ML20117A155

Proposed Tech Specs Re Revision to TS Administrative Controls Section 6.1.F.1 Stating That Either 8 or 12 H Shifts Will Be Considered Normal & 40 H Will Be Considered Nominal

ML20117A155
Person / Time
Site:	Monticello
Issue date:	08/15/1996
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20117A131	List: ML20117A127 ML20117A155 ML20117A428
References
NUDOCS 9608260023
Download: ML20117A155 (62)

Text

l Exhibit B Monticello Nuclear Generating Plant License Amendment Reauest Dated Auaust 15.1996 Technical Specification Pages Marked Up with Proposed Wording Changes Exhibit B consists of the existing Technical Specification pages marked up with the proposed changes. Existing pages affected by this change are listed below:

_P [g!

3 i

iv vi vii 22 31 69 89 99 102 164 188 190 198t 198y 200 227c 227d 227e 229b 229i 229ff 232 233 234 240 241 243 244a 246b 247a 252 B-1 9608260023 960815 PDR ADOCK 05000263 p

PDR

}

Eau 1

3.13 and 4.13 Fire Detection Protection Systems 223,

A.

Fire Detection Instrumentation 223 B.

Fire-Suppression Water System 224-C.

Hose Stations 226 D.

Yard Hydrant Hose Houses 227' E.

Sprinkler Systems' 227a F.

Halon Systems 227b l

G.

Penetration Fire Barriers 227b

-H.

Alternate Shutdown System 227c 3.13 Bases 228 4.13 Bases SS6h 3.14 and 4'.14 Accident Monitoring Instrumentation 229a

[

3 3.15 and 4.15 Inservice Inspection and Testing 229f 3.15 and 4.15 Bases 229g 3.16 and 4.16 Radiation Environmental Monitoring Program 229h A.

Sampling'and Analysis 229h B.

Land Use Census 229j i

C.

Interlaboratory Comparison 229k

)

3.16 and 4.16 Bases 229t 3.17 and 4.17 Control Room Habitability 229u A.

Ventilation System 229u f

B.

Emergency Filtration System 229v

./

3.17 Bases 229y 4.17 Bases 229z i

5.0 DESIGN FEATURES 230 5.1 Site 230 5.2 Reactor 230 5.3 Reactor Vessel 230 5.4 Containment 230 5.5 Fuel Storage

. 231 5.6 Seismic Designs 231 6.0 ADMINISTRATIVE CONTROLS 232 6.1 Organization 232 6.2 Review and Audit 237 6.3 Special Inspection and Audits 243 i

6.4 Action to be taken if a Safety Limit is Exceeded 243 6.5 Plant Operating Procedures 244 6.6 Plant Operating Records 24 C e

248 6.7 Re orting Requirem %nts

-jQ sW GwdconvnewTa\\ QuMGcmow iv

- 4 v E /=/=

v

.m..

1 l

LIST OF TABLES i

Table No.

Eggg 3.1.1 Reactor Protection System (Scram) Instrument Requirements 28 4.1.1 Scram Instrument Functional Tests - Minimum Functional 32 Test Frequencies for Safety Instrumentation and Control Circuits 4.1.2 Scram Instrument Calibration - Minimum Calibration 34 Frequencies for Reactor Protection Instrument Channels 3.2.1 Instrumentation that Initiates Primary Containment 49 Isolation Functions 1

3.2.2 Instrumentation that Initiates Emergency Core Cooling Systems 52 j

i 3.2.3 Instrumentation that Initiates Rod Block 3

3.2.4 Instrumentation that Initiates Reactor Building Ventilation 59 l

Isolation and Standby Gas Treatment System Initiation 3.2.5 Instrumentation that Initiates a Recirculation Pump Trip 60 and Alternate Rod Injection 3.2.6 Instrumentation for Safeguards Bus Degraded Voltage and 60a Loss of Voltage Protection 3.2.7 Instrumentation for Safety / Relief Valve Low-Low Set Logic 60b 3.2.8 Other Instrumentation 60d 3.2.9 Instrumentation for Control Room Habitability Protection 60e 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod 61 Block and Isolation Instrumentation 4.6-1 Snubber Visual Inspection Interval 132a 3.8.1 Radioactive Liquid Effluent Monitoring Instrumentation 1981 3.8.2 Radioactive Geseous Effluent Monitoring Instrumentation 198k 4.8.1 Radioactive Liquid Effluent Monitoring Instrumentation 198m Surveillance Requirements 4.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation 198n Surveillance Requirements 4.8.3 Radioactive Liquid Waste Sampling and Analysis Program 198p 4.8.4 Radioactive Gaseous Waste Sampling and Analysis Program 198s vi REV 132 7/15/92-

LIST OF TABLES (continued)

Eggg Table No.

t 7h 3.13.1 Safety Related Fire Detection Instruments

-227m 3.14.1 Instrumentation for Accident Monitoring 229b 4.14.1 Minimum Test and Calibration Frequency for Accident 229d Monitoring Instrumentacion 4.16.1 Radiation Environmental Monitoring Program (REMP) 229-1 Sample Collection and Analysis 4.16.2 REMP - Maximum Values for the Lower Limits of Detection 229q 4.16.3 REMP - Reporting Levels for Radioactivity Concentrations 229s in Environmental Samples 6.1.1 Minimum Shift Crew Composition 236 4

/

vii REV -120 0/24t&L

Bases:

2.2 The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products.

It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

) }3f3()

The pressure safety limit of 1335 psig as measured in the vessel steam space is equivale to 1375 psig at the lowest elevation of the reactor coolant system. The 1375 paig value was ived from the design pressures of the reactor pressure vessel, coolant piping, and reeircula on pump casing.

The respective design pressures are 1250 psig at 575*F,1148 psig at 562*F, and sig at 575*F.

The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes: ASME Boiler and Pressure Vessel Code Section III-A for the pressure vessel, ASME Boiler and Pressure Vessel Code Section III-C for the recirculation pump casing, and the USAS Piping Code Section B31.1 for the reactor coolant system piping. The ASME Code permits pressure transients up to 10 percent over the vessel design pressure (110s x 1250 - 1375 psig) and the USAS Code permits pressure transients up to 20 percent over the piping design pressure (120% x 1148 -

1378 psig).

The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig and temper-ature of 575'F; this is more than a factor of 1.5 below the yield strength of 42,300 psi at this temperature. At the pressure limit of 1375 psig, the general membrane stress increases to 29,400 psi, still safely below the yield strength.

The reactor coolant system piping provides a comparable margin of protection at the established pressure safety limit.

2.2 BASES 22 REV 02 1/3/01

Table 3.1.1 - Continued

-Thehighdrywellpre_ssurescramfunctionsintheStartupandRunmodeswhennecessary6nkbyclos. int _the e.

contains m solation g during purging for containment inerting or de-inerting$ Verification manual N e bypass c M ion shall be noted in the control room log.

f.

One instrument channel for the functions indicated in the table to allow completion of surveillance testing, provided that:

1.

Redundant instrument channels in the same trip system are capable of initiating the automatic function and are demonstrated to be operable either immediately prior or immediately subsequent to applying the bypass.

2.

While the bypass is applied, surveillance testing shall proceed on a continuous basis and the remaining instrument channels initiating the same function are tested prior to any other. Upon completion of surveillance testing, the bypass is removed.

3.1/4.1 31 REV-52 A/vfat-

Bases Continued:

3.2 increases core voiding, a negative reactivity feedback. liigh pressure sensors initiate the pump trip in the event of an isolation transient.

Low level sensors initiate the trip on loss of feedwater (and the resulting MSIV closure). The recirculation pump trip is only required at high reactor power levels, where the safety / relief valves have insufficient capacity to relieve the steam which continues to be generated after reactor isolation in this unlikely postulated event, requiring the trip to be operable only when in the RUN mode is therefore conservative.

The ATWS high reactor pressure and low-low water level logic also initiates the Alternate Rod Injection System. Two solenoid valves are installed in the scram air header upstream of the hydraulic control units.

Each of the two trip systems energizes a valve to vent the header and causes rod insertion.

[

This greatly reduces the long term consequences of an ATUS event.

Voltage sensing relays are provided on the safeguards bus to transfer the bus to an alternate source when a loss of voltage condition or a degraded voltage condition is sensed. On losa of voltage this transfer occurs immediately. The transfer on degraded voltage has a time delay to prevent transfer during the starting of large loads. The degraded voltage setpoint corresponds to the minimum acceptable safeguards bus voltage for n a ti... m e w...ing-Joad during - 1ere of ::clant :: ident. An allowance for relay tolerance is included.

Safety / relief valve low-low set logic is provided to prevent any safety / relief valve from opening when there is an elevated water leg in the respective discharge line. A high water leg is formed immediately following valve closure due to the vacuum formed when steam condenses in the line.

If the valve reopens before the discharge line vacuum breakers act to return water level to normal, water clearing thrust loads on the discharge line may exceed their design limit. The logic reduces the opening setpoint and increases the blowdown range of three non-APRS valves followir, a scram. A 15-second interval between subsequent valve actuations is provided assuming one valve fails to a.sTealy snre LOcA I od TLT rwa's wv2m s. et1 gust e.

aT r%e L{ 80 \\1

e. 5.s.a M cd IV\\ CCS.

WITAg e.

_e 3.2 BASES 69 REV ^5 7,1/00

Bases Continued 3.3 and 4.3:

consequences of reactivity accidents are functions of the initial neutron flux. The require-mentofatleast3countspergecondassuresthatanytransient, should it occur, begins at or above the initial value of 10~

of rated power used in the analyses of transients from cold conditions.

One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's are provided as an added conservatism.

C.

Scram Insertion Times The control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent fuel damage; i.e.,

to prevent the MCPR from becoming less than the Safety Limit (T.S.2.1.A).

This requires the negative reactivity insertion in any local region of the core and in the overall core to be equivalent to at least the scram reactivity curve used in the transient analysis. The required average scram times for three control rods in all two by two arrays and the required average scram times for all control rods.are based on inserting thia amount of negative reactivity at the specified rate locally and in the overall core. Under these conditions.

• h:-ti: ;;l li=ite'ere ne.er re-+-d d -ing t% *r---*-=*= r-iei=- cartrel r ' ::r.-

"h; li=iting :p:::tirn:1 tr xci:nt i: th:t ::::1 ting frc

turiin: :::p ::1r: :12:rr: ri_. frilure f th turhin: Lir;;; ;7;;;;-

- ;;i ;i; Of thi; C;-- ;i; t ;h;--

th;; th; 2;6;;iT; I;;;tiviti 7

r ter :::riting fr:= th: ::::: rith th: _. r:g: :::p: :: cf 211 th; drir : :: given in th: 25 ::

Sp::ificati:n, p-..id; th; :: quired pret ti:n, and MCPR remains above the Safety Limit (T.S.2.1. A).

T tt e b c.

e

/

he C.PR r af ay Lw k neuar ete.ecAed Aiq ag rrusica regery control rol sen 3.3/4.3 BASES 89 REV-'?

