ML19323B867

Proposed Tech Spec Change to Section 4.13 Re Containment Vent & Purge Sys & 3.8 for Refueling,Fuel Handling & Storage.Safety Evaluation Encl

ML19323B867
Person / Time
Site:	Indian Point
Issue date:	05/05/1980
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML19323B856	List: ML19323B855 ML19323B862 ML19323B867
References
NUDOCS 8005140364
Download: ML19323B867 (9)

Text

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60051403(pL i

e ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGE RELATED TO I

I CONTAINMENT VENT AND PURGE SYSTEM AND REFUELING, FUEL HANDLING AND STORAGE

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i POWER A"'"HORITY OF THE STATE OF NEW YORK  !

! INDIA. ?OINT 3 NUCLEAR POWER PLANT l DOCKET NO. 50-286 i MAY 5, 1980 f

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S7ction Title P. .g o 4.'13 Containm:nt Vcnt and Purgo System 4.13-1l

5. Dacign Features 5.1-1 5.1- Site 5.1-1

. 5.2 Containment 5.2-1 Reactor Containment 5.2-1 Penetrations 5.2-1 Containment Systems 5.2-2 5.3 Reactor 5.3-1 Reactor Core 5.3-1 Reactor Coolant System 5.3-2 5.4 Fuel Storage 5.4-1

6. Administrative Controls 6-1 6.1 Responsibility 6-1 6.2 Organization 6-1 Facility Management and Technical 6-1 Support Facility Staff 6-1 6.3 Facility Staff Qualifications 6-5 6.4 Training 6-5 6.5 Review and Audit 6-5 Plant Operating Review Committee 6-5

1) Function 6-5

2) Composition 6-6

3) Alternates 6-6

4) Meeting Frequency 6-6

5) Quorum 6-6

6) Responsibilities 6-6

7) Authority 6-7

8) Records 6-7 Safety Review Committee 6-8

1) Function 6-8

2) Composition 6-9

3) Alternates 6-9

4) Consultants 6-9

5) Meeting Frequency 6-9

6) Quorum 6-9

7) Review 6-10

8) Audits 6-11

9) Authority 6-11

10) Records 6-12 6.6 Reportable Occurrence Action 6-12 6.7 Safety Limit Violation 6-12 6.8 Procedures 6-13 6.9 Reporting Requirements 6-13 Routine and Reportable Occurrence 6-13 Reports Special Reports 6-18 6.10 Record Retention 6-18 6.11- Radiation and Respiratory Protection 6-19 Program 6.12 High Radiation Area 6-20 Amendment No. 14,l/ iii i

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3.8 R3funling, Funl Handling cnd Storaga Appliccbility

, Applies to operating limitations during refueling, fuel handling, storage operations, and when heavy loads are moved over the reactor when the head is removed.

Objective To ensure that no incident could occur during refueling, fuel handling, and storage operations that would a3versely affect public health and safety.

Specification A. During handling operations, reactor vessel head removal or installation, or movement of heavy loads over the reactor vessel with the head removed, the following conditions shall be satisfied:

1. The equipment door and at least one door in each personnel air lock shall be properly closed. In addition, at least one isolation valve shall be operable or locked closed in each line penetrating the containment and which provides a direct path from containment atmosphere to the outside.

2. Radiation levels in the containment and spent fuel storage areas shall be monitored continuously.

3. The core subcritical neutron flux shall be continuously monitored by t' 3 two source range neutron monitors, each with continuous visual indication in the control room and one with audible indication in the containment available whenever core geometry is being changed. When core geometry is not being changed, at least one source range neutron flux monitor shall be in service.

4. At least one residual heat removal pump and heat exhanger shall be operating except during those core alterations in which the residual heat removal flow interferes with component positioning.

5. During reactor vessel head removal and while loading and unloading fuel in the reactor, Tavg shall be <140% and the minimum boron concentration sufficient to maintain the reactor subcritical by at least 10% Ak/k. The required boron concentration shall be verified by chemical analysis daily.

6. Direct communication between the control room and the refueling cavity manipulator crane shall be available when-ever changes in core geometry are taking place.

Amendment No. g 3.8-1 i

. 7. The containment vent and purge system, including the radiation monitors which initiate isolation, shall be tested and verified to be operable within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to refueling operations.

