ML20076B541

Deficiency Rept Re Containment Gas Analyzer Cell Catalysts Mfg by Comsip,Inc.Initially Reported on 830711.Upgraded Catalysts Installed.Also Reported Per Part 21 by Comsip. Proof & Review Tech Specs & FSAR Section 9.4.6. Encl

ML20076B541
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	08/12/1983
From:	James Smith LONG ISLAND LIGHTING CO.
To:	Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
REF-PT21-83 10CFR-050.55E, 10CFR-50.55E, SNRC-951, NUDOCS 8308190396
Download: ML20076B541 (10)

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> 1 LONG ISLAND LIGHTING COM PANY

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wwww 3 SHOREHAM NUCLEAR POWER STATION ew nve r.+ m u e P.O. BOX 618, NORTH COUNTRY ROAD e WADING RIVER, N.Y.11792 Dkect Dial Numkr August 12, 1983 SNRC-951 Dr. Thomas E. Murley Regional Administrator Office of Inspection & Enforcement - Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 Gas Analyzer Cell Catalysts Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322

Dear Dr. Murley:

On July ll, 1983, in accordance with 10CFR50.55(e), we reported verbally to Region I a potential deficiency involving contain-ment gas analyzer cell catalysts. This had been previously reported under 10CFR21 by COMSIP, Inc. to Mr. R. C. DeYoung, (Director I&E). Subsequent Shoreham specific engineering review indicates that the analyzer cell deficiency applies to Shoreham and is considered a reportable deficiency. Therefore, this letter will serve as our thirty-day written report pertaining to this deficiency.

Description of Deficiency The primary containment atmosphere may contain fission fragment iodine following certain accident conditions involving core damage. Calculations performed by Stone & Webster confirm the potential for a primary containment iodine concentration of 0.016 mm Hg partial pressure. This concentration adversely affects the cell catalyst such that the usable life is decreased to a duration in the order of 8-10 days. After this period, the analyzer would provide erroneous readings for oxygen and hydrogen

, concentrations (percent by volume) in both the primary containment ,

and the recombiner discharge. Loss of indication due to analyzer failure will reduce the ability of the control room to accurately monitor hydrogen /cxygen generating events.

8308190306 830812 PDR ADOCK 05000322 PDR i S FC-8935.1 1

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A August 12, 1983 SNRC-951 Page 2 Corrective Action Our corrective action involves installation of upgraded analyzer cell catalysts manufactured by COMSIP, Inc. This material has been procured by our E&DCR P-4443 ~and is. currently being in-stalled by E&DCR P-4443A. The new model analyzer-cell catalysts are currently undergoing testing by COMSIP and as of July 27, 1983, have passed the 127-day mark of a scheduled 180-day con-secutive test run; The testing is bcing done with a sample atmosphere of 0.02 mm Hg partial pressure _(iodine) . The replace-ment cell catalysts will, therefore, function properly under Shoreham's post-accident monitoring environment. Upon installa-tion of the new cell catalysts, no further modifications will be required. This modification will be completed prior to exceeding a power level of five (5) percent.

Very truly yours, J. L. Smith Manager, Special Projects Shoreham Nuclear Power Station BWC:jm ec: Mr. Richard DeYoung, Director NRC Office of Inspection and Enforcement Div. of Reactor Operations Inspection Washington, D.C. 20555 Mr. J. C. Higgins USNRC Resident Inspector NRC Site Office (TSC)

All Parties Listed in Attachment 1

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. l ATTACRMENT 1 Lawrence Brenner,'Esq. . Herbert H. Brown, Esq.

Administrative Judge Lawrence Coe Lanpher, Esq.

Atomic Safety and Licensing Karla J. Letsche, Esq.

Board Panel Kirkpatrick, Lockhart, Hill U.S. Nuclear Regulatory Christopher & Phillips Commission 8th Floor

. Washington, D.C. 20555 1900 M. Street, N.W.

Washington, D.C. 20036 Dr. Peter A. Morris Administrative Judge Mr. Marc W. Goldsmith

. Atomic Safety and Licensing , Energy Research Group Board Panel 4001 Totten Pond Road U.S. Nuclear Regulatory Waltham, Massachusetts 02154 Commission e Washington, D.C. 20555 MHB Technical Associates "

, 1723 Hamilton Avenue Dr. George A. Ferguson Suite K School of Engineering San Jose, Califcrnia 95125 Howard University 2300 Fifth Street Washington, D.C. 20059 Stephen B. Latham, Esq. .

Twomey, Latham & Shea 33 West Second Street Daniel F. Brown, Esq. P.O. Box 398 Attorney Riverhead, New York 11901 Atomic Safety and Licensing Board Panel ,

U.S. Nuclear Regulatory Ralph Shapiro, Esq. C Commission Cammer and Shapiro, P.C.