11/11/0L

m s

v Basis 3.4 and 4.4:

A.

The design objective of the standby liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted. To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of boron in the reactor core in less than 125 minutes sufficient to bring the reactor from full power to a 3% delta k suberitical condition considering the hot to cold reactivity swing, xenon poisoning and an additional 25% boron concentration margin for p ::ible imper-

-fect-midng-of-the-chemical-solution-in-the-reaetor-water-and-di-lut-len-from-the-water-in-the-cccid :n

. -dreui To d ot.afor I eab sqcrfe 6 mwg }

The time ine' nit (125%nutes) for insertio '6Pth?bo'rWsutu'ti"on was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak.

The ATUS Rule (10CFR50.62) requires the addition of a new design requirement to the generic SLC System design basis.

Changes to flow rate, solution concentration or boron enrichment to meet the ATWS Rule do not invalidate the original system design basis. Paragraph (c)(4) of 10CFR50.62 states that:

"Each boiling water reactor must have a Standby Liquid Control System (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight

~

percent sodium pentaborate solution" (natural boron enrichment).

- The described minimum system parameters (equivalent to 24 gpm, 10.7% concentration and 55 atom

- percent Boron-10 enrichment) will ensure an equivalent injection capability that meets the ATUS rule requirement.

Boron enrichment concentration, solution temperature, and volume (including check of tan'k heater and pipe heat tracing system) are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Only one of the two standby liquid control pumping circuits is

/

needed for proper operation of the system.

If one pumping circuit is found to be inoperable, there

[

is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made.

A reliability analysis indicates that the plant can be operated safely in this manner for ten days.

For additional margin, the allowable out of service time has been reduced to seven days.

The only practical time to test the standby liquid control system is during a refueling outage and by initiation from local stations. Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refueling outage unnecessary.

A test of explosive charges from one manufacturing batch is made to assure that the replacement charges for the tested system are satisfactory. A continual check of the firing circuit continuity is provided by pilot lights in the control room.

The relief valves in the standby liquid control system protect the system piping and positive

-, displacement pumps which are nominally designed for 1500 psi from overpressure. The pressure relief

_.-valves discharge back to the standby liquid control solution tank.

3.4/4.4 BASES 99 REV 127 2/15/01

3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.

One of the following conditions of 4.

Perform the following tests:

inoperability may exist for the period specified:

Item Frecuency

[

a.

One Core Spray subsystem may be inoperable Motor Operated Pursuant to for 7 days, or Valve Operability Specification J

b.

One RHR pump may be inoperable for 30 days, r

ADS Valve Each Operating

)

c.

One low pressure pump or valve (Core Spray Operability-Cycle i

or RHR) may be inoperable with an ADS valve

)

inoperable for 7 days, or Note: Safety / relief valve operability is

(

verified by cycling the valve and observing

)

d.

One cf the two LPCI injection paths may be a compensating change in turbine bypass inoperable for 7 days, or valve position.

OC

-C.On'Wo\\

e.

Two RHR pumps may be inoperable for 7 days, ADS Inhibit Each Operating or Switch Operability Cycle f.

Both of the LPCI injection paths may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or Perform a simulated Each Operating

(

g.

HPCI may be inoperable.for 14 days, automatic actuation test Cycle i

provided RCIC is operable, or (including HPCI transfer to e88 n Po and

/ h.

ADS valve may be inoperable for 14 ;

subsequent low reactor 1.

Two or more ADS valves may be inoperable water level) s for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

('

5.

Perform the following test on the Core 4.

If the requirements or conditions of 3.5. A.1, Spray Ap Instrumentation:

2 or 3 cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor Check once/ day l

shall be placed in a condition in which the

)

affected equipment is not required to be Test Once/ month f

i j

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

t Calibrate Once/3 months i

3.5/4.5 102 i

REV 129 '- / 9 / 91 l

i

=

)

m m.

. =

m

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

4. Pressure Suppression Chamb y-Drywell_Yacuum 4.

Pressure Suppression Chamber-Drywell Vacuum Breakers

, gpre 7153 Breakers Operability and full closure of the a.

When primary containment is requ red, all a.

eight drywell-suppression chamber vacuum drywell-suppression chamber vacuum breakers shall be operable and positioned breakers shall be verified by performance in the closed position as indicated by of the following:

the position indication system, except during testing and except as specified in (1)

Monthly each operable drywell-3.7.A.4.b through 3.7.A.4.d below.

suppression chamber vacuum breaker shall be exercised through an b.

Any drywell-suppression chamber vacuum opening-closing cycle.

breaker may be nonfully closed as indicated by the position indication and (2)

Once each operating cycle, dry-alarm system provided that drywell to well to suppression chamber suppression chamber differential pressure leakage shall be demonstrated to decay does not exceed that shown on be less than that equivalent to a Figure 3.7.1 one-inch diameter ( _-ifice and each vacuum breaker shall be visually c.

Up to two drywell-suppression chamber inspected.

(Containment access vacuum breakers may be inoperable required) provided that: (1) the vacuum breakers are determined to be fully closed and at (3)

Once each operating cycle, vacuum least one position alarm circuit is breaker position indication and operable or (2) the vacuum breaker is alarm systems shall be calibrated secured in the closed position or and functionally tested.

replaced by a blank flange.

(Containment access required) d.

Drywell-suppression chamber vacuum (4)

Once each operating cycle, the IT breakers may be cycled, one at a time, vacuum breakers shall be tested to during containment inerting and determine that the force required deinerting operations to assist in to open each valve from fully purging air or nitrogen from the closed to fully open does not suppression chamber vent header.

exceed that equivalent to 0.5 psi acting on the suppression chamber face of the valve disc.

(Containment access required) 164 REV-130 S/12/^1-3.7/4.7

m.___

____..m._,__.__.___

Bases Continued:

B.

Standby Gas Treatment System, and C.

Secondary Containment Initiating reactor building isolation and operation of the standby gas treatment system to maintain the design negative pressure within the secondary containment provides an adequate test of the reactor building isolation valves and the standby gas treatment system.

Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system operational capability. Secondary Containment Capability Test data obtained under non-cala conditions is to be extrapolated to calm wind conditions using information provided in "Sununary Technical Report to the United States Atomic Energy Commission, Directorate of Licensing, on Secondary Containment leak Rate Test", submitted by letter dated July 23, 1973, and as g,,

described in NSP letter to the NRC dated August 18, 1995, with subject, " Revision 2 to License Amendment

[

Request Dated June 8,1994, Standby Gas Treatment and Secondary Containment Technical Specifications."

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Standby gas treatment system inplace testing procedures will be established utilizing applicable sections of ANSI N510-1989 standard as a procedural guideline only. If painting, fire, or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals, or foreign materials, the same tests and sample analysis should be performed as required for operational use.

Replacement adsorbent should be qualified according to the guidelines of Regulatory Guide 1.52 Revision 2 (March 1978. The charcoal adsorber efficiency test procedures will allow for the removal of a representative ample. The 30*C, 95% relative humidity test per g

ASTM D 3803-89 is the test method to establish the methyl odine removal efficiency of the adsorbent. The sample will be at least two inches in diameter and a len equal to the thickness of the bed.

If the iodine removal efficiency test results are unacceptable, all adsorbent in the system will be replaced. High efficiency particulate filters are installed before and fter the charcoal filters to prevent clogging of the carbon adsorbers and to minimize potential release f particulates to the environment. An efficiency of 991 is adequate to retain particulates that may be rel ased to the reactor building following an accident.

This will be demonstrated by inplace testing with DOP as the testing medium. Any HEPA filters found defective will be replaced with filters qualified pu uant to regulatory guide position C.3.d of Regulatory Guide 1.52 Revision 2 (March 1978). Once per opera ng cycle demonstration of HEPA filter pressure drop, operability of inline heaters at rated power, aut tic initiation of each standby gas treatment system circuit, and leakage tests after maintenance or te ting which could affect leakage, is necessary to assure system performance capability.

4.7 BASES

- ^

188

~___

s REV

C MW A-STw\\ b og g DCC[ D I

l

~,_ _

v w

Bases Continued:

The containment is penetrated by a large number of small diamet - instrumer,.t lines. A program for the periodic testing (see Specification 4.7.D) and examination of the valves in these lines has been developed and a report covering this program was submitted to the AEC on July 27, N The main steam line isolation valves are functionally tested on a more frequent interval to

\\97 3 establish a high degree of reliability.

~

E.

Combustible Gas Control System 0 7 The Combustible Gas Control System (CGCS) is functionally tested once every six months to ensure that the recombiner trains will be available if required.

In addition, calibration and

(

maintenance of essential components is specified once each operating cycle.

190 4.7 BASES REV 0; 12,','"

TABLE 4.8.4

- RADIOACTIVE CASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM (continued)

(Page 2 of 2)

Notes:

The LLD is the smallest concentration of radioactive material in a sample that will be detected with 951 I

a.

probability with SI probability of falsely concluding that a blank observation represents a "real" signal.

Note (a) of Table 4.B.3 is applicable.

b.

Grab samples taken at the discharge of the plant stack and reactor building vent are generally below minimum detectable levels for most nuclides with existing analytical equipment.

For this reason, isotopic analysis data, corrected for holdup time, for samples taken at the steam jet air cjector may be used to calculate noble gas ratios Whenever the steady state radiciodine concentration is greater than 10 percent of the limit of Specification c.

3.6.C.1, daily sampling of reactor coolant for radioactive iodines of I-131 through I-135 is required.

the Whenever a change of 25% or more in calculated Dose Equivalent I-131 is detected under these conditions,il iodine and particulate collection devices for all release points shall be removed and analyzed daily unt it is shown that a pattern exists which can be used to predict the release rate.

Sampling may then revert to weeklg. When samples collected for one da{n 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after removalare analyzed, the corresponding LLD's may be increa factor o 10.

Samples shall be analyzed with To be representative of the average quantities and concentrations of radioactive materials in particulate d.

form in gaseous effluents, samples should be collected in proportion to the rate of flow of the effluent streams.

The principal gamma emitters for which the LLD specification will apply are exclusively the following e.

Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Kr-87, Kr-88, Xe-133, Xe-133m,141, and Ce-144 for radio-nuclides:

Mo-99 Cs-134 Cs-137, Ce-This list does Co-58, Co-60, 2n-65,hese nu,clides are to be detected and reported. particulate emissions.Other peaks which are measurable and not mean that only t identifiable, together with the above nuclides, shall also be identified and reported.

f.

Nuclides which are below the LLD for the analyses shall be reported as "less than" the LLD of the nuclide and should not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations.

When unusual circumstances result in LLD's higher than reported, the reasons shall be documented in the semiannual effluent report.

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period g.

sampled.

/1h.

if the limits of 3.8.B.1 are satisfied for other 3 analysis shall not be gequired prior to purginfeted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after sampling.

H

/.3 nuclides. However, the H analysis shall be comp 1.

In lieu of grab samples, continuous monitoring with bi-weekly analysis using silica-el samplers may be provided.

