:f burn up., the actu al boro n conc entr atio n o:f the cool ant shal l be peri odic ally comp ared vith the pred icte d vs.1. ue. If the di:ff eren ce betw een the obse rved and pred icte d stea dy-s tate conc entr ation s reac hes the equi vale nt o:f one perc ent in reac tivi ty, an eval uatio n 'Will be ma.de to dete rmin e the ca.us e of the disc repa ncy.

Base s To elim inate poss ible erro rs in the calc ulat ions of the init ial. reac ti-vi ty of the core and the reac tivit y depl etion rate

., the pred icte d rela tion betw een fuel burn up and the.b oron conc entr atio n, nece ssar y to main tain adeq uate con-trol char acte risti cs, must be adju sted

{ norm alize d) to accu rate ly refl ect actu al core cond ition s. When full powe r is reac hed init iall y, and with the cont rol rod grou ps in*th e desi red posi tion s, the boro n conc entr atio n is meas ured and the pred icted curv e is adju sted to this poin t ~ As powe.r oper a-tion proc eeds , the meas ured boro n conc entr ation is comp ared '\dth the pred icte conc entr ation and the slop e of the curv d e rela ting b11rnup and reac tivi ty is comp ared with that pred icted . Tb.is proc ess of norm aliza tion shou ld be com-plet ed afte r abou t 10 :perc ent of the tota

l. core burn up. The reaf ter, actu al.

boro n conc entr ation can be comp ared vith pred ictio n, and the reac tivi ty stat us of the core can 'be cont inuo usly eval uate

d. Any reac tivi ty anom aly grea ter than one perc ent woul d be unex pecte .d~ and its occu rren ce woul d be thor ough ly inve stiga ted and eval uate d.

The valu e of one perc ent is cons idere d a safe liln it sinc e a shutd own marg in of at leas t one perc ent with the most reac tive rod in the .ful.l y with dra.m posi tion is el ways main taine d.

4-53

4.11 REACTOR COOLANT SYSTEM VENTS Applicability Applies to Reactor Coolant System Vents.

Objective To ensure that Reactor Coolant System vents are able to perform their design function.

Specification 4.11.1 Each reactor coolant system vent path shall be demonstrated OPERABL E at the frequency specified in the Surveillance Frequency Control Program by cycling each power operated valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING.

BASES Tests specified above are necessary to ensure that the individual Reactor Coolant System Vents will perform their functions. It is not advisable to perform these tests during Plant Power Operation, or when there is significant pressure in the Reactor Coolant System. Tests are, therefore, to be performed during either Cold Shutdown or Refueling .

4.54.

Amendmen t No. 9e, Q7, 274

4.12 AIR TRE ATM ENT SYSTEM 4.12.1 EME RGE NCY CON TRO L ROOM AIR TREA TME NT SYST EM Applicability Applies to the emer genc y control room air treatm ent system and associated components.

Objective To verify that this syste m and associated comp onents will be able to perf9rm its design functions ..

Specification 4.12.'1.1 At the frequ ency specified in the Surveillance Freq uency Control Program, the press ure drop across the combined HEP Afilte rs and charco1;1I a:dsorber banks of AH-F 3A and 38 shall be demonstrated to be less system desig n flow rate (+/-10%). than 6 inches of water at 4.12.1.2

a. The tests and sample analysis required by Spec ification 3.15.1.2 shall be performed initially and at the frequency specified in the Surveillance Freq uenc y Control Program for stand by service or after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation and-follo_wirig significant.pain ting*, *steam, fire or chemical .release in any ventilation zone comm unicating with the system that could contaminate the HEPA filters or charc oal adsbrbers.

b. DOP testing *shall be *performed after each comp lete or partial replacement of the HEPA filter bank or after any structural maintenance on the system housing which could affec t the HEP A filter bank bypass leakage.

c. Halogenated hydrocarbon testing __ shall be pE;3rf otmed after each complete or partial replacement of the charcqal adso rber bank or after any structural maintenance on the system hous ing which could effect the charcoal adso rberb ank bypass leakage.

d. Each AH-E1 BA and B. (AH-F3A and B) fan/filter circu for =?: 15 continuous minutes at the frequ ency spec it shall be operated ified in the Surveillance Frequency Control Program.

4.12.1.3 At the frequ ency specified in the Surveillance.Fr equency Control Program, autom atic initiation 'Of t~e required Control Build ing dampen;; for isolation and recirculation shall be demonstrated as operable; 4.12.1.4 An air distribution test s_ha!I be performed on the

.HEPA filter bank in,itially, and after any maintenc:if'!Ce or testing. that could affec t the air distribution*within the syste m .. The a_ir distribution t;1cross the Hl;PA fi.lter

+/-20% . The test shall be performed at 40,000 cfm bank shall be uniform within

(+/-10%) flow rate.

4.12.1.5 Cont rol Room Envelope unfiltered air inleakag(;3 testing shall be performed in acco rdan ce with the Control Room Envelope Habi tability Program.

