ML20076A971
| ML20076A971 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 08/15/1983 |
| From: | James Smith LONG ISLAND LIGHTING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SNRC-952, NUDOCS 8308180399 | |
| Download: ML20076A971 (11) | |
Text
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I LONG ISI_AND LIGHTING COM PANY M,
U fgy SHOREHAM NUCLEAR POWER STATION
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--===~4 P.O. DOX 618, NORTH COUNTRY ROAD e WADING RIVER N.Y.11792 Direct Dial Number August 15, 1983 SNRC-952 Mr. liarold R. Denton, Director Office of Nuclear Reactor Regulation U.S.
Nuclear Regulatory Commission Washington, DC 20555 Equipment Environmental Qualification SER Outstanding Issue No. 9 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322
Dear Mr. Denton:
This letter provides additional information requested by the NRC staff based on their review of LILCO submittal SNRC-945, dated August 3,
1983.
Two topics are addressed; 1) the bases for our conclusion that 30 events in Section 15 of the Shoreham FSAR do not create a harsh environment, and 2) whether the loss of drywell coolers is an anticipated operational occurrence and, therefore, not covered by 10CFR50.49 or a design basis event which creates a harsh environment.
Of the 38 FSAR Chapter 15 events, 30 events were determined not to create a harsh environment.
The following provides the basis for this determination.
Events 15.1.1 - 15.1.26 and 15.1.29 are considered Anticipated Operational Occurrences based on the 10CFR50 Appendix A definition l
and the frequencies as defined in Chapter 15 of the FSAR.
There i
are no pipe ruptures and no fuel failures postulated for these I
- events, llowever, for some of the events radioactivity is dis-l charged to the suppression pool as a result of SRV actuation, as l
well as llPCI/RCIC operation.
These discharges were anticipated in the development of the plant normal environmental conditions and are accounted for in the 40 year normal dose for the suppression pool.
Therefore, it was determined that no harsh environment was l
created.
Event 15.1.27, Anticipau 3 Transients without Scrara, refers to NEDO 10349 which states, "since there were no calculated fuel failures 8308180399 830815 PDR ADOCK 05000322 l
E PDR rc.asas.:
SNRC-952 August 15, 1983 Page 2 nor any requirements for venting as a result of a transient with no scram or recirculation pump trip, no off-site radiological effects occur".
Again, no pipe ruptures were postulated.
Therefore, it was determined that no harsh environment was created.
Event 15.1.33, Control Rod Drop, does not create a harsh environ-ment based on no rupture postulated and on a Stone & Webster Engineering Corp. (SWEC) calculation which results in a gix month integrated gamma dose in the steam tunnel of 3.4 x 10 rads assuming no airborne source.
This calculation is based on the fuel damage and fission product release described in FSAR Section 15.1.33.5.2.2, with the release assumed to be distributed instanta-neously and uniformly in the steam volume including the main steam lines in the steam tunneg.
This dose is less than the 40 year normal dose of 4.9 x 10 rads (EQR Rev.
5, Fig. D-2) and is therefore considered a non-harsh environment.
Event 15.1.36, Fuel Handling Accident, does not create a harsh environment based on no pipe rupture postulated and on a SWEC calculation which results in a paximum six month integrated dose in t
the reactor building of 2 x 10 rads.
This calculation is based on the fuel damage and fission product release described in FSAR Section 15.1.36.5.2.2, with 100 percent of the released noble gas and 1 percent of the released iodine assumed airborne and 99 percent of the released iodine assumed in the fuel pool water.
This dose is not pignificantly more severe than the 40 year normal dose of 1.8 x 10 rads (EQR Rev.
5, Fig. D-2) and therefore is considered a non-harsh environment.
The Drywell Air Cooling System (DACS) is described in FSAR Section 9.4.6, a copy of which is attached for your convenience (Attachment
- 2).
The maximum normal temperatures given in the Environmental Qualification Report (i.e.,
the EQR, revised by LILCO letter SNRC-917, dated June 27, 1983) Appendix D and upon which equipment qualified lives are based, are consistent with the DACS design.
Operation with temperatures exceeding these values is restricted by Technical Specifications 3/4.6.1.7 and 3/4.7.9, a copy of which is l
attached for your convenience (Attachment 3).
(The next revision of the EQR will eliminate the discrepancy with Technical Specifica-tion Table 3.7.9-1 in the drywell head area temperature.
It has l
been determined that this change will affect neither the qualified i
status of equipment located in the area nor any justifications for l
interim operation.)
As described in the FSAR, the drywell coolers are supplied by l
back-up emergency power and loss of drywell cooling is not con-l sidered an anticipated operational occurrence.
- Also, since operation with high drywell temperature is limited by Technical
l SNRC-952 August 15, 1983 Page 3 Specifications, the loss of drywell cooling is not considered a design basis event.
In addition, the loss of drywell cooling is inherently encompassed by the Shoreham Environmental Qualification Program.