198t 3.8/4.8 REV 140 0/7/0'

pe 3.8 and 4.8 Bases: (continued)

Specification 3.8.B.4.c is pro ded to ensure that the concentration of potentially explosive gas mixtures contained in the compressed orage subsystem is maintained below the flammability limits of hydrogen and oxygen.

Automatic control features e included in the system to prevent the hydrogen and oxygen concentrations from l

reaching these flammabilit limits. Maintaining the concentration of hydrogen below the fla==,bility limit I

provides assurance that e releases of radioactive materials will be controlled in conformance with the require-ments of General Design riterion 60 of Appendix A to 10 CFR Part 50.

j Specification 3.8.B.4 is provided to limit the radioactivity which can be stored in one decay tank. Restricting the, quantity of ra i etivity contained in each gas storage tank provides assurance that in the event of an uncontrolled relea the tanks contents, the resulting total body exposure to an individual at the site restricted area bo dary will not exceed 20 mrem. A flow restrictor in the discharge line of the decay tanks prevents a tank being discharged at an uncontrolled rate.

In addition, interlocks prevent the contents of a tank.from being released with less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of holdup.

Specification 3.8.B.5 establishes a maximum activity at the steam jet air ejector. Restricting the gross radio-activity rate of noble gases form the main condenser provides reasonable assurance that the total body exposure to an individual at the restricted area boundary will not exceed the limits of 10 CFR Part 20 in the event this effluent is inadvertently discharged directly to the environment with minimal treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.

Specification 3.8.B.6 requires the containment to be purged and vented through the standby gas treatment system except during inerting and deinerting operations. This provides for iodine and particulate removal fron'the containment atmosphere.

During outages when the containment is opened for maintenance, the containment ventilation exhaust is directed to the monitored reactor building vent. Use of the 2-inch flow path prevents damage to the standby gas treatment system in the event of a loss of coolant accident during purging or venting. Use of the reactor building plenum and vent flow path for inerting and

(

deinerting operations permits the control room operators to monitor the activity level of the resulting

)

effluent by use of the Reactor Building Vent Wide Range Gas Monitors.

In addition, the Reactor Building Plenum Monitors will automatically terminate releases if their release rate limits are exceeded.

In the event that the reactor building release rate exceeds tne Reactor Building Vent Wide Range Gas Monitor alarm settings, the monitors will alarm in the control room alerting the operators to take actions to N

limit the release of gaseous radioactive effluents. An analysis has been performed which shows that the

/

control room operators would have in excess of two hours to take manual actions to terminate releases

/

without exceeding the permissible levels of radiation exposure in 10 CFR Part 20 Section 20.105(a). The

(

alarm settings for the reactor building vent wide range gas monitors are calculated in accordance with the NRC approved methods in the ODCM to ensure that alarms will alert control room operators prior to the limits of 10CFR Part 20, Section 20.105(a) being exceeded.

b C.

Solid Radioactive Waste Specification 3.8.C.1 provides assurance that the solid radwaste system will be used whenever solid radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR Part 50.36a and General Design Criteria 60 of Appendix A to 10 CFR Part 50.

3.8/4.8 BASES 198y REV 12'. C '?5 '??

=%

A 3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 2.

Both diesel generators are operable and capable of feeding their designated 4160 volt buses.

3.

(a) 4160V Buses #15 and #16 are energized.

(b) 480V Load Centers #103 and #104 are energized.

4.

All station 24/48, 125, and 250 volt t

batteries are charged and in service, and associated battery chargers are operable.

B.

When the mode switch is in Run, the availability of electric power shall be as specified in 3.9.A.

except as specified in 3.9.Brl, 3.0.0.2, 3.7.L.3 I

2xf 2. ?. 5. ': or the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1.

Transmission Lines From and after the date that incoming

[

power is available from only one line, reactor operation is permissible only i

during the succeeding seven days unless i

an additional line is sooner placed in i

l i

i 3.9/4.9 200 REV 101 10/10 '07-l f

-m3

-s

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 11.

Alternate Siutdown System

_ _ 11. Alternate Shutdown System ThesystemcontrolsontheASDSpaneh-c.,. pome 1.

Switches on the alternate shutdown system 1.

(

hall be operable whenever that : :t panel shall be functionally tested once 15 ;.t-al-cre required to be operable, per operating cycle.

^

f.

t'- aa-e el :::.

2.

The alternate shutdown system panel master

/

2.

If system controls required to be transfer switch shall be verified to alarm operable by Specification 3.13.11.1 are in the control room when unlocked once per made or found inoperable, restore the operating cycle.

inoperable system control to operable s

within 7 days, or perform one of the following; a.

Provide equivalent shutdown capability and within 60 days restore the inoperable system controls to operable; or i

i b.

Establish a continuous fire watch in the

\\

cable spreading room and the back-panel

\\

area of the control room and within 60 days restore the inoperable system controls to operable; or c.

Verify the operability of the fire detectors in the cable spreading room and the back-panel area of the control room and establish a hourly fire watch patrol and within 60 days restore the inoperable system controls to operable; or d.

Place the reactor in a condition where the systems for which the system controls at the ASDS are inoperable are not required to be operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

The alternate shutdown system panel master transfer switch shall be locked in the normal position except when in use, being tested or being maintained.

I 3.13/4.13 227c REV -111 3/b av f

W 4@[D TABLE 3.13.1 SAFETY RELATED FIRE DETECTION INSTRUMENTS Y

Minimum Instruments Operable v

Fire Zone location ligar Flame Smoke 1A "B" RHR Room 3

1B "A" RHR Room 3

1C RCIC Room 3

IE HPCI Room 2

IF Reactor Building-Torus Compartment 11 2A Reactor Bldg. 935' elev - TIP Drive Area 1

2B Reactor Bldg. 935' elev - CRD HCU Area East 10 2C Reactor Bldg. 935' elev - CRD HCU Area West 11 2::

QN Reactor Bldg. 935' - LPCI Injection Valve Area 1

3B Reactor Bldg. 962' elev - SBLC Area 2

3C Reactor Bldg. 962' elev - South 5

3D Reactor Bldg. 962' elev - RBCCW Pump Area 4

4A Reactor Bldg. 985' elev - South 4

4B Reactor Bldg. 985' elev - RBCCW Hx Area 5

4D SBGT System Room 2

SA Reactor Bldg. 1001' elev - South 7

SB Reacto-Bldg. 1001' elev - North 3

SC Reactor Bldg. - Fuel Pool Cooling Pump Area 1

6 Reactor Building 1027' elev 5

7A Battery Room 1

7B Battery Room 1

7C Battery Room 1

B Cable Spreading Room 7

3.13/4.13 227d REV 111 3/29/89

_._.m I

f.

-o^

TABLE 3.13.1

+

e N

SAFETY REIATED FIRE DETECTION INSTRUMENTS I

D y

Minimum Instruments Operable Fire Zone Location

}{ eat Flame Smoke 12A Turbine Bldg. - 911' - 4.16 KV Switchgear 3

13C Turbine Bldg. - 911' evel - MCC 133 Area 1

F ^.

Turbine Bldg. - 931' - 4.16 KV Switchgear 2

15A//sd-

1. 12 DG Room & Day Tank Room 3

l 15B/jqp

1. 11 DG Room & Day Tank Room 3

L 16 Turbine Bldg. 931' elev - Cable Corridor 3

17 Turbine Bldg. 941' elev - Cable Corridor 3

19A Turbine Bldg. 931' elev - Water Treatment Area 5

19B Turbine Bldg. 931' elev - MCC 142-143 Area 1

19C Turbine Bldg. 931' elev - W Pipe Chase 1

20 Heating Boiler Room 1

23A Intake Structure Pump Room 3

31A ist Floor - Reactor Building Addition - Division I 3

31B lat Floor - Reactor Building Addition - Division II 15 32A 2nd Floor - Reactor Building Addition - Division I 6

32B 2nd Floor - Reactor Building Addition - Division II 4

/

33 3rd Floor - Reactor Building Addition 5

(, )

f 3.13/4.13 227e REV 111

?,'? ?,'? ?

t Table 3.14.1 Instrumentation for Accident Monitoring Function Total No. of Minimum No. of Required Instrument Channels Operable Channels Conditions

• Reactor Vessel Fuel Zone Water Level 2

1 A, B Safety / Relief Valve Position 2

1 A, C (One Channel Pressure Switch and One Channel Thermocouple Position Indication per Valve) i Drywell Wide Range Pressure 2

1 A, B Suppression Pool Wide Range Level 2

1 A, B y

Suppression Pool Temperature 2

1 A, D

(

Drywell High Range Radiation 2

1 A, D Drywell and Suppression Pool 2

1 A, B Hydrogen and Oxygen Monitor i

Offgas Stack Vide Range Radiation 2

1 A, D Reactor Bldg Vent Wide Range Radiation 2

1 A, D

• Required Conditions A. When the number of channels made or found to be inoperable is such that the number of operable channels is less tl.an the total number of channels, either restore the inoperable channels to operable status within seven days, or prepare and submit a special report to the Commission pursuant to Technical Specification E..S.3-within the next 30 days outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status.

l.J. h 3.14/4.14 229b REV 112

~/10/00.

I I

---. -~

---. r

w.

v s

[

F 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 2( Welds i austenitic air'.ess steel pipi four ches or larg in diame r cent ning reactor coolant a a te erature abov 200 degre F during

\\ete -

%p er operatio, includin reactor ves 1 tachments d appurte nces, shall e included i n augment inspectio rogram meeting t requirements of Gener Letter 88-01.

B.

Inservice Testing 1.

Inservice Testing of Quality Group A, B, and C pumps and valves shall be performed in accordance with the requirements for ASME Code Class 1, 2 and 3 pumps and valves, respectively, contained in Section XI of the ASME boiler and Preocure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) except where relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55(a)(g)(6)(1), or where alternate testing is justified in accordance with Generic Letter 89-04.

2.

Nothing in the ASME Boiler and Pressure Vessel code shall be construed to supersede the requirements of any Technical Specification.

3.15/4.15 229ff REV i-27 2/15/91

-.._-_.-_-..--___-..m.

i v

w i

t t

?

3.0 LIMITING CONDITIONS FDR OPERATION 4.0 SURVEILIANCE REQUIREMENTS i

3.

Deviations are permitted from the required sampling schedule if samples are unobtainable due to hazardous conditions, seasonable unavaila-bility, or to malfunction of automatic sampling equipment.

If the latter occurs, every effort shall be made to complete corrective action l

prior to the end of the next sampling period.

l r

4.

With the level of radioactivity in an environ-l mental sampling medium exceeding the reporting-levels of Table 4.16.3 when averaged over any calendar quarter, submit a special report to the

/

Commission within 30 days from the end of the

[l affected calendar quarter pursuant to

/

Specification -5

..?.

When more than one of

(

the radio s in Table 4.16.3 are detected in ling medium, this report shall be (G[,h submitted if:

concentration (1) # concentration (2) limit level (1) limit level (2) + */* > 1.0 j

When radionuclides other than those in Table 4.16.3 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications t

3.8.A.2, 3.8.B.2, or 3.8.B.3.