4-55 Ame ndme nt No.66, 68, 1_49, 176, 223, 264, 274, 282

BASES Pressure drop across the combined HEPA filters and charcqaladsorbers of less than 6 inches of w;3ter at tfie:system des,lgri-flow rate will indic ate that the filters and adsorbers are not clogg~d by excessiv~ amounts of foreign rry~tter.-

Pressure drop should be determined at the frequency specified in the Surveillance Frequenc y Control Program to show system performance capability. * *

The frequency of tests and sample analysis are necessary to show t!lat the.HEPA filters and ct,arcoa(adsor.bers'can perform as evaluated; Tests of the charcoaf adsorbers wi(h -

halogenated *hydrocarbon shall-be perforrhe9 in accor.dan.ce with i;ipprovE;Jd test procedures.

Repla_cemen.t adsorbent should be qualified acco rding to ASTMD~803-1989. Ttie charc~aJ adsorber efficiency test procedwes should allow for_theremov~I of one* ~dsoi:ber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtqiningeat *,east' two samples.

Each sample should be at least two inches in diam eter and a leng~tJ.eqLla:I to the thickness of the bed .. If test results'are unacceptable all adso rbent_in the syste'm $hall be replaced. Tests of the HE;PA filters With..OOp aero~olshall al_S9 be performed in accordance with approved test procedures. Any* HEPA filters found-defective should be replaced with filters qualified according to Regulatory Guide 1:52 Marett .1978.

Operation for~ 15 con~inµous.minutel5 at the frequ ency specified in the Surveillance FreqLJency Control Program demonstrates*OPERABILITY of the system.* PeriOdi.c operation ensures that blockage/fan or motor failure, Or excessive vibra tion can be detected for corrective action.

If significant painting, steam, fire or chemical .relea se occurs such that the HEPA fiiter or charcoal ads.orber couitj become contamjnaied frc:im the fumes, chemicals qr foreign materials, the same tests and sample analysis shaJLbe *perf ormed as req1;1ired for operational use. The determination of significanc~ shall be .made by the Vice President-TM! ;Unit 1.

Dem9ristration of the automatic initiation of.the recirculatio.n mode of operation is necessary to assu.re sy~tem perfqrmarice -c::apability. D~mpers

required for control building isolation and recirci,Jlatiqn are*specified iri UFSAR S"ections 7.4.5 and 9.8.L *

Control Room .Envelope unfiltered air inleakage testing verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary.and in'bt he c:RE. The details of the testing are specified in the Control Room E:'~velope Habitability-.Program.

__ The CR!= is considered habitable when the radio logical dose to CRE occupants calculated in the licensing basis analyses of-OBA consequences is no more than 5 rem TEDE and the CRE occupants are protected from hazardous qhen iicals and smoke. Air inleaka'ge testing veri.fies that the unfiltered air inleakag~ into* the GAE is no greater than the flow rate ass1,Jmed in the licen§ing basis analyses of OBA consequences

. When* unfiltered air in leakage is greater than the assumed-flow rat~. Section 3.15: 1.5 must be entered. The required actions allow time to restore the CRE boundary to OPERABLE statu s provided mitigating actions can* ensure that the CRE remains within the _licensing Qasis habitabili ty limits for the'1occupants 'following an .

accident. Compensatory r11easi.lres are discus,sed in Regulatory .Guide 1.196, Section C.2. 7.3, (Ref. 1) whiph endorses, with exceptions, NEI 99:03, Section-S.4 and Appendix F (Ref. 2).

These compensatory measures may also be used as mitigating actions as req1..,1ired by Section 3.15.1.5. Temporary analytical methods may also be used as compensatory measures to 4-55a Amendmerit No. 56,1 79,2 18,2 23,2 26,2 64,2 74,2 82

restore OPERABILITY (Ref. 3). Options for restoring the CRE boundary to OPERAB LE status include changing the licensing basis OBA consequence analysi~. repairing the CRE boun_dary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status.

References

1. Regulatory Guide 1.196.

2. NEI 99-03, "Control Room Habitability Assessment Guidance", June 2001.

3. Letter from Eric J. Leeds (NRC} to Jarn*es W. Davis (NEI) dated January 30, 2004, "NEI

, Draft White Paper, Use of Generic Letter 91-18 Process and Alternative Source Terms in the Context of Control Room Habitability." (ADAMS Accession No. ML040300694).

~"-~

tf'**.

1 4-55b Amendm ent No. 264

4.12.2 REACTOR BUILDING PURGE AIR TREATMENT SYSTEM Deleted 4-55c Amendm ent No. 56, 68, 108, 149, 167, 170,176 ,218,22 6,240,2 46,264

CONTROLLED COPY r'j'°':"') 4.12.3 AUXILIARY AND FUEL HANDLING BUILDING AIR TREATMENT SYSTE M

. /

DELETED 4-55d Amendment No. 55, 76, 122,17 7,248

CONTROllED C PY THIS PAGE LEFT BLANK INTENTI ONALLY 4-55e Amendment No. 55, 122, 157,179 ,218,248

4.12.4 FUEL HANDLING BUILDING ESF AIR TREA TMEN T SYSTEM Applicability Applies to Fuel Handlihg Building (FHB) ESF Air Treatment System associated components. and Objective .

To vedft.t hat this system and associated components will be able to perform its design functions. . .

Specifi~ation 4.12.4.1 Each refueling interval prior to movement of.irradiated fuel:

a. The pressure drop across the entire filtration unit shall be demonstrat1:1d to be less than 7.0 inches of water at 6,000 cfm flow rate (+/-10%). *

b. . The tests arid sample analysis required by SpeGification 3.15.4.2 shall be performed.