This is due to the fact that a LOCA automatically causes an isolation of the drywell cooling system.
The resulting LOCA environment is a combination of the mass and energy release from the break, fuel damage and the loss of the drywell cooling system.
In the event of a simple loss of drywell cooling with no other accident, the drywell temperature would gradually increase, but there would be no other effect on drywell atmosphere.
In conclusion, loss of drywell cooling has been adequately account-ed for in the plant design and plant operational restrictions and, therefore, this circumstance is not subject to the requirements of 10CFR50.49.
It is our understanding that this information satisfies all out-standing staff requests for information needed to resolve SER Outstanding Issue No.
9.
Should there be any questions, please contact this office.
Very truly yours, 6
ss L.
Smith Manager, Special Projects Shoreham Nuclear Power Station JFE/ law S3 Enclosure cc:
J.
liiggins All Parties Listed in Attachment 1 i
i 1
ATTACHMENT 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 me 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 Washington, D.C.
20555 MHB Technical Associates 1723 Hamilton Avenue Dr. George A. Ferguson Suite K School of Engineering San Jose, California 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 i
Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Ralph Shapiro, Esq.
Commission Cammer and Shapiro, P.C.
Washington, D.C.
20555 9 East 40th Street New York, New York 10016 Bernard M.
Bordenick, Esq.
j 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 3045 Porter Street Washington, D.C.
20008 i
s
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4 SNPS-1 FSAR
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9.4.6 Drywell Air Cooling System l:
9.4.6.1 Decian 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
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 10,000 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.
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.
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The units are designed to operate during al'1 normal plant operations.
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 spare.
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 i
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, end all primary containment isolation valves in the cooling water piping are closed automatically.
c 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.
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SNPS-1 FSAR 9.4.6.4 Tests and Inspections C;
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 WS.
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I CONTAINMENT SYSTEMS
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ORYWELL AVERAGE AIR TEMPERATURE 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 average 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 SHUTOOWN 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
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average of the temperatures at the following locations and shall be determined to be within the limit at least once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:
Elevation Azimuth a.
68'-0" 13*, 320*
b.
80'-0" 190*, in CR0 area c.
83'-0" 25*, 135*, 255' d.
110'-0" 165*, 350' e.
145'-0" 55*, 230*
f.
162'-6" Reactor Vessel Centerline SHOREHAM - UNIT 1 3/4 6-12 l
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PLANT SYSTEMS PRDOF & R3H COPY 3/4.7.9 AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION 3.7.9 The temperature of each area shown 'in Table 3.7.9-1 shall be maintained within the limits indicated.
kPPLICABILITY: 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:
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 rsquired by a.
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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TABLE 3.7.9-1 AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT (*F) a.
Control Room 90 b.
Ch(11er Equipmen't Room (El 63')
104 c.
Relay Room 104 d.
Emergency Switchgear Rooms 104 e.
Battery Rooms 104 4
f.
Diesel Generator Rooms 120 g.
Screenwell House 104 h.
Reactor Building - Secondary Containment 1.
General Areas 104 2.
Refueling Area 110,
1.
Reactor Building - Primary Containment 1.
Genera ~ Areas 150
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2.
Area Beneath RPV 150*
3.
Drywell Head Area 185 A
165'F during Scram.
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CONTAINMENT SYSTEMS BASES 3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY c
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.
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3/4.6.1.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE The limitations on drywell and suppression chamber internal pressure ensure that the containment peak pressure of 46.0 psig does not exceed the design pressure of 48 psig during LOCA conditions or that the external i
pressure differential does not exceed the design maximum external pressure differential of 5.7 psid. The upper Timit 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 l
value and the containment external pressure requirements are satisfied.
3/4.6.1.7 DRYW LL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 340*F during LOCA conditions and is consistent with the safety analysis.
3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM 4
The 18-inch drywell and suppression chamber purge supply and exhaust isolation valves are required to be sealed closed during plant cperation 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.
i I
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."
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SH0RENAM - UNIT 1 8 3/4 6-2 I
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PLANT SYSTEMS M y g. g}fg BASES 3/4 7.7 FIRE SUPPRESSION SYSTEMS The OPERA 8ILITY 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.
CO systems, Halon systems and fire hose stations.
The collective capability 2
of the fire suppression systems is adequate to minimize potential damage to safety related equipment and is a major element in the, facility fire protectien 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 allowed to provide an alternate means of fire fighting than if the inoperable 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, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant.
~
3/4.7.8 FIRE RATED ASSEMBLIES The OPERABILITY of the fire barriers and barrier penetrations ensure that fire damage 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.
3/4.7.9 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety related equipment will 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 req 0 ired to be OPERABLE consistent with
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the assumptions of the feedwater controller failure analysis for FSAR Chapter 15.
SHOREHAM - UNIT 1 8 3/4 7-4
.