This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiation Environmental Monitoring Report.

i 2291 3.16/4.16 REV "O 7/1/00

~

v s

i 6.0 ADMINISTRATIVE CONTROLS j

l 6.1 Organization A.

The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for the safe operation and maintenance of the plant.

During eriods when the Plant Manager is unavailable, this responsibility may be delegated to

\\

oth personnel.

)

Th. " L M :-'p W MMk b erMOR.r, a designated individual during periods of absence from the control or's office) shall be responsible for the control room command function.

room n sh a

B.

Offsite and Onsite Organizations

)

Onsite and offsite organizations shall be established for plant operation and corporate g

management, respectively. The onsite and offsite organizations shall include positions for

)

activities affecting plant safety.

(

1.

Lines of authority, responsibility and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating i

organization positions. These relationships shall be documented and updated, as appropriate,

/

in the form of organization charts, function descriptions of department responsiblities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements are documented in corporate and plant procedures, or the Updated Safety Analysis Report or the Operational Quality Assurance Plan.

2.

'Ihe Vice President Nuclear Generation shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the s

staff in operating, maintaining and providing technical support to the plant to ensure

)

nuclear safety. This position has the responsiblity for the Fire Protection Program.

3.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall

/

have sufficient organizational freedom to ensure their independence from operating pressures.

6.1 232 REV-iin sp*foF

C.

Plant Staff 1.

Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.1.1.

2.

At least one licensed operator shall be in the control room when fuel is in the reactor.

3.

At least two licensed operators shall be present in the control room during cold startup, scheduled reactor shutdown, and during recovery from reactor trips.

4.

An individual qualified in radiation protection procedures shall be on site when g

fuel is in the reactor.

c-d[g 5

All alterations of the reactor core shall be directly supervised by a license Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

g 6.

A fire brigade of at least five members shall be maintained on site at all f@

times.*

The fire brigade shall not include the three members of the shift ge5 organization required for safe shutdown of the reactor from outside the contro room.

7.

The General Superintendent, Operations shall be formerly licensed *as a Senior Reactor Operator or hold a current Senior Reactor Operator License.

/

8.

At least one member of plant management holding a current Senior Reactor Operator Lice

[

shall be assigned to the plant operations group on a long term basis (approximate 1 o

years). This individual will not be assigned to a rotating shift.

D.

Each member of the unit staff shall meet or exceed the minimum qualifications g ANSI N18.1-1971 for comparable positions, except for (1) the Experint:xf: t ":dicti:r Pr t:: _.. who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents, and (3) the General Ex;;r'nd xt, Operations who shall meet the requirement of ANSI N18.1-1971 except that NRC lic e requirements are as specified in A

Specification 6.1.C.7.

The training program shall e under the direction of a designated member o6ge.&

ofy :'. rr. States Power management.

• Fire Brigade composition may be less than the minimum requir ents for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of Fire Briga members provided immediate action is taken to restore the Fire Brigade to within the minimum requirements.

6.1 233 l

f E.

A training program for individuals serving in the fire brigade shall be maintained under the direction of a designated member of Northern States Power management.

This program shall meet the requirement of Section 27 of the NFPA Code - 1976 with the exception of training scheduling.

Fire brigade training shall be scheduled as set forth in the training program.

F.

Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g.,

senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance geeg )

s4, he# syn E personnel.

Procedures shall include the following provisions:

qJe g'"l,-- SA '

O (g) of 1.

Adequate shift coverage shall be maintained withou ne fgo'd\\,s)\\

heavy use of overtime.

The objective shall b o have operating i

I personnel work a normal "-Serr 2-y,

'^ ;. ;r

_:S while the plant is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:

a.

An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time.

b.

Overtime should be limited for all nuclear plant staff personnel so that total work time does not exceed 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any

'48-hour period, not more than 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> in any seven day period, all excluding shift turnover time.

Individuals should not be required to work more than 15 consecutive days without two consecutive days off.

c.

A break of at least eight hours including shift turnover time should be allowed between work periods.

d.

Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

6.1 234 REV 1:0 7,'1 '. '9 0

7.

Authority The SAC shall be advisory to the Vice President, Nuclear Generation.

8.

Records a vi Y AM 6f Minutes shall by prepared and retained for al schedule meetings of the Safety-Audit Committee.

The minutes shall be distribute ithin one month of the meeting to the Vice President, Nuclear Generation, the 0:::_ 1 M__ ;:r "url::: 21...;_, each member of the SAC, and others designated by the Chairman or Vice Chairman. There shall be a formal approval of the minutes.

9.

Procedures A written charter for the SAC shall be prepared that contains:

a.

Subjects within the purview of the group, b.

Responsibility and authority of the group.

c.

Mechanisms for convening meetings.

d.

Provisions of use of specialists or subgroups, Authority to obtain access to the nuclear power plant operating record files e.

and operating personnel when assigned audit functions.

f.

Requirements for distributilon of reports and minutes prepared by the group to others in the NSP Organization.

6.2 240 REV 95 7/?/?'

_..--_-___.m._-

.__.._____.._._...___m m.___..._

.. m.

k B.

Operations Committee (OC) i 1.

Membership The Operations Committee shall consist of at least six (6) regular members drawn from the key super-visors of the on-site supervisory staff.

The Plant Manager shall serve as Chairman of the.

OC and shall appoint a regular member to act as Vice Chairman in his absence. Alternates to the regular members shall be designated in writing by the Chairman, or Vice Chairman in the Chairman's absence, to serve on a temporary basis. No more than two alternates shall participate as voting members of the Operations Committee at any one time.

2.

Meeting Frequency The Operations Committee will meet on call by the Chairman or as requested by individual members and at least monthly.

3.

Quorum A quorum shall include a majority of the membership, incl,uding the Chairman or Vice Chairman.

/

4.

Responsibilities - The following subjects shall by reviewed by the Operations Committee:

a.

Proposed tests and experiments and their results.

b.

Modifications to plant systems or equipment as described in the Updated Safety Analysis Report and having nuclear safety significance or which involve an unreviewed safety question as defined in 10 CFR 50.59.

c.

Proposals which would effect permanent changes to normal and emergency operating procedures and any other proposed changes or procedures that are determined by the Plant Manager to affect nuclear safety.

d.

Proposed changes to the Technical Specifications or operating license.

e.

All reported or suspected violations of Technical Sp'ecifications, operating license requirements, administrative procedures, or operating procedures. Results of investi-gations, including evaluation and recommendations to prevent recurrence, will be reported, in writing, to the C;ncr:1.";n;g;r "nsi;;r Picnt; and to the Chairman of the Safety Audit Committee.

m Nsu w

c_%

-' ~ --

R

c.

Mechanism for scheduling meetings d.

Meeting agenda e.

Use of subcommittee f.

Review and approval, by members, of OC actions g.

Distribution of minutes 6.3 Special Inspections and Audits A.

An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified off-site Northern States Power Company personnel or an outside fire protection consultant.

B.

An inspection and audit by an outside qualified fire protection consultant shall be performed at intervals no greater than three years.

\\ Nice. hrediled N M toe bwemdN 6.4 Action to be Taken if a Safety Limit is Exceeded If a Safety Limit is exceeded, the reactor shall be shut d immediately. An immediate report shall be made to the Commission and to the Ormer:1 Men ; r _ :12:r "12 te, or his designated y

alternate in his absence. A complete' analysis of the circumstances leading up to and

/

resulting from the situation, together with recommendations by the Operations Committee, shall also by prepared. This report shall by submitted to the Commission, to the C;n;r:1 M;;;g;r L;.;;r I'lauta and the Chairman of the Safety Audit Committee within 14 days of the occurrence.

Reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission.

6.2 - 6.4 243 REV % 3/27/S1

bNoMOVA bef @CC,b

• B.

Radiolocical 1.

a.

A Radiation Protection Program, consistent with the requ of 10 CFR 20, shall be developed and followed. The Radiation Protection Pro shall consist of the following:

(1) A Radiation Protection Plan, which shall be a c lete definition of radiation protection policy and program (2) Procedures which implement the requirements of the R iation Protection Plan The Radiation Protection Plan and implementing proced es, with the exception of those non-safety related procedures governing work activi es exclusively applicable to or performed by health physics personnel, shall b reviewed by the operations Committee and approved by a member of plant management de s nated by the Plant Manager. Health physics procedures not reviewed by the Operatig Committee shall be reviewed and app-roved by the Sup;rintendent, " dictier "rct :: ca.

b.

In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of f

10CFR20, each high radiation area in which the intensity of radiation is greater than 100 g

mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit.1 Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by more of the following:

1) A radiation monitoring device m..

ontinuously indicates the radiation dose rate in ir4'Te5 $^ % R.

the area.

g are hu e, (2) A radiation monitor ng device on inuously integrates the radiation dose rate in g*

the area and alarms when a prc :.. integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose-r t 1 c:1 in th arca h:2 i::n d*

/

a d '_iched and personnel have been made knowledgeable o em.

IS f

(3) An individual qualified in radiation protection procedures d.: i: :cipped with a gg radiation dose rate monitoring device. This individual ch_il be responsible for providing positivW control over the activities within the area and shall perform cGTe d,*

periodic radiation surveillance at the frequency specified by the unit ": lth "h micist-

+A Radiation Work Permit.

. The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr.

In addition doors shall be locked or attended, to prevent unauthorized entry into these areas and the keys or key devices for locked doors shall be maintained under the administrative control of the Plant Manager.

1-Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the Radiation Work Permit issuance requirement during the performance of their assigned

/

radiation protection duties, provided they comply with approved radiation protection procedures

[

for entry into high radiation areas. This footnote applies only to high radiation areas of 1000 mrem /hr or less.

6.5

^T

@b 244a O O The_ o pp\\lco1Q REV 120- 3/2G/ n

i E.

Ot. site Doss Calculation M*nual (ODCM)

The ODCM shall be-approved by the Commission prior to initial implementation. Changes to the ODCM shall satisfy the following requirements:

1.

Shall by submitted to the Commission with the Semi-Annual Radioactive Effluent release report for l

the period in which the change (s) were made effective. This submittal shall contain:

a.

sufficiently detailed information to totally support the rationale for the change without benefit of additionally or supplemental information.

Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with a revision date, together with appropriate analyses or evaluations-justifying the change (s).

l b.

a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c.

documentation of the fact that the change has been reviewed and found acceptable r

by the Operations Committee.

j 2.

Shall become effective upon review and acceptance by the Operations Committee.

f i

F.

Security-l

+

t Procedures shall be developed to implement the requirements of the Security Plan and the Security j

Cor.tingency Plan. These implementing procedures, with the exception of those non-safety related procedures governing work activities exclusively applicable to or performed by security personnel, shall be reviewed by the Operations Committee and approved by a member of plant management designated by.the Plant Manager.

Security procedures not reviewed by the Operations Committee shall be reviewed and approved by the Superintendent, Securitygenddesus6ees.

C.