4.12.4.2 Testing necessa*ry to denioristrate operability shall be performed as follows:

a. The tests a.nd sample analysis required by.Specification 3.15.4.2 shaHbe performed followjn_g sigilific!;lnt pain~ing, steam; fire, or chemical release in any ventilati,cm zone corniilunicating with the.~ystem that could contaminate the HEPA filters dr charcoal adsorbers.

b. DOP'testiflg sti~.11 be,perf9rrn~d after e*ach_ compl~te or.p_artial replacem.~nt bf.a HEPA filt~r bank, arid .aftl;}r ~ny structural mair:ite_narice on. the system housing that could affect the HEPA filter bailk,bypass*leakage. * *

c. Haloge,nate~ hydrocarbon testing sl)all be performed after each complete or partial replac$meint of a charcoal adsorber bank, and aft.er- any .structural maintenance on*the system housing

that could affect charcoal adsorber ba.nk bypass leakage.

4.12.4.3 Each filter train shali be operated for.~15 continuous minutes at the frequency specified in the Surveillance Frequency Control Pro9ra m.

4.12.4.4 An air flow distrib!Jtion test shall be performed c;m the HEPA filter bank initially and after any rn,aintenance or te.sting that could affec~ the.air fl9w*.distribution within the system. The distribution across the HEPA filter bank shall be uniform within

+/-20%. Th's test shall be performed at 6,000 cfm +/- 10% flow rate.

4-55f Amend ment No. 122,2 74,28 2

Bases Tf:ie FHB ESF Air Treatment System is a system which is norm any kept in a "stand Te.sts ar:id sample anafys.is assure that-the HEPA filters and by" operating status.

charcoal adsorb ers can perform as evaluated.

The charc,oal adsorb er efficiency test procedure stiould allow-f or:-the remov al of a san:iple from one adsorber test canister. Eat:h sample should be at least two thickness of .the bed: The in°place test criteria for activat a inches in diairi'eter and len-gth equal to the ed charcoal will meet the. guidelines of ANS1-N51'0-1980. The labora tciiyte st'of charcoal wiil beper forme d in accord ance witti'A STM D:3a03a198_9; If li;ibotatory test results are unacceptable,

all aqsorb ent in the system shall be replaced with an adsorbent qualified in accorqance with ASTM-03803, H:189. Any. HEPA filters found defective Will be replaced with filters qualified if\ accordance with ANSI-N 509-1980.

Pressure drop across the entire filtration _unit of less. ~han 7.0 inches of water at Jhe system design flow rate will indicate that the filters and adsorbers are hot clogge d by excess ive amounts of foreign matter.

Opeiration fo~ ~ 15 contiri\JOUS minutes at the frequency specified in the Surveillance Frequency Control Program_demonstrates OPERABILITY of the system. Period i9 operation ensure s that blockage, fan or motor .failure, or excessive vibration can be detected for corrective action.

If significant painting, st~am, fire, or chemi calrele ~se occurs such that the HEPA filter or charcoal adsorb er could becom e contaminated from the.fumes, chemic als or foreign.material, the same tests and samp.le analysis shall be performed as required for operat ional movem ent of irradiated fuel. The determination of what is significant shall be made by the Vice Presid enfTMI Unit 1.

4-55g Amend ment No. 122, 157, 179,2 18,22 6,274 ,282

CONTROUJED y t':~'t\

(~--t~)

4.13 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE Applic ability Applies to leakage testing of byproduct. source, and special nuclear radioac tive material sources.

Objective To assure that leakage from byproduct. source, and special nuclear radioactive material sources does not exceed allowable limits.

Specif kati on Tests for leakage and/or contamination shall be performed by the license e or by other persons specifi cally authorized by the Commission or an agreement State. as follows:

1. Each sealed source, except startup sources previously subjec t to flux. containing radioac tive material. other than Hydrogen 31 withcore half-li fe greate r than 30 days and in any form other than gas shall abe tested for leakage and/or contamination at interva ls not to exceed six

(~~i) months. *

....? The period ic leak.te st required does not apply to sealed sources that are stored and not being used. The sources excepted from this test shall be tested for leakage prior to any use or transfe r to another unless. they have been leak tested within six rronths prior to the dateuser use or transfe r. In the absence of a certifi cate from a transfe ror of indicat ing that a test has been made within six rronths prior to the transfe r, sealed sources shal 1 not be put into use until tested.

3. Each sealed source shall be tested within 31 days prior to being subjected to core flux and following repair or maintenance to the source.

4.14 DELETED 4-56 (page 4-57 deleted)

I Anendnent No. fl9. 129 3-31-81

COPY f.~ :?~"'~7~~.

t\,s:): 4.15 MAIN STEAM SYSTEM INSERVICE INSPECTION Appl icabil i t:t This technical specification applies to the inservice inspection of four welds in the Hain Steam System identified as HS-0001, MS-0002, HS-0003, and HS-0004L of the TMJ-1 Jnservice Inspection Program.