8. No movement of irradiatLd fuel in the reactor shall be made until the reactor has been subcritical for at least 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />. In addition, movement of fuel in the reactor before the reactor has been suberitical for equal to or greater than 365 hours0.00422 days <br />0.101 hours <br />6.035053e-4 weeks <br />1.388825e-4 months <br /> will necessitate operation of the Containment Building Vent and Purge System through the HEPA filters and charcoal adsorbers. For this case operability of the Containment Building Vent and Purge System shall be established in accordance with Section 4.13 of the Technical Specifications. In the event that more than one region of fuel (72 assemblies) is to be discharged from the reactor, those assemblies in excess of one region shall not be dis-charged before an interval of 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> has elapsed after shutdown.

9. Whenever movement of irradiated fuel is being made, the minimum water lever in the area of movement shall be maintained 23 feet over the top of irradiated fuel assemblies seated within the reactor pressure vessel.

10. Hoists or cranes utilized in handling irradiated fuel shall be dead-load tested before movement begins. The load assumed by the hoists or cranes for this test must be equal to or greater than maximum load to be assumed by the hoists or cranes during the refueling operation. A thorough visual inspection of the hoists or cranes shall be made after the dead-load test and prior to fuel handling. A test of inter-locks shall also be performed.

11. The fuel storage building emergency ventilation system shall be operable whenever irradiated fuel is being handled within the fuel storage building. The emergency ventilation system may be inoperable when irradiated fuel is in the fuel storage building, provided irradiated fuel is not being handled and neither the spent fuel cask nor the cask crane are moved over the spent fuel pit during the period of inoperability.

Amendment No. g 3.8-2

6. The fuel storego building emergancy ventilation cystem chall be op3rable whensvar irradiated fuel is being handled within the fuel storage building. The emergency ventilation system

. may be inoperable when irradiated fuel is in the fuel storage building, provided irradiated fuel is not being handled and neither the spent fuel cask nor the cask crane are moved over the spent fuel pit during the period of inoperability.

Basis The equipment and general procedures to be utilized during refueling, fuel handling, and storage are discussed in the FSAR. Detailed instructions, the above specified precautions, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling, fuel handling, reactor maintenance or storage o e ations that would result in a hazari to public health and safety.

Whenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumentation.

Continuous monitoring of radiatian levels and neutron flux provides immediate indication of an unsafa condition. The residual heat removal pump is used to maintain a uniform boron concentration.

The shutdown margin indicated will keep the core subcritical, even if all control rods were withdrawn from the core. During refueling, the reactor refueling cavity is filled with approximately 342,000 gallons of water from the refueling water storage tank with a boron concentration of 200.0 ppm. A shutdown margin of 10% Ak/k in the cold condition with all rods inserted will also maintain the core sub-critical even if no control rods were inserted into the reactor.(2)

Periodic checks of refueling water boron concentration and residual heat removal pump operation insure the proper shutdown margin.

The requirement for direct communications allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

In addition to the above safeguards, interlocks are utilized during refueling to ensure safe handling. An excess weight interlock is provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. The spent fuel transfer mechanism can accomodate onif one fuel assembly at a time.

The 120-hour decay time following the subcritical condition and l the 23 feet of water above the top of the irradiated fuel assemblies are consistent with the assumptions used in the dose calculaion for the fuel-handling accident.

The waiting time of 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> required following plant shutdown before unloading more than one region of fuel from the reactor assures that the maximum pool water temperature will be within design objectives as stated in the FSAR.

Amendment No. g 3.8-4

The requirement for the fuel storage building emergency ventilation system to be operable is established in accordance with standard testing requirements to assure that the system will function to reduce the offsite dose to within acceptable limits in the event of a fuel-handling accident. The system is actuated upon receipt of a signal from the area high activity alarm or by a manually-operated switch. The system is tested prior to fuel handling and is in a standby basis.

When fuel in the reactor is moved before the reactor has been suberitical for at least 365 hours0.00422 days <br />0.101 hours <br />6.035053e-4 weeks <br />1.388825e-4 months <br />, the limitations on the contain-ment vent and purge system ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorbers prior to discharge to the atmosphere.

The minimum spent fuel pit boron concentration and the 90-day restriction of the movement of the spent fuel cask to allow the irradiated fuel to decay were specified in order to minimize the consequences of an unlikely sideways cask drop.