Temporary Channes to Procedures Temporary changes to those procedures which are required to be reviewed by the Operations Committee described in A, B, C, D, E and F above, which do not change the intent of the original procedures may be made with the concurrence of two members of the unit management staff, at least one of whom holds a j

Senior Operator License.

Such changes should be documented, reviewed by the Operations Committee and

[

]

approved by a member of plant management designated by the Plant Manager within one month. Temporary l

changes to health physics and security procedures not reviewed by the Operations Committee shall be reviewed by the i; ri.._ _ __.... R A

..r P-"'--

for health physics procedures and the Superintendent, 1

Security

'--- for security procedures.b i

G.

a w e er, h % se a es s.

A.

u

B.

Records Retained for Plant Life (continued) 11.

ords fo nvironne al Qual ion w ch are vere the ovisio of-paragra' 6.8 y

J.Y Records of the service lives of all safety-related snubbers, including the date at which

(

the service life conumences and associated installation and maintenance records.

y q'

(7

/

i 4

6.6 247a REV-59 12,'2 9,'? 1

l i

2.

Environmental Special Reports When radioactivity levels in samples exceed limits specified in Table 4.16.3 an Environmental Special Report j

shall be submitted within 30 days from the end of the affected calendar quarter.

For certain cases i

involving long analysis time, determination of quarterly averages may extend beyond the 30 days period.

f In these cases the potential for exceeding the quarterly limits will be reported within the 30 day period to be followed by the Environmental Special Report as soon as practicable.

l 3.

Other Environmental Reports (non-radiological. non-aaustici l

a.

Environmental events that indicate or could result in a significant environmental impact

[

causally related to plant operation. The following are examples-xcessi a hird ir;- -

j tion; onsite plant or animal disease outbreaks; unusual mer any species protecte

[

by Endangered Species Act of 1973; increase in nuisance organisms or conditions; or TfhD P'IO' sry j

excessive environmental impact caused by herbicide application to transmission corridors

~

associated with the plant.

This report shall be submitted within 30 days of the avant

+

and shall (a) describe, analyze, and evaluate the event, including extent and magnitude i

of the impact and plant operating characteristics, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses.

f b.

Proposed changes, tests or experiments which may result in a significant increase in any adverse environmental impact which was not previously reviewed or evaluated in the Final Environmental Statement or supplements thereto. This report shall include an evaluation j

of the environmental impact of the proposed activity and shall be submitted 30 days prior to implementing the proposed change, test or experiment.

7 D.

Special Reports

.i Unless otherwise indicated, speciat reports required by the Technical Specification shall be

[

g/'

submitted within the time period specified for each report, b

6.7 252 REV 109 2/15/S9 l

m.

L.

m

-- m- -

Exhibit C i

Monticello Nuclear Generating Plant License Amendment Reauest Dated Auaust 15.1996 Revised Monticello Technical Specification Pages l

Exhibit C consists of revised Technical Specification pages that incorporate the proposed changes. The pages included in this exhibit are as listed below.

Pace iv vi vii

)

22 31 69 i

89 99 102 164 188 190 198t 198y 200 227c 227d 227e i

229b 229i 229ff j

232 233 234 240 241 243 244a 246b 247a 252 C-1

.~

f.aER 3.13 end 4.13 Fira Datsetion Protection Systams 223 1

A.

Fire Detection Instrumentation 223 l

B.

Fire Suppression Water System 224 i

C.

Hose Stations 226 D.

Yard Hydrant Hose Houses 227 E.

Sprinkler Systems 227a F.

Halon Systems 227b G.

Penetration Fire Barriers 227b H.

Alternate Shutdown System 227c i

3.13 Bases 228

{

4.13 Bases 228b

<l 3.14 and 4.14 Accident Monitoring Instrumentation 229a l

l 3.15 and 4.15 Inservice Inspection and Testing 229f 3.15 and 4.15 Bases 229g 3.16 and 4.16 Radiation Environmental Monitoring Program 229h A.

Sampling and Analysis 229h B.

Land Use Census 229j C.

Interlaboratory Comparison 229k i

3.16 and 4.16 Bases 229t 3.17 and 4.17 Control Room Habitability 229u A.

Ventilation System 229u B.

Emergency Filtration System 229v 3.17 Bases 229y 4.17 Bases 229z 5.0 DESIGN FEATURES 230 5.1 Site 230 5.2 Roactor 230 5.3 Reactor Vessel 230 5.4 Containment 230 5.5 Fuel Storage 231 5.6 Seismic Designs 231 6.0 ADMINISTRATIVE CONTROLS 232 i

6.1 Organization 232 6.2 Review and Audit 237 6.3 Special Inspection and Audits 243 6.4 Action to be taken if a Safety Limit is Exceeded 243 l

6.5 Plant Operating Procedures 244 6.6 Plant Operating Records 246c l

6.7 Reporting Requirements 248 6.8 Environmental Qualification 253 l

iv REV

LIST OF TABLES Table No.

Eggg 3.1.1 Reactor Protection System (St. caw) Instrument Requirements 28 4.1.1 Scram Instrument Functional Tests - Minimum Functional 32 Test Frequencies for Safety Instrumentation and Control Circuits 4.1.2 Scram Instrument Calibration - Minimum Calibration 34 Frequencies for Reactor Protection Instrument Channels 3.2.1 Instrumenration that Initiates Primary Containment 49 Isolation Functions 3.2.2 Instrumentation that Initiates Emergency Core Cooling Systems 52 3.2.3 Instrumentation that Initiates Rod Block 56 l

3.2.4 Instrumentation that Initiates Reactor Building Ventilation 59 Isolation and Standby Cas Treatment System Initiation 3.2.5 Instrumentation that Initiates a Recirculation Pump Trip 60 and Alternate Rod Injection 3.2.6 Instrumentation for Safeguards Bus Degraded Voltage and 60a Loss of Voltage Protection 3.2.7 Instrumentation for Safety / Relief Valve Low-Low Set Logic 60b 3.2.8 Other Instrumentation 60d 3.2.9 Instrumentation for Control Room Habitability Protection 60e 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod 61 Block and Isolation Instrumentation 4.6-1 Snubber Visual Inspection Interval 132a 3.8.1 Radioactive Liquid Effluent Monitoring Instrumentation 1981 4.8.2 Re.dioactive Gaseous Effluent Monitoring Instrumentation 198k 4.8.1 Radioactive Liquid Effluent Monitoring Instrumentation 198m Surveillance Requirements 4.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation 198n Surveillance Requirements 4,8.3 Radioactive Liquid Wasta Sampling and Analysis Program 198p 4.8.4 Radioactive Gaseous Waste Sampling and Analysis Program 198s vi REV

l LIST OF TABLES (continued)

Table No.

Eggg 3.13.1 Safety Related Fire Detection Instruments 227d l

3.14.1 Instrumentation for Accident Monitoring 229b i

4.14.1 Minimum Test and Calibration Frequency for Accident 229d j

Monitoring Instrumentation 4.16.1 Radiation Environmental Monitoring Program (REMP) 229-1 l

Sample Collection and Analysis 1

l 4.16.2 REMP - Maximum Values for the Lower Limits of Detection 229q 4.16.3 REMP - Reporting Levels for Radioactivity Concentrations 229s i

in Environmental Samples 1

I 6.1.1 Minimum Shift Crew Composition 236 vii REV

l Bases:

2.2 The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products.

It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

The pressure safety limit of 1335 psig as measured in the vessel steam space is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The 1375 psig value was derived from the design pressures of the reactor pressure vessel, coolant piping, and recirculation pump casing.

The respective design pressures are 1250 psig at 575*F,1148 psig at 562*F, and 1380 psig at 575"F.

l The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes: ASME Boiler and Pressure Vessel Code Section III-A for the pressure vessel, ASME Boiler and Pressure Vessel Code Section III-C for the recirculation pump casing, and the USAS Piping Code Section B31.1 for the reactor coolant system piping. The ASME Code permits pressure transients up to 10 percent over the vessel design pressure (110% x 1250 - 1375 psig) and the USAS Code permits pressure transients up to 20 parcent over the piping design pressure (120% x 1148 -

1378 psig).

The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig and temper-ature of 575*F; this is more than a factor of 1.5 below the yield strength of 42,300 psi at this temperature. At the pressure limit of 1375 psig, the general membrane stress increases to 29,400 psi, still safely below the yield strength.

The reactor coolant system piping provides a comparable margin of protection at the established pressure safety limit.

2.2 BASES 22 REV

1 Table 3.1.1 - Continued The high drywell pressure scram functions in the Startup and Run modes when necessary during purging for e.

containment inerting or de-inerting only by closing the manual containment isolation valves.

of the bypass condition shall be noted in the control room log.

Verification f.

One instrument channel for the functions indicated in the table to allow completion of surveillance testing, provided that:

1.

Redundant instrument channels in the same trip system are capable of initiating the automatic function bypass.and are demonstrated to be operable either immediately prior or immediately subsequent to applying the 2.

While the bypass is applied, surveillance testing shall proceed on a continuous basis and the remaining instrument channels initiating the same function are tested prior to any other.

Upon completion of surveillance testing, the bypass is removed.

3.1/4.1 31 REV l

Bases Continued:

3.2 increases core voiding, a negative reactivity feedback. High pressure sensors initiate the pump trip in the event of an isolation transient. Iow level sensors initiate the trip on loss of feedwater (and the resulting MSIV closure). The recirculation pump trip is only required at high reactor power levels, where the safety / relief valves have insufficient capacity to relieve the steam which continues to be generated after reactor isolation in this unlikely postulated event, requiring the trip to be operable only when in the RUN mode is therefore conservative.

The ATVS high reactor pressure and low-low water level logic also initiates the Alternate Rod Injection System. Two solenoid valves are installed in the scram air header upstream of the hydraulic control units. Each of the two trip systems energizes a valve to vent the header and causes rod insertion.

This greatly reduces the long term consequences of an ATVS event.

Voltage sensing relays are provided on the safeguards bus to transfer the bus to an alternate I

source when a loss of voltage condition or a degraded voltage condition is sensed. On loss of voltage this transfer occure immediately. The transfer on degraded voltage has a time delay to prevent transfer during the starting of large loads. The degraded voltage setpoint l

corresponds to the minimum acceptable safeguards bus voltage for a steady state IDCA load that maintains adequate voltage at the 480 V essential MCCs. An allowance for relay tolerance is included.

[

Safety / relief valve low-low set logic is provided to prevent any safety / relief valve from i

opening when there is an elevated water leg in the respective discharge line. A high water leg is formed immediately following valve closure due to the vacuum formed when l

steam condenses in the line.

If the valve reopens before the discharge line vacuum breakers act to return water level to normal, water clearing thrust loads on the discharge line may exceed their design limit. The logic reduces the opening setpoint

[

and increases the blowdown range of three non-APRS valves following a scram. A 15-second interval between subsequent valve actuations is provided assuming one valve fails to i

3.2 BASES 69 REV

...m.

-___.__....-.--_--_..~._.m.

. ~.. _. ~.. _.. _.