Objective The objective of the Inservice Inspection Program is to provide assurance of the continuing integrity of that portion of the Hain Steam System in which a postulated failure would produce pressures in excess of the compartment wall and/or slab capacities.

soecificatjo n 4.15.1. The four weld joints identified above shall be 100 percent inspected in accordance with the ASHE Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant components, defined in the TMI-1 Inservice Inspection Program. Inspections are to be performed at a frequency of once every 3-1/2 years (or during the nearest refueling outage).

Prior to initial plant operation, a preoperational inspection of the identified weld joints will be performed and any data acquired will be recorded to form a baseline on which to compare results of subsequent inspections.

Calculations (Reference 1) postulated that breaks in the main steam lines at the containment penetrations in small compartments No. 2 and No. 5 could produce pressures in excess of wall and/or slab capacities.

Inspections are conducted at an inspection frequency of 3 1/2 year intervals following initial plant startup. These inspections have revealed that no degradation of the welds has occurred during the inspection cycles up to and including the 9R outage inspection. Consequently, as further degradation is not expected to occur, justificatio n to extend the inspection frequency to once every ten (10) years is being developed. The conclusions of the technical benefit review will be submitted to the NRC for evaluation in a Technical Specification change request.

Reference (1) UFSAR, Appendix 14A, Section 7.2.1 4-58 Amendment No. jjf",167

4.16 REACTOR INTE RNA LS VENT VALVES SUR VEIL LAN CE Applicability Applies to Reactor Internals Vent Valves.

Objective To verify that no reactor internals vent valve is stuck in the open position and that each valve continues to exhib it freedom of move ment .

Specification Item . Test Frequency 4.16.1 Reac tor Internals Demonstrate Operability Vent Valves At the frequency specified By: in the Surveillance Frequency Control

a. Conducting a remote Program visual inspection of visually accessible sur-faces at the valve body and disc sealing faces and evaluating any observed surface irregu-larities.

b. Verifying that the valve is not stuck in an open position, and

c. Verifying through manu al actuation that the valve is fully open with a force of :Si 400 lbs. (applied vertically upward).

Base s Verifying vent valve freed om of movement insur es that coola nt flow does not bypass the core throu gh reactor intern als vent valves durin g operation and there fore insures the cons ervatism at Core Protection Safe ty limits as delineated in Figures and the flux/flow trip setpoint. 2.1-1 and 2.1-3, 4-59 Ame ndm ent No. 6a, 149, 274

4.17 SHOCK SUPPRESSORS (SNUBBER~~

SURVEILLANCE REQUIREMENTS

4. I 7.1 Each snubb er shall be demonstrated OPERABLE by performance of the following inspection progr am.

a. Snub ber Tvpe s As used in this specification, type of snubber shall mean snubbers of the same design and manu factu rer, irrespective of capacity.

b. Visual Inspections Snub bers are categorized as inaccessible or accessible during reactor operation and may be treate d independently. The Director-Radiological Health and Safety, \\ill ensure that a review is perfo rmed for ALA RA considerations on all snubb ers which are located in radiation areas for the determination of their accessibility. This review shall be in accordance with the recommendations of Regu latory Guides 8.8 and 8 .10. The determinatio n shall be based upon the kno\'t11 or projected radiation levels at each snubb er location which would render the area inaccessible durin g react or opera tion and based upon the expected time to perfo rm the visual inspection. Snub bers may also be determined to be inaccessible becau se of their physi cal location due to an existing industrial safety hazar d at the specific snubber location. This deten nination shall be re,..iewed and appro ved by the mana geme nt position responsible for occupational safety.

Snub bers accessible during reacto r operation shall be inspected in accordance \i.,ith the sched ule stated bel~w. Snub bers scheduled for inspection that are inaccessible during reactor opera tion becau se of physical location or radiation levels shall be inspected during the next reacto r shutd O\m great er than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months where acces s is restored* unles s previously inspected in accor dance "ith the sched ule stated below.

Visual inspections shall include all safety related snubb ers and shall be performed in accor dance

\\ith the following schedule:

No. Inoperable Snubbers of Each Subse quent Visual Tvpe per Inspection Period Inspection Period**#

0 24 months +/- 25%

I 16 months +/- 25%

2 6 months +/-25%

3,4 124 days +/-25%

5, 6, 7 62 days +/-25%

8 or more 31 days +/-25%

Snub bers may conti nue to be inaccessible during reacto r shutdO\m great er than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months (e.g. if purgi ng of the reactor building is.not permitted).

- The inspection interval for each type of snubb er shall not be lengthened more than one step at a time unless a gener ic probl em has been identified and corrected; in that event the inspection interval may be lengthened one step the first time and two steps therea fter if no inoperable snubbers of that type are found

The provisions of Table 1.2 are not applicable.

