ML20079J682

Application to Amend License DPR-16,consisting of Tech Spec Change Limiting Oxygen Content to 4%.Class III Exemption Fee Encl

ML20079J682
Person / Time
Site:	Oyster Creek
Issue date:	12/15/1982
From:	Fiedler P GENERAL PUBLIC UTILITIES CORP.
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8212280313
Download: ML20079J682 (7)

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OPU Nuclear Ngg gf 100 Pars!nterpace ippany, New Parkway Jersey 07054 201 263-6500 TELEX 136-482 Writer's Direct Dial Number:

December 15, 1982 Director Nuclear Reactor Regulation Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Oyster Creek Nacicar Generati2g Station Docket No. 50-219 Final Rule on Interim Requirements Related to flydrogen Control - Exemption Request dated August 2, 1982 By letter dated August 2, 1982, the Oyster Creek Nuclear Generating Station requested an exemption from the Final Interim Requirements Related to flydrogen Cc.ntrol which was published in the Federal Register on Deember 2, 1981, pp. 53484 - 58486. At this time, it was stated that a site speciHc evaluation would be performed which will demonstrate that the exemptions and deferral previously requested will not endanger life or property or the common defense and security and are otherwise in the public interest.

This letter transmits the site specific evaluation for Oyster Creek.

As mentioned in the evaluation, a change to the Oyster Creek Technical Specifications is currently being analy:cd which will limit the oxygen content to 4%. It is anticipated that this change request will be completed and submitted upon approval of the requested exemption.

It has been previously determined that this exemption request is Class III in accordance with 10CFR 170.22.

Enclosed is a check for $4,000.

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80 8212280313 821215 p) l PDR ADOCK 05000219 P PDR b MI  %

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GPU Nuclear is a part of the General Public Utilities System

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I Hydrogen Control - Exemption Request Page 2 If you have any questions on this, please call Mr. James Knubel j at (201) 299-2264.

i-Very truly yours, f

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• w-  :?jsu

. 3. fir =ler Vice President and

Director - Oyster Creek

, b1f Enclosures cc: Ronald C. Haynes, Administrator

Region I i U.S. Nuclear Regulatory Commission

631 Park Avenue King of Prussia, PA 19406 i

NRC Resident Inspector Oyster Creek Nuclear Generating Station i

i Forked River, NJ 08731 i

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EVALUATION OF OXYGEN GENERATION DURING A HYPOTHETICAL ACCIDENT AT OYSTER CREEK ,

An evaluation of oxy 0en during a. hypothetical event at Oyster Creek was conducted.. Specific areas reviewed were:

1) Applicability of the Owner's Group Analysis (NEDO 22155) to Oyster Creek.

I 2') Reactor Vessel vent capability versus the NRC requirement (10CFR50.44C3111).

3) Venting Capability of the Isolation Condenser System.

The conclusions of the evaluation are:

A) .i. Oyster Creek is bounded in the oxygen generation rate by the ocse case analysis of the limiting plant and limiting event (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> boiling time).

11. Oyster Creek cannot take credit for.those sensitivity study cases with nitrogen dilution to 1 30 psig containnent oressure or beyond, because Oystar Creek's suppression pool is designed to 35 psig.

B) Oyster Creek has sufficient venting capability via ERV's for noncondensible gases accumulated in the reactor vessel head. It is highly reliable, remotely controllable and has sufficient redundancy to serve its

! p u rp os e. Therefore, the requirement for installing additional high point vents should be exempted.

C) The isolation condenser system at Oyster Creek is equipped with venting lines for noncondensible gases during normal power operation. Once the isolation condenser system is in operation, the vent lines will be automatically shut and further noncondensible gases, if i any, will be swept through the system by the circulating-fluid and will not jeopardize the function of the system. Therefore, a deferral for investigating the modification of its remote operability should not endanger the unit's overall safety.

Limiting Combustible Gas Determination For the determination of radiolytic oxygen generation following a LOCA, the limiting parameter is the ratio of core thermal power to drywell volume. The plant with the highest i

ratio would result in the highest oxygen concentration in the contali .aent. For Oyster Creek the licensed thermal power is 1930 MW'and the drywell volume is 180,000 cubic feet. Using 102% o MWt/ftjthelicensed'powertheratioturnsout-to'he0.0109

, which is'well below the corresponding ratio for the limiting plant (Brown's Ferry / Peach Bottom) that the O ner's Group chose for the base case analysis (0.0301) MWt/ft ).

Therefore, the' oxygen generation ratio for Oyster Creek is bounded by the generic analysis for Mark I containments.

Limitina Event Determination For the BWR, the limiting factor that would result in  ;

the highest oxygen generation is the time duration and extent of boiling for the reactor water in the core. In the Owner's i GrcJp report, extensive evidence has been shown that the oxygen generation in subcooled water is negligible. Consequently, the t limiting avent was selected such that the time duration and  ;

extent of noiling is maximized. .

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I The selected. event was an isolation transient with only low pressure core cooling available for a generic BWR/4 plant.