Bases Continued 3.3 and 4.3:

[

consequences of reactivity accidents are functions of the initial neutron flux. The require-ment of at least 3 counts per second assures that any transient, should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from cG d conditions. One operable SRM channel would be adequate to monitor the approach to criticality T

using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's are provided as an added conservatism.

C.

Scram Insertion Times The control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than the Safety Limit (T.S.2.1.A).

This requires the negative reactivity insertion in any local region of the core and in the overall core to be equivalent to at least the scram reactivity curve used in the transient analysis. The required average scram times for three control rods in all two by two arrays and the required average scram times for all control rods are based on inserting this amount of negative reactivity at the specified rate locally and in the overall core. Under these conditions, the CPR safety limit is never exceeded during any transient requiring control rod scram, and therefore MCP~L remains above the Safety Limit (T.S.2.1.A).

3.3/4.3 BASES 89 REV

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

t Basis 3.4 and 4.4:

I A. The design objective of the standby liquid control system is to provide the capability of bringing, the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted. To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of boron in the reactor core in less than 125 minutes sufficient to bring the reactor from full power to a 3% delta k suberitical condition considering the hot to cold reactivity swing, xenon poisoning and an additional 25% boron concentration margin to allow for leakage and imperfect mixing.

The time requirement (125 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak.

The AWS Rule (10CFR50.62) requires the addition of a new design requirement to the generic SLC System design basis. Changes to flow rate, solution concentration or boron enrichment to meet the A WS Rule do not invalidate the original system design basis.

Paregraph (c)(4) of lOCHt50.62 states that:

"Each boiling water reactor must have a Standby Liquid Control System (SILS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution" (natural boron enrichment).

The described minimum system parameters (equivalent to 24 gpa, 10.7% concentration and 55 atom t

percent Boron-10 enrichment) will ensure an equivalent injection capability that meets the ATWS rule requirement.

Boron enrichment concentration, solution temperature, and volume (including check of tank heater and pipe heat tracing system) are checked on a frequency to assure a high reliability of operation of the i

system should it ever be required. Only one of the two standby liquid control pumping circuits is needed for proper operation of the system.

If one pumping circuit is found to be inoperable, there r

is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made. A reliability. analysis indicates that the plant can be operated safely in this manner for ten days.

For additional margin, the allowable out of service time has been reduced to seven j

days.

The only practical time to test the standby liquid control system is during a refueling outage and by initiation from local stations. Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refueling outage unnecessary. A test of explosive charges from one manufacturing batch is made to assure that the i

replacement charges for the tested system are satisfactory. A continual check of the firing circuit continuity is provided by pilot lights in the control room.

f The relief valves in the standby liquid control system protect the system piping and positive displacement pumps which are nominally designed for 1500 psi from overpressure. '1he pressure relief valves discharge back to the standby liquid control solution tank.

3.4/4.4 BASES 99 REV

3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 3.

One of the following conditions of 4.

Perform the following tests:

inoperability may exist for the period specified:

Item Frecuency a.

One Core Spray subsystem may be inoperable Motor Operated Pursuant to for 7 days, or Valve Operability Specification b.

One RHR pump may be inoperable for 30 days, U#

ADS Valve Each Operating c.

One low pressure pump or valve (Core Spray Operability Cycle or RHR) may be inoperable with an ADS valve inoperable for 7 days, or Note: Safety / relief valve operability is d.

One of the two LPCI injection paths may be verified by cycling the valve and observing inoperable for 7 days, or a compensating change in turbine bypass or control valve position.

Two RHR pumps may be inoperable for 7 days, e.

or ADS Inhibit Each Operating Switch Operability Cycle f.

Both of the LPCI inj.ection paths may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or Perform a simulated Each Operating g.

HPCI may be inoperable for 14 days, automatic actuation test Cycle provided RCIC is operable, or (including HPCI transfer to e

P es on po aM h.

One ADS valve may be inoperable for 14 days, or subsequent low reactor 1.

Two or more ADS valves may be inoperable water level) for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

5.

Perform the following test on the Core 4.

If the requirements or conditions of 3.5. A.l.

Spray Ap Instrumentation:

2 or 3 cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor Check Once/ day shall be placed in a condition in which the affected equipment is not required to be Test Once/ month operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Calibrate Once/3 months 3.5/4.5 102 Rev.

l l

1 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 4.

Pressure Suppression Chamber-Drywell Vacuum 4.

Pressure Suppression Chamber-Drywell Vacuum Breakers Breakers l

a.

When primag containment integrity is a.

Operability and full closure of the required, all eight drywell-suppression drywell-suppression chamber vacuum breakers chamber vacuum breakers shall be operable shall be verified by performance of the and positioned in the closed position as following:

indicated by the position indication system, except during testing and except as (1)

Monthly each operable drywell-specified in 3.7.A.4.b through 3.7.A.4.d suppression chamber vacuum breaker below.

shall be exercised through an opening-closing cycle.

b.

Any drywell-suppression chamber vacuum breaker may be nonfully closed as indicated (2)

Once each operating cycle, dry-by the position indication and alarm system well to suppression chamber provided that drywell to suppression leakage shall be demonstrated to chamber differential pressure decay does be less than that equivalent to a not exceed that shown on Figure 3.7.1 c te-inch diameter orifice and each vacuum breaker shall be visually c.

Up to two drywell-suppression chamber inspected.

(Containment access vacuum breakers may be inoperable provided required) that: (1) the vacuum breakers are determined to be fully closed and at least (3)

Once each operating cycle, vacuum one position alarm circuit is operable or breaker position indication and (2) the vacuum breaker is secured in the alarm systems shall be calibrated closed position or replaced by a blank and functionally tested, flange.

(Containment access required) d.

Drywell-suppression chamber vacuum breakers (4)

Once each operating cycle, the may be cycled, one at a time, during vacuum breakers shall be tested to containment inerting and deinerting determine that the force required operations to assist in purging air or to open each valve from fully nitrogen from the suppression chamber vent closed to fully open does not header.

exceed that equivalent to 0.5 psi acting on the suppression chamber face of the valve disc.

(Containment access required) 3.7/4.7 164 REV

f l

r Bases Continued:

r B. Standby Cas Treatment System and C.

Secondary Containment I

' Initiating reactor building isolation and operation of the standby gas treatment system tu maintain the design l

negative pressure within the secondary containment provides an adequate test of the reactor building isolation t

valves and the standby gas treatment system.

Periodic testing gives sufficient confidence cf reactor building i

integrity and standby gas treatment system operational capability. Secondary Containment Capability Test data

[

obtained under non-cala conditions is to be extrapolated to calm wind conditions using information provided in

.t

" Summary Technical Report to the United States Atomic Energy Commission, Directorate of Licensing, on Secondary l

Containment Ieak Rate Test", submitted by letter dated July 23, 1973, and as described in NSP letter to the NRC l

dated August 18, 1995, with subject, " Revision 2 to License Amendment Request Dated June 8,1994, Standby Cas Treatment and Secondary Containment Technical Specifications."

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers l

can perform as evaluated.

Standby gas treatment system inplace testing procedures will be established utilizing applicable sections of ANSI N510-1989 standard as a procedural guideline only.

If painting, fire, or chemical l

release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals, or foreign materials, the same tests and sample analysis should be performed as required for operational use.

Replacement adsorbent should be qualified according to the guidelines of Regulatory Guide 1.52 l

Revision 2 (March 1978), except testing should be IAW ASTM D3803-1989. The charcoal adsorber efficiency test l

}

procedures will allow for the removal of a representative sample. The 30*C, 954 relative humidity test per ASTM D 3803-89 is the test method to establish the methyl iodine removal efficiency of the adsorbent. The sample

'}'

will be at least two inches in diameter and a length equal to the thickness of the bed.

If the iodine removal efficiency test results are unacceptable, all adsorbent in the system will be replaced. High efficiency i

particulate filters are installed before and after the charcoal filters to prevent clogging of the carbon j

adsorbers and to minimize potential release of particulates to the environment. An efficiency of 994 is-l adequate to retain particulates that may be released to the reactor building following an accident. This will i

be demonstrated by inplace testing with DOP as the testing medium. Any HEPA filters found defective will be l

replaced with filters qualified pursuant to regulatory guide position C.3.d of Regulatory Guide 1.52 Revision 2 (March 1978). Once per operating cycle demonstration of HEPA filter pressure drop, operability of inline i

heaters at rated power, automatic initiation of each standby gas treatment system circuit, and leakage tests f

after maintenance or testing which could affect leakage, is necessary to assure system performance capability.

i I

l 4.7 BASES 188 REV 44 10/2/95

Bases Continued:

The containment is penetrated by a large number of small diameter instrument lines. A program for the periodic testing (see Specification 4.7.D) and examination of the valves in these lines has been developed and a report covering this program was submitted to the AEC on July 27, 1973.

l The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability.

E.

Combustible Gas Control System The Combustible Gas Control System (CGCS) is functionally tested once every six months to ensure that the recombiner trains will be available if required.

In addition, calibration and maintenance of essential components is specified once each operating cycle.

4.7 BASES 190 REV

TABLE 4.8.4

- RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM (continued)

(Page 2 of 2)

Notes:

a.

The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95%

probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

Note (a) of Table 4.8.3 is applicable.

b.

Grab samples taken at the discharge of the plant stack and reactor building vent are generally below minimum detectable levels for most nuclides with existing analytical equipment.

For this reason, isotopic analysis data, corrected for holdup time, for samples taken at the steam jet air ejector may be used to calculate noble gas ratios Whenever the steady state radioiodine concentration is greater than 10 percent of the limit of Specification c.

3.6.C.1, daily sampling of reactor coolant for radioactive iodines of I-131 through I-135 is required. Whenever a change of 25% or more in calculated Dose Equivalent I-131 is detected under these conditions the iodine and particulate collection devices for all release points shall be removed and analysed daily untii it is shown that a pattern exists which can be used to predict the release rate.

Sampling may then revert to weekly.

When samples collected for one day are analyzed, the corresponding LLD's may be increased by a factor of 10.

Samples shall be analyzed within 98 hours0.00113 days <br />0.0272 hours <br />1.62037e-4 weeks <br />3.7289e-5 months <br /> after removal.

d.

To be representative of the average quantities and concentrations of radioactive materials in particulate form in gaseous effluents, samples should be collected in proportion to the rate of flow of the effluent streams.

The principal gamma emitters for which the LLD specification will apply are exclusively the following radio-e.

nuclides:

Kr-87, Kr-88, Xe-133, Xe-133m Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported, f.

Nuclides which are below the LLD for the analyses shall be reported as "less than" the LLD of the nuclide and should not be reported as being present at che LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations. When unusual circumstances result in LLD's higher than reported, the reasons shall be documented in the semiannual effluent report.

g.

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period sampled.

h.

H3 analysis shall not be required prior to venting if the limits of 3.8.B.1 are satisfied for other nuclides.

The analysis shall be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after sampling, however.

1.

In lieu of grab samples, continuous monitoring with weekly analysis using silica-gel samplers may be provided.

l 3.8/4.8 198t REV l

~

3.8 and 4.8 Bases: (continued)

Specification 3.8.B.4.c is provided to ensure that the concentration of potentially explosive gas mixtures Automatic control features are included ln the system to prevent the hydrogen and oxygen concen reaching these flammability limits.