4-60 Amendment No. 30, l0i, 110, 175, 179 219

COPY SHOCK SUPPRESSORS <SNUBBERS>

SURVEILLANCE REQUIREMENTS (Continued)

c. Refueling Outage Inspections At leas t once each refu elin g cycle during shutd visual inspection shall be performed of all safe own, a snubbers attached to sections of safe ty systems ty rela ted that have experienced unexpected, pote ntia lly piping tran sien ts as determined from a review of ope damaging and a visual inspection of the systems. rational data

d. Visual Insp~ction Acceptance Crit eria Visual inspections shall veri fy: Cl> that ther visi ble indi cati ons of damage or impaired ope e are no

<2> attachments to the foundation or supportinrabi lity and are secure. Snubbers which appear inoperable g stru ctur e of visual inspections may be determined OPERABLE as a resu lt purpose of esta blis hing the next visual inspecti for the terv al, provided that : Cl> the cause of the rejeon in-clea rly esta blis hed and remedied for that part ctio n is snubber and for other snubbers that may be genicul ar susc epti ble, and (2) the affe cted snubber is eric ally test ed in the as found condition and determin func tion ally per Spe cifi cati on *4.1 7-lf . Hhen the rese rvoied OPERABLE

. of a snubber is found to be uncovered by flui r out let port snubber shal l on)y be declared operable if func d, the test ing in both extension and retr acti on dire tional sati sfac tory and an engineering evaluation conc ctio ns ts this snubber is operable. ludes that

e. Functional Tests*

At leas t once each refu elin g inte rval during repr esen tativ e sample of snubbers shal l be?teste shutdown, a one of the following ~ample plans. The samp d using be sele tted prio r to the test period and cannle plan shal l during the test period. The NRC Regional Adm ot be changed shal l be noti fied fn writing of the sample planinistrator prfo r to the test period, or the sample plan selected*

prio r test period shall be used: used in the

1) At leas t 101 of the tota l each type in the plan t shall be func tion ally test of snubber in use place or in a bench test . For each snub ed eith er in-that does not meet the functional test acce ber of a type ptance crit eria of Spe cifi cati on 4.1 7.lf , an add ition of that type of snubber shal l be fun ctio nall y al 101 unti l no more fail ures are found or unti l all test ed of that type have been func tion ally test ed; or snubbers

.,..~ . -- - '

The four 550,000 lb reac tor coolant pump snub included. The func tion al test program for reac bers are not

~) snubbers is implemented in accordance with the tor coolant pump othe r requirements of the snubber test ing prog schedule and ram.

4-61 Amendment No. ~ )Rf(. ~ 149

co*

SHOO< SUPPRESSORS (SNL.e8ERS)

SURVEILLANCE REQUIREMENTS (Continued)

2) A represe ntative sample of eac:h type of snubber shall be functio nally tested in accordance with Figure 4.17-1. "C" is the total number of snubbers of a type found not meeting the acceptance requirements of Specifi caticn 4.17.lf . _The cunulat ive number of snubbers of a* type tested is Ilenoted by "N". At the end of each day's testing , the new values of 11 N11 and "C" (previous day's total plus cur~nt day's increments) shall be plotted on Figure 4.17-1. l f at any time the point plotted falls in the "Reject" region all snut:lbers of that type

shall be functio nally tested. lf at any time the point plotted falls in the "Accept" region testing of that type of snubber may be tennina ted. \Vhen the point plotted lies in the "Continue Testing" region, additio nal snubbers of that type shall be tested until the point falls in the "Accept" region or the "Reject" region, or all the snubbers of that type have been tested. Testing equipment failure during functio nal testing may invalid ate that day's testing and allow that day's testing to resune anew at a later time, provided all snubbers testeo with the failed equipment during the day of equipment failure are reteste d.

The represe ntative sample selecte d for functio nal test sample plans shall be randomly selecte d from the snubbers of each type and reviewed before beginning.the testing . The review. shall ensure as far as practic able that they are represe ntative of the various configu rations , opetating environments, and the range of size and capacit y of snubbers of each type. Snubbers placed in the same locatio n as snubbers which failed the previous functio nal test shall be reteste d at the time of the next functio nal test but shall not be included in the sample plan. If, during* the functio nal test, additio nal sampling is required d~ to failure of only one type of snubber, the functio nal test ~esults sr.all be reviewed at that time to determine if additio nal samples should be limited to the type of snubber whic:h has failed the functio nal testing .

f *. Functional Test Acceotance Criteri a

The snubber functio nal test shall verify that:

1) Snubber activat ion (restra ining action or lockup) is achieved within the specifi ed velocit y range in both tension and compression. *

2) Snubber release rate (bleed) is achieved in both tension and canpression, within the specifi ed range.

3) Fastene rs for attachn ent of the snub.be!' to ttie component and to the snubber anchorage, are secure.

4-62

AIT";endment No. 10, 106

C PY SHOCK SUPPRESSORS (SNU88ERS)

SURVEILLANCE REQUIREMENTS (Continued)

Testing methods may be used to measure parameters indirect ly, or parameters other than those specifie d, if those results can be corxelat ed to the specifie d parameters through establis hed methods.

g. Functional Test Failure Analysis

1. Cause of Failure Evaluation An engineering evaluati on shall be made of each failure to meet the function al test acceptance criteria to determine the cause of the failure. The results of this evaluati on shall be used, if applicab le, in selectin g snubbers to be tested in an effort to determine the operabi lity of other snubbers, irrespec tive of type, which may be subject to the same failure mode.

2. Damage Evaluation For the snubbers found inoperable, an engineering evaluati on shall be performed on the components to which the inoperab le snubbers are attached . The purpose of this engineering evaluati on shall be to determine if the components to which the inoperab le snubbers are attached were adversely affected by the inopera bility of the snubbers in order to ensure that the component remains capable of meeting the designed service.