Washington, D.C. 20555 9 East 40th Street New York, New York 10016 Bernard M. Bordenick, Esq.

David A. Repka, Esq. Matthew J. Kelly, Esq.

U.S. Nuclear Regulatory State of New York Commission Department of Public Service Washington, D.C. 20555 Three Empire State Plaza Albany, New York 12223 Mr. James Dougherty l

3045 Porter Street Washington, D.C. 20008 l

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Attachment 2 o

4 ' SNPS-1 FSAR

,J E m 9.4.6 Drywell Air Cooling System , .

'9.4.6.1 Design Bases )

The drywell air cooling system circulates the atmosphere within the drywell to remove heat and maintain design temperatures. The drywell will be maintained at a maximum temperature of 146 F (avg. 127 F) during normal operation. The control rod drive. area design temperature is 150 F, while maximum allowed temperature in the area is 165 F. In the proximity of electrical equipment, maximum temperature is 130 F. The system is not a safety related system.

9.4.6.2 System Description 1

During normal operation air is circulated as shown in Fig. 9.4.6-1 within the drywell by two unit coolers each

, comprised of four cooling coils, four 10l000 cfm fans, intake and discharge dampers, and supply and return ductwork. Air is supplied by three of the four fans in each unit to the lower level of the drywell and returned to the unit from the top of the drywell. Water from the reactor building closed loop cooling water system is used as the cooling medium for the coils. No control of air flow or water flow is provided for this system.

However, should the requirements for cooling decrease, one or more of the operating fans can be shut down manually from the main control room. g NJ To ensure continuous operation during loss of offsite power and no accident signal present, the drywell unit cooler fans, dampers, and valves are connected to the emergency power supply.

The units are designed to operate during al'1 normal plant operations, j 9.4.6.3 Safety Evaluation The unit coolers are designed to meet the cooling requirements of the drywell with three fans in each unit running and one fan as a spafe. Upon failure of any one of the three running fans or -

associated dampers in each unit, an alarm will sound in the main control room and the spare fan will be started manually from the main control room. ,,

Upon indication of high pressure in the drywell or low reactor water level signals, the drywell unit coolers are automatically shut down, and - all primary containment isolation valves in the cooling water piping are closed automatically.

I The system is not safety related. However, all ductwork and .

! equipment are seismically supported to ensure they stay in place and do not damage safety related equipment in the area.

O 9.4-16

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. SNPS-1 FSAR A w 9.4.6.4 Tests and Inspections All components are tested and inspected as separate components and as integrated systems. After the system is completely installed, air flows are measured and adjusted to meet design flow rates. During plant normal shutdown, the system will be inspected and readjusted, if required, to meet design flow rates.

9.4.6.5 Instrumentation Application Drywell unit cooler controls including selector switches, monitors, and system alarms are located on panels in the main control room.

Alarms are provided in the main control room for the following conditions:

1. Control Rod Drive area high temperature

2. Upper drywell exhaust high temperature

3. Drywell head area exhaust high temperature

4. Reactor building closed loop cooling water return water high temperature

5. Drywell unit cooler high supply air temperature

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9.4-17 Revision 1 - March 1976

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• Attachment 3 PtEF & Et;_P!/ 2py

, CONTAINMENT SYSTEMS DRYWELL AVERAGE AIR TEMPERATURE ,

l LIMITING CONDITION FOR OPERATI'ON 3.6.1.7 Drywell average air temperature shall not exceed 145*F.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With the drywell average air temperature greater than 145*F, reduce the cverage air temperature to within the limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.7 The drywell average air temperature shall be the volume weighted

(- average of the temperatures at the following locations and shall be determined to be within the limit at least once per 24 hoers:

Elevation Azimuth

a. 68'-0" 13*, 320 -

b. 80'-0" 190*, in CRD area

c. 83'-0" 25*, 135*, 255*

d. 110'-0" '165*, 350'

e. 145'-0" 55*, 230*

f. 162'-6" ,

Reactor Vessel Centerline i

1 SH0kEHAM - UNIT 1 3/4 6-12 l l

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PLANT SYSTEMS g 'g g]

(, 3/4.7.9 AREA TEMPERATURE MONITORING

, LIMITING CONDITION FOR OPERATION 3.7.9 The temperatureof each area shown in Table 3.7.9-1 shall be maintained within the limits indicated.

PPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE.

ACTION:

With one or more areas exceeding the temperature limit (s) shown in Table 3.7.9-1:

a. For more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, in lieu of any report required by Specification 6.9.1, prepare and submit' a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days providing a record of the amount by which and the cumulative time the temperature in the affected area exceeded its limit and an analysis to demonstrate the continued OPERABILITY of the affected equipment.

b.

By more than 30*F, in addition to the Special Report required above,

' within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either restore the area to within its temperature limit or declare the equipment in the affected area inoperable.