Maintaining the concentration of hydrogen below the finmmability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the require-ments of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

Specification 3.8.B.4.e is provided to limit the radioactivity which can be stored in one decay tank.

the quantity of radioactivity c ntained in each gas storage tank provides assurance that in the event of anRestricting uncontrolled release of the tanks contents the r restricted area bounda g will not exceed 26 mrem.esulting total body exposure to an individual at the site prevents a tank from being discharged at an uncontrolled rate.A flow restrictor in the discharge line of the decay tanks tank from being released with less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of holdup.

In addition, interlocks prevent the contents of a l

Specification 3.8.B.5 establishes a maximum activity at the steam jet ai; activity rate of noble gases form the main condenser provides reasonable assurance that the total body exposureRestricting the aector.

to an individual at the restricted area boundary will not exceed the limits of 10 CFR Part 20 in the event this effluent is inadvertently discharged directly to the environment with minimal treatment.

implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50This specification Specification 3.8.B.6 requires the containment to be pur system except during inerting and deinerting operations.ged and vented through the standby gas treatment from the containment atmosphere.

During outages when the containmentThis provides for lodine and particulate removal is opened for maintenance, the containment ventilation exhaust is directed to the monitored reactor building vent.

flow path prevents damage to the standby gas treatment system in the event of a loss of coolant accident Use of the 2-inch during purging or venting.

Use of the reactor building plenum and vent flow path for inerting and deinerting operations permits the control room operators to monitor the activity level of the resulting effluent by use of the Reactor Building Vent Wide Range Gas Monitors.

In addition, Plenum Monitors will automatically terminate releases if their release rate limits are exceeded.the Reactor Building event that the reactor building release rate exceeds the Reactor Building Vent Wide Range Cac Monitor In the alarm settings, the release of gaseous radioactive effluents.the monitors will alarm in the control room alerting the operators to take actions to limit control room operators would have in excess of two hours to take manual actions to terminate releasesAn analysis has been perfo without exceeding the permissible levels of radiation exposure in 10 CFR Part 20 Section 20.105(a).

NRC approved methods in the ODCM to ensure that alarms will alert control room operators prior The limits of 10CFR Part 20, Section 20.105(a) being exceeded C.

Solid Radioactive Waste Specification 3.8.C.1 provides assurance that the solid radwaste system will be used whenever solid radwastes require processing and packaging prior to being shipped offsite.

This specification im requirements of 10 CFR Part 50.36a and General Design Criteria 60 of Appendix A to 10 CFR Part 50.plements the 3.8/4.8 BASES 19p 1

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

3.0 LIMITING CONDITIONS ftR OPERATION 4.0 SURVEILIANCE REQUIREMENTS I

2.

Both diesel generators are operable and capable of feedb+g their designated 4160 volt buses.

3.

(a) 4160V Buses #15 and #16 are energized.

(b) 480V Ioad Centers #103 and #104 are energized.

4.

All station 24/48, 125, and 250 volt batteries are charged and in service, and associated battery chargers are

'>perable.

B.

When the

--A switch is in Run, the availabilit.y of electric power shall be as specified in 3.9.A. except as specified in g

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

1.

Transmission Lines From and after the date that incoming power is available from only one line, reactor operation is permissible only during the succeeding seven days unless an additional line is sooner placed in 3.9/4.9 200 REV

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS H.

Alternate Shutdown System H.

Alternate Shutdown System 1.

Switches on the alternate shutdown system 1.

The system controls on the ASDS panel shall be operable whenever that system /

Panel shall be functionally tested once component is required to be operable.

Per operating cycle.

Thealternateshutdownsystem$anelmaster 2.

2.

If system controls required to be operable transfer switch shall be verif ed to alarm by Specification 3.13.H.1 are made of in the control room when unlocked once per found inoperable, restore the inoperable operating cycle system control to operable within 7 days, or perform one of the following:

a.

Provide equivalent shutdown capability and within 60 days restore the inoperable system controls to operable; or b.

Establish a continuous fire watch in the cable apreading room and the back-panel area of the control room and wit' In 60 days restore the inoperable sys.om controls te operable; or c.

Verify the operability of the fire detectors in the cable spreading room and the back-panel area of the control room and establish a hourly fire watch patrol and within 60 days restore the inoperable system controls to opero.dle; or d.

Place the reactor in a condition where the systems for which the system controls at the ASDS are inoperable are not required to be operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

The alternate shutdown system panel master transfer switch shall be locked in the normal position except when in use, being test.ed or being maintained.

227c REV 3.13/4.13

TABLE 3.13.1 SAFETY REIATED FIRE DETECTION INSTRUMENTS Fire l

Detection Minimum Instruments Operable Zone Location Heat Flame Smoke 1A "B" RHR Room 3

1B "A" RHR Room 3

1C RCIC Room 3

l 1E HPCI Room 2

1F Reactor Building-Torus Compartment 11 l

2A Reactor Bldg. 935' elev - TIP Drive Area 1

2B Reactor Bldg. 935' elev - CRD HCU Area East 10 2C Reactor Bldg. 935' elev - CRD HCU Area West 11

)

2G/2H Reactor Bldg. 935' - LPCI Injection Valve Area 1

3B Reactor Bldg. 962' elev - SBLC Area 2

3C Reactor Bldg. 962' elev - South 5

3D Reactor Bldg. 962' elev - RBCCW Pump Area 4

4A Reactor Bldg. 985' elev - South 4

4B Reactor Bldg. 985' elev - RBCCW Hx Area 5

4D SBGT System Room 2

SA Reactor Bldg. 1001' elev - South 7

SB Reactor Bldg. 1001' elev - North 3

i SC Reactor Bldg. - Fuel Fool Cooling Pump Area 1

6 Reactor Building 1027' elev 5

[

7A Battery Room 1

t 7B Battery Room 1

7C Battery Room 1

r 8

Cable Spreading Room 7

3.13/4.13 227d REV l

TABLE 3.13.1 SAFETY REIATED FIRE DETECTION INSTRUMENTS ~

Fire l

Detection Minimum Instruments Operable Zone Location iltal Flame Smoke 12A Turbine Bldg. - 911' - 4.16 KV Switchgear 3

13C Turbine b1dg. - 911' evel - MCC 133 Area 1

14A Turbine Bldg. - 931' - 4.16 KV Switchgear 2

15A/15C

1. 12 DG Room & Day Tank Room 3

15B/15D

1. 11 DG Room & Day Tank Room 3

16 Turbine Bldg. 931' elev - Cable Corridor 3

17 Turbine Bldg. 941' elev - Cable Corridor 3

19A Turbine Bldg. 931' elr* - Water Treatment Area 5

19B Turbine Bldg. 931' elev - MCC 142-143 Area 1

19C Turbine Bldg. 931' elev - W Pipe Chase 1

20 Heating Boiler Room 1

23A Intake Structure Pump Room 3

31A ist Floor - Reactor Building Addition - Division I 3

31B ist Floor - Reactor Building Addition - Division II 15 32A 2nd Floor - Reactor Building Addition - Division I 6

32B 2nd Floor - Reactor Building Addition - Division II 4

33 3rd Floor - Reactor Building' Addition 5

3.13/4.13 227e REV

I i

i Table 3.14.1 Instrumentation for Accident Monitoring Function Total No. of Minimum No. of Required Instrument Channels Operable Channels Conditions

• i Reactor Vessel Fuel Zone Water 14 vel 2

1 A, B Safety / Relief Valve Position 2

1 A, C (One Channel Pressure Switch and One Channel Thermocouple Position Indication per Valve)

Drywell Wide Range Pressure 2

1 A, 8 Suppression Pool Wide Range Imvel 2

1 A, B Suppression Pool Temperature 2

1 A, D Drywell High Range Radiation 2

1 A, D Drywell and Suppression Pool 2

1 AB Hydrogen and Oxygen Monitor Offgas Stack Wide Range Radiation 2

1 A, D Reactor Bldg Vent Wide Range Radiation 2

1 A, D

• Required Conditions A. When the number of channels made or found to be inoperable is such that the number of operable channels is less than the total number of channels, either restore the inoperable channels to operable status within seven days, or prepare and submit a special report to the Commission pursuant to Technical i

Specification 6.7.D within the next 30 days outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status.

3.14/4.14 229b REV 4

....._y m

. -., - -,,, ~ -

r-,---

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 2.

Deleted.

k B.

Inservice Testing 1.

Inservice Testing of Quality Group A, B, and C pumps and valves shall be performed in accordance with the requirements for ASME Code Class 1, 2 and 3 pumps and valves, respectively, contained in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CR 50, Section 50.55a(g) except where relief has been granted by the Commission pursuant to 10 c m 50, Section 50.55(a)(g)(6)(1), or where alternate testing is justified in accordance with Generic Letter 89-04, 2.

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

l 3.15/4.15 229ff REV

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 3.

Deviations are permitted from the required sampling schedule if samples are unobtainable due to hazardous conditions, seasonable unavailability, or to malfunction of automatic sampling equipment. If the latter occurs, every effort shall be made to complete corrective action prior to the end of the next sampling period.

4.

With the level of radioactivity in an environmental nampling medium exceeding the reporting levels of Table 4.16.3 when averaged over any calendar quarter, submit a special report to the Commission within 30 days from the end of the affected calendar quarter pursuant to Specification 6.7.C.2.

When more l

than one of the radio-nuclides in Table 4.16.3 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2) limit level (1)

+

limit level (2) +...> 1.0 When radionuclides other than those in Table 4.16.3 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dore to an individual is equal to or greater than the calendar year limits of Specifications 3.8.A.2, 3.8.B.2, or 3.8.B.3.

This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiation Environmental Monitoring Report.

2291 REV

t 6.0 ADMINISTRATIVE CONTROLS 6.1 Organize. ion 4

A. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for the safe operation and maintenance of the plant.

I During periods when the Plant Manager is unavailable, this responsibility may be delegated to other qualified supervisory personnel.

7 The Shift Supervisor (or, a designated individual during periods of absence from the control l

room and shift supervisor's office) shall be responsible for the control room command function.

[

B. Offsite and Onsite Organizations

}

Onsite and offsite organizations shall be established for plant operation and corporate i

management, respectively. The onsite and offsite organizations shall include positions for activities affecting plant safety.

i 1.

Lines of authority, responsibility and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating

[

organization positions. These relationships shall be documented and updated, as appropriate,.

in the form of organization charts, function descriptions of department responsiblities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements are documented in corporate and plant procedures, or the l

Updated Safety Analysis Report or the Operational Quality Assurance Plan.

2.

The Vice President Nuclear Generation shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining and providing technical support to the plant to ensure nuclear safety. This position has the responsiblity for the Fire Protection Program.

1 3.

The individuals who train the operating staff and those who carry out health physics and j

quality assurance functions may report to the appropriate onsite manager; however, they shall i

have sufficient organizational freedom to ensure their independence from operating pressures.