If any snubber selected for function al testing either fails to activate (leek~p) or fails*to extend or retract, i.e.,

frozen-i n-place, the cause will be evaluated and, if caused by manufacturer or design deficien cy, all snubbers of the same type which are,subj ect to the same defect shall be evaluate d in a manner to ensure operabi lity. This testing requirement shall be independent of the r~quirements stated in Specific ation 4.17.le fOl' snubbers not meeting the function al test acceptance c4iteria .

h.

Functional Testina of Recaired and Reolaced Snubbers

- Snubbers which fail the visual inspecti on or the function al test acceptance criteria shall be repaired or replaced. Replacement snubbexs and snubbers which have repairs which might affect the function al test result shall have been tested to meet the function al test criteria before installa tion in the unit.

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4-63 Amendment No. l Du

SHOCK SUPPRESSORS CSNUBBERS)

SURVEILLANCE REQUIREMENTS <Continued>

i. Snubber Seal Service Program A snubber seal service life program whereby the seal service life of hydsha ll be developed raul monitored to ensure that the service lifeic snubbers is between surv eill anc e inspections. The desiis not exceeded life for the various seals shall be esta blisgnated service engineering information. The sea ls sha ll hed based on tha t the indicated service life will not bebe replaced so during a period when the snubber is required exceeded OPERABLE. The seal replacements sha ll be to be docu the documentation shall be reta ined in acco mented and Spe cifi cati on 6.10.2.m. rdance with f'f~)

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~,., -" Amendment No. ~ J.,tt5", 149

0W8)

Bases All safet y relat ed hydraulic snubbers are integ rity and oper abili ty. The inspe ction visu ally inspected for over all includes verif icati on of proper orien tatio n, adequate hydraulic fluid leve l, and prop to piping and struc tures . er attachment of snubber The visu al insp ectio n frequency is based upon maintainin snubber prote ction *. Thus, the required inspe g a cons tant leve l of ction with the observed snubber failu res. The number of inop inter val varie s inve rsely during a required inspe ction determines the*t ime inter erabl e snubbers found requ ired insp ectio n. Insp ectio ns performed before that val for the next may be used as a new refer ence poin t to determine the inter val has elapsed However, the resu lts of such early inspe ction s performed next insp ectio n.

requ ired time inter val has elapsed (nominal time less before the orig inal lengthen the requ ired inspe ction inter val. Any inspe 25%) may not be used to

.requ ire a shor ter inspe ction inter val will override ction whose resu lts Those snubbers which are inac cess ible during reac tor the previous schedule.

requ ired to be inspe cted in accordance with the indic operation are not but*must be inspe cted during the next shutdown when ated inspe ction inter val access is resto red.

When the cause of the rejec tion of a snubber by visu estab lishe d and remedied for that snubber and al inspe ction is clea rly for any be gene rical ly susc eptib le, that snubber may be exem othe r snubbers that may inop erab le if it is determined operable by func tiona pted from being counted as

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\; '.} ,,,, susc eptib le snubbers are those snubbers which are of l testi ng. Generically and have the same design featu res direc tly relat ed toa ?pec ific make or ~odel snubbers by visu al inspe ction , or are simi liarl y locat rejec tion of the same environmental cond ition s such as temperature,. radia ed or exposed to the tion, and v~bration.

When a snubber is found inoperable, an engineerin addi tion to the determination of the snubber modeg ofevaluation is performed, in determine if any safe ty-re lated component or system failu re, in orde= to affec ted by the inop erab ility of the snubber. The engin has been adversely shal l determine whether or not the snubber mode of failu eering evalu ation sign ifica nt effe ct or degradation on the supported co~p re has imparted a onent or syste ~.

To provide assurance of snubber func tiona l sampling a~d acceptance crite ria methods aterelia bilit y, one of the two used:

1. Functionally test 10% of a type of snubber with an addi teste d for each func tiona l testi ng failu re, or tiona l 10%

2. Func tiona lly test a sample size and determine samp rejec tion using Figure 4.17-1. le acceptance or Figure 4.17-1 was developed using "Wald's Sequential as described in "Quality Control and Indu stria l Stat Prob abili ty Ratio Plan" Duncan. istic s" by Acheson J.

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4-65 Amendment No. l 06

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.Snubber sea l ser vic e lif e is eva infoT"iaation *through consideratio luated via manufacturer input and conditions and ass oci ate d ins tal n of the snubber ser vic e The requirement to'fflOnitor lat ion and maintenance reccn:ts.

the snu included to ensure tha t the sea bber sea l ser vic e lif e is perfonnance eva lua tio n in view ls per iod ica lly undergo a-~-

con dit ion s. ihe se records will of the ir age and ope rat ing fut ure con sid era tio n -~f snubbe provide sta tis tic al bases for requirements for the maintenance r sea l ser vic e lif e. The-*

ser vic e lif e are no t intended to of records and the snubber sea l aff ect pla nt oper.ation.

