ML19351F293

Proposed Revision to 801015 Tech Specs Sections 3.5 & 6 Re Valve Position Indication,Shift Technical Advisor,Sys Integrity & Improved Iodine Measurements Program,Per TMI-2 Lessons Learned Task Force Requirements

ML19351F293
Person / Time
Site:	Cooper
Issue date:	01/06/1981
From:	NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML19351F291	List: ML19351F293
References
RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8101120167
Download: ML19351F293 (6)

Text

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O TABLE OF CONTENTS (Cont'd.)

Page No.

SURVEILLANCE LIMITING CONDITIONS FOR OPERATION R"QUIREMENTS 3.14 Fire Detection System 4.14 216b 3.15 Fire Suppression Water System 4.15 216b 3.16 Spray and/or Sprinkler System (Fire Protection) 4.16 216e 3.17 Carbon Dioxide System 4.17 216f 3.18 Fire Hose Stations 4.18 216g 3.19 Fire Barrier Penetration Fire Seals 4.19 216h 3.20 Yard Fire Hydrant and Hydrant Hose House 4.20 2161 l

MAJOR DESIGN FEATURES 5.1 Site Features 217 5.2 Reactor 217 5.3 Reactor Vessel 217 5.4 Containment 217 5.5 Fuel Storage 218 5.6 Seismic Design 218 5.17 Barge Traffic 218 ADMINISTRATIVE CONTROLS 6.1 Organization 219 6.2 Review and Audit 220 6.2.1.A Station Operations Review Committee (SORd) 220 A.1 Membership 220 A.2 Meeting Frequency 220 A.3 Quorum 220 A.4 Responsibilities 220 A.5 Authority 221 l

A.6 Records 221 6.2.1.B NPPD Safety Review and Audit Board (SRAB) 222 l

B.1 Membership 223 B.2 Meeting Frequency 223 B.3 Quorum 223 B.4 Responsibilities 223 B.5 Authority 224 B.6 Records 225 B.7 Procedures 225 I

B.8 Fire Protection Inspection 225

$(01120 in

TABLE OF CONTENTS (Cont'd.)

Page No.

SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 6.3 Station Operating Procedures 226 6.3.1 (Introduction) 226 6.3.2 (Integrated and System Procedures) 226 6.3.3 (Maintenance and Test Procedures) 226 6.3.4 (Radiation Control Procedures) 226 6.3.5 (High Radiation Areas) 226a 6.3.6 (Implementation Review of Procedures) 226a 6.3.7 (Temporary Changes to Procedures) 226a 6.3.8 (Drills) 226a 6.4 Actions to be Taken in the Event of Occurrences Specified In Section 6.7.2.A 227 6.5 Actions to be Taken if a Safety Limit is Exceeded 227 6.6 Station Operating Recrods 228 6.6.1 (5 year retention) 228 6.6.2 (life retention) 228 6.6.3 (2 year retention) 229 6.7 Station Reporting Requirements 230 6.7.1 Routine Reports 230

.A (Introduction) 230

.B Startup Report 230

.C Annual Reports 230

.D Monthly Operating Report 231 6.7.2 Reportable Occurrences 231

.A Prompt Notification with Written Followup 232

.B Thirty Day Written Reports 234 6.7.3 Unique Reporting Requirements 235 6.8 Environmental Qualification 235a l

6.9 Systems Integrity Monitoring Program 235a l

6,10 Iodine Monitoring Program 235a i

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LIMITIJC CONDITIO:IS RM OPERATIO;I SUhVEILLA:!CE REQUIRCIE:ITS 3.6.D Safety and Relief Valvas 4.6.D Safety and Relief Valves 1.

During reactor power operating condi-1.

Approximately half of the safety valves tions and prior to reactor startup and relief valves shall be checked or from a Cold Condition, or whenever replaced with bench checked valves reactor coolant pressure is greater once per operating cycle. All valves than atmospheric and temperature will be tested every two cycles, greater than 212 F, all three safety valves and the safety modes of all The set point of the safety valves relief valves shall be operable, ex-shall be as specified in Specification cept as specified in 3.6.D.2.

2.2.

2.

2.

At least one of the reliet valves shall be disassembled and inspected each re-a.

From and after the date that the fueling outage, safety valve function of one relief valve is made or found to be inopera-3.

The integrity of the relief safety valve ble, continued reactor operation is bellows on any three stage valve permissible only during the succeeding shall be continuously monitored.

thirty days unless such valve function is sooner made operable.

4.

The operability of the bellows monitoring system shall be demonstrated once every b.

From and af ter the date that the safety three months when three stage valves valve function of two relief valves is are ins talled.

made or found to be inoperable, con-tinued reactor operation is permissible 5.

Once per operating cycle, with the.

only during the succeeding seven days reactor pressure > 100 psig, each relief unless such valve function is sooner valve shall be' manually opened until made ope rable.

the main turbine bypass valves have closed to compensate for relief valve 3.

If Specification 3.6.D.1 is not met, opening.

an orderly shutdown shall be initiated 6.

and the reactor coolant pressure shall a.