When the spent fuel cask is being placed in or removed from its position in the spent fuel pit, mechanical stops incorporated'in the bridge rails make it impossible for the bridge of the crane to travel further north than a point directly over the spot reserved for the cask in the pit. Thus, it will be possible to handle the spent fuel cask with the 40-ton hook and to move new fuel to the new fuel elevator with a 5-ton hook, but it will be impossible to carry any objective over the spent fuel storage area with either the '

40 or 5-ton hook of the fuel storage building carne.

l Amendment No. ,35' 3.8-5 1

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4.13 containment Vent and Purge System Applicability This specification applies to the surveillance requirements of the containment vent and purge system when reactor fuel is antici-pated to be moved before the reactor has been subcritical for at least 365 hours0.00422 days <br />0.101 hours <br />6.035053e-4 weeks <br />1.388825e-4 months <br />.

Objective To verify the operability of the containment vent and purge system when reactor fuel is anticipated to be moved before the reactor has been subcritical for at least 365 hours0.00422 days <br />0.101 hours <br />6.035053e-4 weeks <br />1.388825e-4 months <br />.

Specification If fuel movement is to take place before the reactor has been suberitical for at least 365 hours0.00422 days <br />0.101 hours <br />6.035053e-4 weeks <br />1.388825e-4 months <br />, the containment vent and purge system shall be demonstrated operable as follows:

a. Within 18 months prior to fuel mcVement:

1. Verify that the charcoal adsorbers remove >99% of halo genated hydrocarbon refrigerant test gas when they are tested in-place while operating the ventilation system at the operatin g flow rate i 10%.

2. Verifying that the HEPA filter banks renove 199% of I the DOP when they are tested in-place while operatin g 1 the ventilction system at the operatin g flow rate I i 10%.

b. Within 18 nonths prior to fuel noverent and after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation, subject a representative sanple of carbon from the charcoal adsorbers to a laboratory analysis and verify within 31 days a re noval efficiency of 290% for radioactive nethyl iodine at an operating air I flow velocity 120% per test 5.b in Table 2 of Re gulatory Guide 1.52, March 1978.

Basis The operability of this system and the resulting iodine renoval capacity are consistent with accident analyscs. The reprsentative carbon sanple will be two inches in dianeter with a len gth _ equal to the thickness of the bed.

Anendnent No. 4.13-1 l

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ATTACHMENT II SAFETY EVALUATION RELATED TO CONTAINMENT VENT AND PURGE SYSTEM AND REFUELING, FUEL HANDLING AND STORAGE POWER AUTHORITY OF THE STATE OF NEW YORK INDIAN POINT 3 NUCLEAR POWER PLANT ,

DOCKET NO. 50-286 MAY 5, 1980

"'Shetion I - De cription of Modification The operating limitations for Refueling, Fuel Handling and Storage e are being changed to require that during refueling operations., the containment vent and purge system shall be exhausting through HEPA filters and charcoal adsorbers when irradiated fuel has decayed less than (365) hours. Surveillance requirements to yerify the operability of the containment vent and purge system are also being added.

The operating limitation for Refueling, Fuel Handling and Storage are also being changed to require additional precautions for the control of heavy loads over the reactor when the head is removed, during the other than refueling operations.

Section II - Purpose of Modification The purpose of these modifications is to (1) ensure that the potential consequences of a fuel handling accident inside containment are appropriately within the guidelines of 10 CFR Part 100, and (2) prevent lifting of heavy loads over the reactor when the head is removed for other than refueling operations so as not to damage the reactor core.

Section III - Impact of the Change These modifications will not alter the conclusions reached in the FSAR and SER accident analyses.

Section IV - Implementation of the Modification The modifications as proposed will not impact the ALARA c>r Fire Protection Program at IP3.

Section V - Conclusion The incorporation of these modifications: a) will not increase the probability nor the consequences of an accident or mal-function of equipment important to safety as previously evalu-ated in the Safety Analysis Report; b) will not increase the possibility for an accident or malfunction of a different type than any avaluated previously in the Safety Analysis Report; and c) will not reduce the margin of safety as defined in the basis for any Technical Specification, and d) does not constitute an unreviewed safety question.

Section VI - References (a) IP3 FSAR (b) IP3 SER k 7