A technique and method for fun rea cto r coo lan t pump snubbers is ctional tes tin g of the 550,000 lb.

functional tes t program sha ll be cur ren tly under development. The Jul y 1. 1985, whichever is ea rli developed by Cycle 6 ref uel ing or sha ll be implemented in accordancer. The functional tes t program requirements of 1:he .program. e with the schedule and oth er A lis t of ind ivi du al snubbe information is maintained atrsthe with appropriate det ail ed ,

J pennanent del eti on of a snubbe pla nt sit e. As a bas is for -

snubbers, an engineering ana lysr from the lis t of saf ety rel ate d the ori gin al saf ety ana lys is desis must be perfonned to ver ify tha t exceeded. Snubber add itio ns and ign cri teT ia are eit he r met or in accordance with 10 CFR S0.59 del eti on s are reported to the NRC

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Amendment No. *1,J. 110

liEDOOPY Y)

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FIGURE 4.17-1 SNUBBER FUNCTIONAL TEST - s;.I*:?LE PLAN 2 Amendment No. ~ , 167

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(Pages 4-69 through 4-76A deleted) 4-68

4.19 STEAM GENE RATO R (SG) TUBE INTEGRITY Applicability: When ever the reactor coolant average tempe rature is above 200°F Surveillance Requirements (SR}:

Each steam generator shall be determined to have tube integrity by performance of the following:

4.19.1 Verify SG tube integrity in accordance with the Steam Gene rator Program.

4.19.2 Verify that each inspected SG tube that satisfies the tube plugging criteria is plugg ed in accordance with the Steam Generator Program prior to excee ding an average reactor coolant temperature of 200°F following an SG tube inspe ction.

BASES:

BACKGROUND Steam generator (SG} tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.

The SG tubes have a number of important safety functions.

Steam generator tubes are an integral part of the reactor coolant press ure boundary (RCPB) and, as such, are relied on to maintain the prima ry system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secon dary syste ms to remove heat from the primary system. This Specification addre sses only the RCPB integrity function of the SG. The SG heat remov al function is addressed by TS Section 3.4.

SG tube integrity means that the tubes are capable of perfo rming their intended RCPB safety function consistent with the licens ing basis, including applicable regulatory requirements.

Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting

, intergranular attack, and stress corrosion cracking, along with other mech anically induced phenomena such as denting and wear. These degra dation mechanisms can impair tube integrity if they are not mana ged effectively.

The SG performance criteria are used to manage SG tube degradation.

Specification 6.19, "Steam Generator (SG) Program," requir es that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.19, tube integr ity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and 4-71 Amendment No. 47, 261, 279 (12 22 78)

BACKGROUND (continued) ,

operational leakage. The SG performance criteria are described in Specification 6.19. Meeting the SG performance criteria provid es reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are define d by the Steam Generator Program Guidelines (Ref. 1).

APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of a SGTR event assumes a bound ing primary to secondary leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assum es the contaminated secondary fluid is only briefly released to the atmos phere via safety valves..and the majority is discharged to the main conde nser.

The analysis for design basis accidents and transients other than a SGTR assum e the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary leakage from all SGs of 1 gallon per minute or is assum ed to increase to 1 gallon per minute as a result of accident-induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is conservatively assumed to be equal to, or greate r than, the TS 3.1.4, "Reactor Coolant System Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel.

The dose consequences of these events are within the limits of GDC 19 (Ref.

2), 10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.g.,

a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO TS 3.1.1.2.a The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the plugging criteria be plugge d in

. accordance with the Steam Gener ator Program.

During a SG inspection, any inspected tube that satisfies the Steam Gener ator Program plugging criteria is removed from service by plugging.

If a tube was determined to satisfy the plugging criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall between the tube-to-tube sheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.

The tube-to-tubesheet weld is not considered part of the tube.

4-78 Amendment No~ 47, 153, 237, 2e1, 271, 279 (12 22 78)

LCO (continued)

A SG tube has tube integrity when it satisfies the SG performance criteria.

The SG performance criteria are defined in Specificati on 6.19, "Steam Generator Program," and describe acceptable SG tube performance. The

Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accid ent conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defin ed as, "The gross structural failure of the tube wall. The condition typica lly corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plast ic) tearing of the tube material at the ends of the degradation." Tube collap se is defined as, "For the load displacement curve for a given structure, collap se occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect ori burst or collapse. In that context, the term "significant is defin ed as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and seco ndary classifications will be base d on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code

, Section Ill, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the desig n specification. This includes safet y factors and applicable design basis loads based on ASM E Code, Section Ill, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

The accident induced leakage performance criterion ensures that the primary to secondary leakage caused by a design basis accident, other than a SGTR , is within the accident analysis assum ptions. The accid ent analysis assu mes that accident induced leakage does not exceed 1 gpm per SG.

4-79 Amen dmen t No. 47, 163, 20S, 209, 2S1, 271

LCO (continued) The accident induced leakage rate includes any prima ry to secondary leakage existing prior to the accident in addition to prima ry to secondary leakage induced during the accident.

The operational leakage performance criterion provi des an observable indication of SG tube conditions during plant operation.

The limit on operational leakage is contained in TS 3.1.6.3, "LEA KAGE," and limits primary to secondary leakage through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR unde r the stress conditions of a LOCA or a main steam line break. If this amou nt of leakage is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differentia l pressures across SG tubes can only be experienced when the reactor coolant system average temperature is above 200° F.

RCS conditions are far less challenging when avera ge temperature is at or below 200°F; primary to secondary differential press ure is low, resulting in lower stresses and reduced potential for leakage.

ACTIONS The ACTI ONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensa tory actions for each affected SG tube. Complying with the Required Actio ns may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of assoc iated Required Actions.