SURVEILLANCE REQUIREMENTS 4.7.9 The temperature in each of the areas shown in Table 3.7.9-1 shall be determined to be within its limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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SHOREHAM - UNIT 1 3/4 7-36

.. PR00F & REVH COPY

,, TABLE 3.7.9-1 AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT (*F)

a. Control Room 90

b. Chiller Equipmen't Room (El 63') 104

c. Relay Room 104

d. Emergency Switchgear Rooms 104

e. Battery Rooms 104

f. bieselGeneratorRooms 120

g. Screenwell House 104

h. Reactor Building - Secondary Containment

1. General Areas 104

2. Refueling Area .

110 ,

i. Reactor Building - Primary Containment

1. General Areas 150 (f . 2.

3.

Area Beneath RPV Drywell Head Area 150*

185 -

A 165*F during Scram.

SHOREHAM - UNIT 1 3/4 7-37  ;

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CONTAINMENT SYSTEMS g !g { ,

i BASES 3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY .

~This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the unit. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 48 psig in the event of a.LOCA. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

3/4.6.1.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE The ifmi'tations on drywell and suppression chamber internal pressure -

ensure that the containment peak prassure of 46.0 psig does not exceed the design, pressure of 48 psig during LOCA conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 5.7 psid. The upper limit of Figure 3.6.1.6-1 will limit the total pressure to 46.0 psig which is less than the design pressure and is consistent with the safety analysis. The lower limit of Figure 3.6.1.6-1 ensures that the peak LOCA temperature will not exceed the safety analysis value and the containment external pressure requirements are satisfied.

3/4.6.1.7 DRYVELL AVERAGE AIR TEMPERATURE .

The limitation on drywell average air temperature ensures that the l containment peak air temperature does not exceed the design temperature of l 340*F during LOCA conditions and is consistent with the safety analysis. .

3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM The 18-inch drywell and suppression chamber purge supply and exhaust isolation valves are required to be sealed closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA.or steam line break accident. Maintaining these valves sealed closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the containment purge system. To provide assurance that the 18-inch valves cannot be inadvertently opened, they are sealed closed in accordance with Standard Review Plan 6.2.4, which includes me'chanical devices to seal or lock the valve closed or prevent power from being supplied to the valve operator.

The use of the drywell and suppression chamber purge lines.is restricted to the 4-inch and the 6-inch purge supply and exhaust isolation valves since, unlike the 18-inch valves, the 4-inch and the 6-inch valves will close during a LOCA or steam line' break accident and therefore the SITE BOUNDARY dose guidelines of 10 CFR Part 100 would not be exceeded in the event of an accident during PURGING operations. The design of the 4-inch and the 6-inch purge supply and exhaust

• isolation valves meets the requirements of Branch Technical Position CSB 6-4,

. " Containment Purging During Normal Plant Operations."

i SHOREMAM - UNIT 1 8 3/4 6-2 .

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r PLANT SY. STEMS BASES hk h {}ljg i

3/47[7 FIRE SUPPRESSION SYSTEMS i

The OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available ,to confine and extinguish fires occurring in any portion of the facility where safety related equipment is located. The fire suppression system consists of the water system, deluge, CO2 systems, Halon systems and fire hose stations. The collective capability of the fire suppression systems is adequate to minimize potential damage to safety related equipment and is a major element in the, facility fire protecticn program.

In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service.

When the inoperable fire fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is all' owed to provide an alternate means of fire fighting than if the inoperatile equipment is the primary means of fire suppression.

The surveillance requirements provide assurances that the minimum

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OPERABILITY requirements of the fire suppression systems are met. An allowance is made for ensuring a sufficient' volume of Halon in the Halon cylinders by verifying the weight and pressure of the tanks.

In the event the fire suppression water system becomes inoperable, imediate corrective measures must be taken since this system provides the '

major fire suppression capability of the plant. ~

3/4.7.8 FIRE RATED ASSEF"dLIES . .

The OPERABILITY fire damage of the fire barriers and barrier penetrations ensure that will be limited.

These design features minimize the possibility of a single fire involving more than one fire area prior to detection and extinguishment. The fire barriers, fire barrier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodically inspected to verify their OPERABILITY.

I 3/4.7.9 AREA TEMPERATURE MONITORING The araa temperature limitations ensure that safety related equipment will

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not be subjected to temperatures in excess of their environmental qualification temperatures.

Exposure to excessive temperatures may degrade equipment and can cause loss of its OPERABILITY.

3/4.7.10 MAIN TURBINE BYPASS SYSTEM The main turbine bypass system is required to be OPERABLE consistent with

.{ the assumptions of the feedwater controller failure analysis for FSAR Chapter 15.

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SHOREHAM - UNIT 1 8 3/4 7-4

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