A boiling time period of up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> was assumed, 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> of which was taken before the operator took action to depressurize the vessel. Depressurization of the vessel to the piessure at which the Shutdown Cuoling System can be initiated was  ;

conservatively assumed to take 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Then an additional 2 >

hours was assumed for preparing the Shutdown Cooling System to ,

cool the reactor water. Once it is started, the reactor water  ;

will be cooled rapidly (within a few minutes) to subcooled.  ;

r For Oyster Creek, the same type of containment (Mark I) l as the generic analysis is required. Although the Oyster Creek  !

reactor vessel is a different type (BWR/2) from that of the BWR/4, its Low Pressure Core Spiay System basically has the same capability and is at least as reliable as was assumed in the generic analysis. Therefore, the limiting boiling time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> should be applicable to Oyster Creek. In addition, the ECCS of Oyster Creek is composed of another independent Isolation Condenser System which is absent in the generic case. The Isolation Condenser System at Oyster Creek is expected to function for emergency core cooling for any over-pressure / loss of inventory event, even without AC power.

By' including this system, the limiting boiling time would be greatly reduced. Nevertheless, the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> boiling time was accepted as conservative and bounding for Oyster Creek.

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I Containment Inerting System The containment inerting system at Oyster Creek maintains an atmosphere of nitrogen gas in the drywell -and suppression chamber during power operation. Oxygen is maintained at a concentration not to exceed 5 percent by.

volume, to prevent hydrogen-oxygen combustion following a postulated loss-of-coolant accident in which hydrogen is produced by metal-water reaction.

If a loss-of-coolant accident occurs, the ECCS safeguard features will prevent the generation of hydrogen in quantitles large encugh to ignite in the containment. Thus, the reduction of oxygen in the primary containment is not a technical requirement. However, the containment inerting system is provided to obviate the possibility of an energy release within the primsry containment from a hydrogen-oxygen reaction under conditions more severe than those currently analyzed.

Although the containment inerting system of Oyster Creek can be used to dilute the containment following a LOCA in order to preclude a hydrogen combustion event, such dilution will pressurize the containment and the containment's integrity would be jeopardized. NRC's current regulation-(10CFR50.44) specifies that containment repressurization be limited to 50 percent of the containment design pressure. For Gyster Creek the suppression chamber was designed to 35 psig and the

, Pressure Control Valves (PCV) close at 2 osig; thus cases 1, 3, 4, 5 and 6 with nitrogen repressurization, in the generic analysis sensitivity study, do not apply to Oyster Creek. For the reason cited earlier, repressurization by nitrogen injection at Oyster Creek is considered not desirable.

The generic base case analysis and others (cases 2 and

7) with initial oxygen concentration of 4% and no nitrogen repressurization show acceptable results (below the Regulatory Guide 1.7 limit of 5%) up to an extended time period (1000 days after accident). Althougn the Oyster Creek Technical Specification set the oxygen concentration to be less than 5%,

the plant routinely measures oxygen concentration to be below 3%. During power operation the oxygen analyzer is calibrated every week. In addition, there is an automatic annunciator for high oxygen concentration. Pure nitrogen is used for pneumatic operated instrument lines, and the containment atmosphere leakage will also be replaced by pure nitrogen makeup injection.

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In order to take credit for the Owner's Group bounding case study, the Oyster Creek Technical Specification on oxygen content should be changed to 4%. Subsequently, all plant operating procedures and alarm setpoint settings should be changed accordingly. This will warrant that oxygen co'ncentration will not reach the flammability limit at Oyster Cr: k during the hypothetical event postulated.

10CFR 50.44.C 3111 Reactor Vessel High Point Vent There are five (5) electromatic relief valves (ERV) and sixteen (16) safety valves in the Oyster Creek reactor vessel steam lines that inherently have the capability for venting noncondensible gases. The ERV's are safetygrade and can be operated from the control room. A stuck open relief valve or safety valve has been analyzed in the unit's FSAR. Therefore, installation of more vent valves for high point venting is considered not necessary and should thus be exenpted.

Venting Caoability of the Isolation Condensers The operation of isolation condensers is expected in the event of a high reactor pressure or low reactor water level condition. When these conditions occur during non-LOCA transients, the core remains covered and hydrogen generation does not result. During a LOCA, the automatic depressurization system (ADS) and the low pressure core spray will ensure sufficient core cooling is provided to meet the limits of 10 CFR 50.46.

At the high points of each isolation condenser loop steam piping there is a vent line for noncondensibles. The vents are, however, designed and used for venting the noncondensibles continuously to the main turbine steam header when the plant is operating and the isolation condenser system is on standby. This is done to remove noncondensible gases from the system which would otherwise collect at these high points and impair the initial actuation and operatioq capacity o' the isolation condenser system. These vents are automatically shut of f when the emergency condenser system goes into operation. The existing high point vent lines in the isolation condenser system partially meet the intent of 10CFR 50.44 C 111. In operation, the system fluid flow will sweep noncondensibles through the system and therefore, no accumulation of noncondensible gases would cause the loss of function of the isolation condenser system.

i In the event of a design basis large break LOCA resulting in hydrogen generation in the core, the vent valves cannot be remotely operated from the control room and hence o

will n't be able to vent the post-accident noncondensible gases out of the reactor vessel. GPU Nuclear is currently engaged in investigating the feasibility of modifying the. venting system to fully comply with the requirements. A preliminary assessment is expected by the second outage after August 1982, (i.e. Cycle 11).

In conclusion, the ERV's still provide the necessary capability of venting gases from the reactor. vessel and they are fully operable from the control room. Also, the isolation condensers are not required to mitigate the effects of LOCA events since the ADS and core spray systems serve this function adequately. Thus, this deferral will not reduce the plant's overall margin of safety.

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