3.1.1.2.a.(3.)a. and 3.1.1.2.a.(3.)b.

3.1.1.2.a.(3.) applies if it is discovered that one or more SG tubes examined_in an inservice inspection satisfy the tube plugging criteria but were not plugged in accordance with the Steam Gene rator Program as required by Surveillance Requirement 4.19.2. An evalu ation of SG tube integrity of the affected tube(s) must be rnade. Steam generator tube integrity is based on meeting the SG performance criter ia described in the Steam Generator Program. The SG plugging crit~r ia define limits qn SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criter ia will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integ rity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradatio n prior to the next SG tube,inspection. If it is determined that tube integ rity is not being maintained, 3.1.1.2.a.(4.) applies.

4-80 Amendment No 116, 149, 163, 206, 237, 261, 271, 279

ACTIONS (continued)

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, Required Action 3.1.1.2.a.(3.)b. allows plant operation. to continue until the next refueling outage or SG inspection provided the inspe ction interval continues to be supported by an operational asses sment that reflects the affected tubes. However, the affected tube(s

) must be plugged prior to exceeding a reactor coolant average tempe rature of 200°F following the next refueling outage or SG inspection.

This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

3.1.1.2.a.(4.)

If the Required Actions and associated Completion limes of Condition 3.1.1.2.a.(3.) are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOW N within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENT SR 4.19.1:

During shutdown periods the SGs are inspected as requir ed by this SR and the Steam Generator Program. NEI 97-06, "Steam Gene rator Program Guidelines" (Ref. 1), and its referenced EPRI Guide lines, establish the content of the Steam Generator Program.

Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessmen t of the SG tubes is performed. The condition monitoring assessmen t determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performanc e criteria have been met for the previous operating period.

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes conta in flaws satisfying the tube plugging criteria. Inspection scope (i.e.,

which tubes or areas of tubing within the SG are to be inspected) is a functi on of existing and potential degradation locations. The Steam Generator Program also 4-81 Amen dmen t No. 47, 83, 91 , 103, 129, 149, 163, 167, 206, 209, 237, 261, 279

SURVEILLANCE REQUIREMENTS (continued) specifies the inspection methods to be used to find poten tial degradation.

Inspection methods are a function of degradation morphology

, non-destructive examination (NOE) technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.19.1. The frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Gene rator Program uses information on existing degradations and growth rates to deterrr:iine an inspection frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next sched uled inspection. In addition, Specification 6.19 contains presc riptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between sched uled inspections. If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 6.19 until subsequent inspections suppo rt extending the inspection interval.

SURVEILLANCE REQUIREMENT SR 4.19.2:

During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from servic e by plugging.

The tube plugging criteria delineated in Specification 6.19 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube plugging criteria, in conjunction with other elements of the Steam Gene rator Program, ensure that the SG performance criteria will contin ue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performanc e criteria.

4-82 Amen dmen t No. 47, 86, 116, 149, 163, 206, 209, 237, 261, 271, 279

The frequency of "prior to exceeding an average reactor coolant

~ temperature of 200°F following an SG tube inspection" ensures that the

\tffl Surveillance has been completed and all tubes meeting the plugging criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

REFERENCES

1. NEI 97-06, "Steam Generator Program Guidelines".

2. 10 CFR 50 Appendix A, GDC 19.

3. 10 CFR 100.

4. ASME Boiler and Pressure Vessel Code, Section Ill, Subsection NB.

5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Genera tor Tubes t August 1976.

6. EPRI, "Pressurized Water Reactor Steam Generator Examination Guideli nes".

4-83 (Pages 4-84 through 4-85 deleted)

Amendment No. 47, 129, 206, 209, 237, 261,27 1,279

4.20 REACTOR BUILDING AIR TEMPE RATUR E Applicability This specification applies to the average air temperature of the primary containment during power operations.

Objective To assure that the temperatures used in the safety analysis at the reactor building are not exceeded.

Specification 4.20.1 When the reactor is critical, the reactor building temperature will be checked at the frequency specified in the Surveillance Frequency Control Program. If any detector exceeds 130°F (120°F below elevation 320) the arithmetic average will be computed to assure compliance with Specification 3. 17.1.

4-86 Amendm ent No. 44,-47, 274

V 4.21 RADIOACTIVE EFFLUENT INSTRUMENTATION Deleted 4.21..1-'-- Radioactive* Li.quid Effluent Instrumentation Deleted 4.21.2 Radioactive Gaseous Process &Effluent Monitoring Instrumentation Deleted 4.22 RADIOACTIVE EFFLUENTS Deleted 4.22.1 Liquid Effluents Deleted 4.22.2 Gaseous Effluents Deleted 4.22.3 Solid Radioactive Waste Deleted 4.22.4 Total Dose Deleted 4.23 RADIOLOGICAL ENVIRONMENTAL MONITORING

Deleted ,

4.23.1 Monitoring Program Deleted

4. 23. 2* Land Use Census Deleted

4. 23. 3 lnterl aboratory Comparison Progr.am Deleted (r'-" ,. ~,,~-.,."

-,J 4-87

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(4-88 thru 4-122 deleted)

Amendment No. 7h JI, i,J, ni, 11f,-117, Ul, l]J, l77, 197

ML19106A161

Text

Subject:

Enclosure:

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