Operability of the relief valve position be reduced to a cool shutdown condi-indicating pressure switches and the tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, safety valve position indicating thermocouples shall be demonstrated 4.

From and af ter the date that position once per operating cycle.

indication on any one relief valve is made or found to be inoperable, contin-b.

An Instrument Check of the safety and ued reactor operation is permissible relief valve position indicating devices only durind the succeeding thirty days shall be performed conthly, unless such valve position indication is sooner uade operable.

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3.6.D & 4.6.D BASES (cont'd.)

The relief and safety valves are bench tested every second operating cycle to ensure that their set points are within the j; 1 percent tolerance. Additionally, once per operating cycle, each relief valve is tested manually with reactor pressure above 100 psig to demonstrate its ability to pass steam.

The requirements established above apply when the nuclear system can be pres-surized above ambient conditions.

These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed.

Ilowever, these transients are much less severe, in terms of pressure, than starting at rated conditions.

The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.

The position indicating pressure switches for the relief valves and the thermo-couples for the safety valves serve as a diagnostic aid to the operator in the event of a safety / relief valve failure.

If position indication is lost, alternate means are available to the operator to determine if a safety valve is leaking.

E.

Jet Pumps Failure of a jet pump nozzle assembly hold down mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design bases double-ended line break.

Therefore, if a failure occurs, repairs must be made.

The detection technique is as follows. With the two recirculation pumps balanced in speed to within j; 5%, the flou rates in both recirculation loops will be verified by Control Room monitoring instrumenta. If the two flow rate values do not dif fer by more than 10%, riser and nozzle assembly intedrity nas been verified.

If they do dif fer by 10% or more, the core flow rate measured by the jet pump diffuser differential pressure system cust be checked against the core flow rate derived f rom the measured values of loop flow to core flow correlation.

If the dif ference between measured and derived core flow rate is 10% or more (with the derived value higher) dif fuser measure-ments will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the plant shut down for repairs.

If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the af fected drive pump will "run out" to a substantially higher flow rate (approximately 115% to 120% for a single nozzle failure).

If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.

In addition, the af fected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive dif ferential pressure but the net ef fect would be a slight decrease (3% to 6%) in the total core flou measured.

This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.

Finally, the af fected jet pump dif fuser dif ferential pressure signal would be reduced because the backflow would be less than the normal forward flow.

A nozzle-riser system failure could also generate the coincident failure of a

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G.

A Fire Brigade of at least 5 members shall be maintained at all times. This excludes the 3 members of the minimum shift crew necessary for safe shutdowns, and other personnel required for other essential functions during a fire emergency.

Three fire Brigade members shall be from the Operations Department and 2 support members may be frou other departments inclusive of Security personnel.

Fire Brigade composition may be less than the mininum 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.

H.

In order to perf orm the function of accident assessment an engineer f rom the normal plant engineering staf f shall be assigned to each shif t during reactor operation.

If the lack of qualified engineers necessitates, an additional senior reactor operator assigned to each shif t may substitute in the performance of the accident assessment function. This requirement is ef fective until January 1, 1981.

6.1.4 The minimum qualifications, training, replacement training, and retraining of plant personnel at the time of fuel loading or appointment to the active position shall meet the requirements as described in the American National Standards Institute N-18.1-1971,

" Selection and Training of Personnel for Nuclear Power Plants".

The Assistant to Station Superintendent qualifications shall comply with Section 4.2 of ANSI-N18.1-1971.

The Chemistry and llealth Physics Supervisor shall meet or exceed the qualifications of Regulatory Guide 1.8, Sept. 1975; personnel qualification equivalency as stated in the Regulatory Guide may be proposed in selected cases. The minimum frequency of the retraining program shall be every two years. The training program shall be under the direction of a designated member of the plant staff.

A.

A training program for the fire brigade will be maintained under the direction of the plant training coordinator and shall meet or exceed the requirements' of Section 27 of the NFPA Code 1976, except for Fire Brigade training sessions which shall be held at least quarterly.

The training program requirements will be provided by a quali-fied fire protection engineer, thru the Risk 11anager.

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6.8 Environesntal Qualification A.

By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of:

Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", December 1979.

Copies of these documents are attached to Order for Modification of License DPR-46 dated October 24, 1980.

B.

By no later than December 1,1980, complete and audit 1ble records must be ceailable and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588.

Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.

6.9 Systems Integrity Monitoring Program A program shall be established to reduce leakage f rom systems outside the primary containment that would or could contain highly radioactive fluids during a serious accident to as low as practical levels.

This program shall include provisions establishing preventive maintenance and periodic visual inspection requirements, and leak testing requirements for each system at a frequency not to exceed refueling cycle intervals.

[p' 6.10 Iodine Monitoring Program A program shall be established to ensure the capability to accurately determine the airborne lodine concentration in vital areas under accident I

conditions.

This program shall include training of personnel, procedures for monitoring and provisions for maintenance of sampling and analysis equipment.

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