{

1 t

6.1 232 REV 1

i

...- m..

.m.

+ -

w r

~ --

i C.

Plant Staff 1.

Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.1.1.

2.

At least one licensed operator shall be in the control room when fuel is in the l

reactor.

j 3.

At least two licensed operators shall be present in the control room during cold startup, scheduled reactor shutdown, and during recovery from reactor trips.

4.

An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.

5.

All alterations of the reactor core shall be directly supervised by a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

l 6.

A fire brigade of at least five members shall be maintained on site at all times.* The fire brigade shall not include the three members of the shift organization required for safe shutdown of the reactor-from outside the control room.

7.

The General Superintendent. Operations shall be formerly licensed as a Senior Reactor f

Operator or hold a current Senior Reactor Operator License.

[

8.

At least one member of plant management holding a current Senior Reactor Operator License j

shall be assigned to the plant operations group on a long term basis (approximately two i

years). This individual will not be assigned to a rotating shift.

D.

Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the General Superintendent Radiation Services who shall meet l

or exceed the qualifications of Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents, and (3) the General Superintendent, Operations who shall meet the

]

requirement of ANSI N18.1-1971 except that NRC license requirements are as specified in Specification 6.1.C.7.

The training program shall be under the direction of a designated member of Northern States Power management.

l

• Fire-Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of Fire Brigade members provided immediate action is taken to restore the Fire Brigade to within the minimum requirements.

I i

6.1 233 REV

~- -

E.

A training program for individuals serving in the. fire brigade shall be maintained under the direction of a designated member of Northern States Power management. This program shall meet the requirement of Section 27 of the NFPA Code - 1976 with the exception of training scheduling.

Fire brigade training shall be scheduled as set forth in the training program.

F.

Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g.,

senior reactor operators, reactor operators, i

health physicists, auxiliary operators, and key maintenance l

personnel. Procedures shall include the following provisions:

1.

Adequate shift coverage shall be unintained without routine heavy use of overtime. The objective shall be to have operating i

personnel work a normal 8 or 12-hour day, nominal 40-hour week g-while the plant is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:

a.

An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time.

b.

Overtime should be limited for all nuclear plant staff personnel so that total work time does not exceed 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour. period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, not more than 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> in any seven day period, all excluding shift turnover time.

Individuals should net be required to work more than 15 consecutive days without two consecutive days off.

c.

A break of at least eigilt hours including shift turnover time should be allowed between work periods.

d.

Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

6.1 234 REV

7.

Authority The SAC shall be advisory to the Vice President, Nuclear Generation.

8.

Records Minutes shall by prepared and retained for all scheduled meetings of the Safety Audit Committee. The minutes shall be distributed within one month of the meeting to the Vice President, Nuclear Generation, the Plant Manager, each member of the SAC, and l

others designated by the Chairman or Vice Chairman. There shall be a formal approval of the minutes.

9.

Procedures A written charter for the SAC shall be prepared that contains:

a.

Subjects within the purview of the group.

b.

Responsibility and authority of the group.

c.

Mechanisms for convening meetings.

d.

Provisions of use of specialists or subgroups.

e.

Authority to obtain access to the nuclear power plant operating record files and operating personnel when assigned audit functions.

f.

Requirements for distribution of reports and minutes prepared by the group to others in the NSP Organization.

6.2 240 REV

j B.

Operations Co-ittee (OC) 1.

Membership The Operations Committee shall consist of at least six (6) regular members drawn from the key super-visors of the on-site supervisory staff. The Plant Manager shall serve as Chairman of the OC and shall appoint a regular member to act as Vice Chairman in his absence. Alternates to the regular members shall be designated in writing by the Chairman, or Vice Chairman in the Chairman's absence, to serve on a temporary basis. No more than two alternates shall participate as voting members of-the Operations Committee at any one time.

2.

Meeting Frequency The Operations Committee will meet on call by the Chairman or as requested by individual members and at least monthly.

3.

Quorum A quorum shall include a majority of the membership, including the Chairman or Vice Chairman.

4.

Responsibilities - The following subjects shall by reviewed by the Operations Committee:

a.

Proposed tests and experiments and their resules, i

b.

Modifications to plant systems or equipment as described in the Updated Safety Analysis Report and having nuclear safety significance or which irvolve an unreviewed safety question as defined in 10 CFR 50.59.

c.

Proposals which would effect permanent changes to normal and emergency operating procedures and any other proposed changes or procedures that are determined by the Plant Manager to affect nuclear safety.

d.

Proposed changes to the Technical Specifications or operating license.

e.

All reported or suspected violations of Technical Specifications, operating license requirements, administrative procedure::, or operating procedures.

Results of investi-gations, including evaluation and recommendations to prevent recurrence, will be reported, in writing, to the Vice President Nuclear Generation and to the Chairman g

of the Safety Audit Committee.

i 6.2 241 REV

.m m

i i

i c.

Mechanism for scheduling meetings j

d.

Meeting agenda e.

Use of subcommittee s

f.

Review and approval, by members, of OC actions i

i g.

Distribution of minutes 6.3 Soecial Inspections and Audits I

A.

An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either quclified off-site Northern States Power i

company personnel or an outside fire protection consultant.

j B.

An inspection and audit by an outside qualified fire protection consultant shall

[

be performed at intervals no greater than three years.

6.4 Action to be Taken if a Safety Timit is Exceeded If a Safety Limit is exceeded, the reactor shall be shut down immediately. An immediate report j

shall be made to the Commission and to the Vice President Nucitar Generation, or his designated dj alternate in his absence. A complete analysis of the circumstances leading up to and resulting from the situation, together with recommendations by the Operations Committee, shall also be prepared. This report i

shall by submitted to the Commission, to the Vice President Nuclear Generation and the Chairman of the Safety

}l Audit Comunittee within 14 days of the occurrence.

Reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission, t

t t

6 T

6.2 - 6.4 243 l

B.

Radiological i

1.

a.

A Radiation Protection Program, consistent-with the requirements of 10 CFR 20, shall be developed and followed. The Radiation Protection Program shall consist of the following:

t (1) A Radiation Protection Plan, which shall be a.'omplete definition of i

radiation protection policy and program l

l (2) Procedures which implement the requirements of the Radiation Protection Plan l

The Radiation Protection Plan end implementing procedures, with the exception of those non-safety related procedures governing work activities exclusively applicable to or performed by health physics personnel, shall be reviewed by the Operations Committee j

and approved by a member of plant management designated by the Plant Manager. Health physics procedures not reviewed by the Operations Committee shall be reviewed and app-l l

roved by the General Superintendent, Radiation Services.

l b.

In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of k

10CER20, each high radiation area in which the intensity of radiation is greater than 100 mres/hr but less than 1000 aren/hr shall be barricaded and conspicuously posted as a high

[

i radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit.1 Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

(1) A radiation monitoring device that continuously indicates the radiation dose rate in l

the area.

i, (2) A radiation monitoring device that continuously integrates the radiation dose rate in 1

the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rates in the area have been determined and personnel have been made knowledgeable of them.

[

t (3) An individual qualified in radiation protection procedures with a radiation dose rate monitoring device. This individual is responsible for_providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the radiation protection procedures 1

l or the applicable Radiation Work Permit.

The above procedure shall also apply to each high radiation area in which the intensity of i

c.

radiation is greater than 1000 area /hr.

In addition doors shall be locked or attended, to prevent unauthorized entry into these areas and the keys or key devices for locked doors shall be maintained under the administrative control of the Plant Manager.

1* Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from f

the Radiation Work Permit issuance requirement during the performance of their assigned i

radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas. This footnote applies only to high radiation areas of i

1000 arem/hr or less.

.t 6.5 244a REV

E.

Offsite Dose Calculation Marual (ODCM) i The ODCM shall be approved by the Commission prior to initial implementation. Changes to the ODCM shall l

catisfy the following requirements:

1.

Shall by submitted to the Commission with the Semi-Annual Radioactive Effluent release report for

'[

the period in which the change (s) were made effective. This submittal shall contain:

sufficiently detailed information to totally support the rationale for the change without l

a.

benefit of additionally or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with a revision date, together with appropriate analyses or evaluations justifying the change (s).

i b.

a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and j

c.

documentation of the fact that the change has been reviewed and found acceptable by the Operations Committee.

l 2.

Shall become effective upon review and acceptance by the Operations Committee.

f F.

Security l

t Procedures shall be developed to implement the requirements of the Security Plan and the Security

(

Contingency. Plan. These implementing procedures, with the exception of those non-safety related procedures governing work activities exclusively applicable to or performed by security personnel, shall be reviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager. Security procedures not reviewed by the Operations Committee shall be reviewed and approved by the Superintendent, Security.

l G.

Temoorary Channes to Procedures Temporary changes to those procedures which are required to be reviewed by the Operations Committee described in A, B, C, D, E and F above, which do not change the intent of the original procedures may be made with the concurrence of two members of the unit management staff, at least one of whom holds a Senior Operator License. Such changes should be documented, reviewed by the Operations Committee and spproved by a member of plant management designated by the Plant Manager within one month. Temporary j

changes to health physics and security procedures not reviewed by the Operations Committee shall be j

reviewed by the General Superintendent, Radiation Services for health physics procedures and the Superintendent, Security for security procedures.

i 6.5 246b l

i

i B.

Records Retained for Plant Life (continued)

I i

11.

Records of the service lives of all safety-related snubbers, including the date at which the service life commences and associated installation and maintenance records.

t t

t t

i i

l l

6.6 247a REV t

I v

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

w

- -w

r 2.

Environmental Soecial Reports I

When radioactivity levels in samples exceed limits specified in Table 4.16.3 an Environmental Special Report shall be submitted within 30 days from the end of the affected calendar quarter.

For certain cases involving long analysis time, determination of quarterly averages may extend beyond the 30 days period.

In these cases the potential for exceeding the quarterly limits will be reported within the 30 day period to be followed by the Environmental Special Report as soon as practicable.

[

r 3.

Other Environmental Reports (non-radiological.~non-aauntic) n.

Environmental events that indicate or could result in a significant environmental impact causally related to plant operation.

The following are examples: excessive bird impac-tion; onsite plant or animal disease outbreaks; unusual mortality of any species protected l

by Endangered Species Act of 1973; increase in nuisance organisms or conditions; or excessive environmental impact caused by herbicide application to transmission corridors 7

associated with the plant. This report shall be submitted within 30 days of the event and shall (a) describe, analyze, and evaluate the event, including extent and magnitude l

of the impact and plant operating characteristics, (b) describe the probable cause of the i

event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses.

~

b.

Proposed changes, tests or experiments which may result in a significant increase in any adverse environmental impact which was not previously reviewed or evaluated in the Final Environmental Statement or supplements thereto. This report shall include an evaluation of the environmental impact of the proposed activity and shall be submitted 30 days prior i

to implementing the proposed change, test or experiment.

D.

Special Reports l

Unless otherwise indicated, special reports required by the Technical Specification shall be submitted within the time period specified for each report.

4 6.7 252 REV 1

-