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4 LONG ISLAND LIGHTING COM PANY #Ed"O w SHOREHAM NUCLEAR POWER STATION P.O. DOX Gia, NORTH COUNTRY ROAD e WADING RIVER, N.Y.11792 ?
February 5, 1979 SNRC-357
Mr. Harold R. Denton, Director Office of Nuclear Reacter Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322
Dear Mr. Denton: During a meeting held in your office on Decemoer 14, 1978, the status of tM remaining SER open items were discussed with the NRC Staff. As a result of those discussions, it was agreed that we would provide additional information clarifying a number of the open items.
Enclosed herewith are fifteen (15) sets of the additional information provided in response to those requests for SER open items numbered 6, 7, 11, 13, 19, and 20 plus confirmatory issues numbered 4 and 12. In addition, we are enclosing draft copies of a pending FSAR figure change to reflect an improvement in the intake canal slope stability, the bases of which are discussed in FSAR Section 2.5.5 and Appendix 2L. Subsequent submittals will be made on or about Feb uary 28, 1979 which will provide additional information relative to SER open items 2, 5, and 20 plus confirmatory issues 1, 8, 15, 16 and 17. Information for any remaining Applicant-action items will be forwarded to the NRC as it becomes available.
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ry t uly yours, MS DOCUME;T CO?iTNilS
.; POOR QUAUTY PAGES o - [D M .
. P. No arro, Project' Manager Shoreham Nuclear Power Station
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. SHOREHAM OUTSTANDING ISSUE NUMBER 6
RV SUPPORTS AND INTERNALS EVALUATION
This issue concerns the evaluation of RV supports and internals, particularly following ruptures of pipes in the vicinity of a reactor vessel nozzle safe-end.
A description of the analysis methods utilized in this evaluation may be found in the Plant Design Assessment Report (DAR) for SRV and LOCA Loads - Shoreham Nuclear Power Station, Unit 1, Revision 3, November 1978, Section 9.2.2. The DAR was forwarded to the Commission on December 11, 1978 via letter SNRC-347.
The attached table provides a summary of results for the loading combination of normal plus annulus pressurization plus the safe shutdown earthquake. In every case, the calculated value is within allowable limits. , , .
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. SHOREHA'1 REACTOR VESSEL SUPPORT + INTERNALS LOAD ASSESSMENT . . n ' - . . * Calculated Component Load Combination Criterion Allowable 2 < 15.3 ksi (Stress) Support Skirt N + (AP2 ggg )k' Faulted 43.1 ksi (Stress) 2 Faulted 60.0 ksi (Stress) < 40.0 ksi (Stress) . Stabilizer Bracket N + (AP + SSE )b 1292 kips (Load) 1093 kips (Load) Stabilizer N + (AP + SSE )' Faulted ' 2 2L < 43.2 ksi (Stress) Shroud Support N + (AP -+ SSE )' Faulted 55.9 ksi (Stress) 2 54.6 psi (AP Buckling) 30.3 psi (AP Buckling) Core Support N + (AP2 + SSE ) Faulted 2 3.7 ksi (Stress) Shroud (Buckling) N + (AP2 , ggg )4 Faulted 17.2 ksi (Stress)
* 2 12.3 ksi (Stress) Top Guide Ocam N + (AP2 ggg )h Faulted S0.7 ksi (Stress) 2 8.7 kips (Load) 4.4 kips (Load) - . Top Guide Holdown Latch N + (AP + SSE ) Faulted
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. SHOREHAM OUTSTANDING ISSUE NUMBER 7
SHOREHAM MSIV ENVIRONMENTAL QUALIFICATION
The Shoreham FSAR references Report APED 5750 which describes the main steam isolation valves (MSIV) and control systems and documents the analysis and the test program conducted to resolve any concern regarding closure of the main steam isola- tion valve under main steam line break conditions. The approach taken was to test a typical full-sized valve under conditions which approximate as closely as possible the event of a steam line break outside the drywell. The primary objective was to demonstrate proper closure of the valve under such conditions. The test facility was designed to provide the appropriate conditions of pressure, flow rate, and quality at the test valve within the limitations of the existing coal fired power plant to which the test facility was attached.
The test was performed on an MSIV manufactured by the Crane Company. The MSIV's supplied for the Shoreham Plant are manufactured by the Rockwell Corporation. The information provided here includes the commonality and differences between the Crane and Rockwell valves to provide assurance that Test Report APED 5750 is an appropriate reference for Shoreham MSIV environmental qualification.
1. Common to both Crane and Rockwell Valves are:
a. Y-type body pattern;
b. Use of both spring and air as closure mechanisms;
c. Design pressure and temperature of 1250 psig and 575 F, respectively;
d. Adjustable closing time of three to ten seconds.
2. Differences between the Crane and Rockwell valves are:
a. Solenoid #3 for controlling the pilot air operated valve is a 125 VDC for the Rockwell design vs. 120 VAC for the Crane design.
No degradation of safety results from the change from AC to DC solenoids. -: ,
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b. On each Rockwell valve there are three (3) solenoids mounted by a nipple on the outside of a Hoffman Nema-4 junction box. Sol #1 - ASCO Cat. No. HTK8320A20 - 60 Hz AC Test Solenoid Sol #2 - ASCO Cat. No. HV166592 - 60 Hz AC Pilot Solenoid Sol #3 - ASCO Cat. No. HV166592 - 125 VDC DC Pilot Solenoid The #3 and #2 solenoid valves are separete on this Crane design vs. a double valve on the Rockwell design. These arrangements however are considered to represent the same effective configuration. The hardware of the double valve is ported to work as two valves in one frame. APED-5750A and Supplements 1 & 2 documented the results of a successful LOCA environmental test on AC solenoid #1 which is common to both valves. The solenoid vendor, Automatic Switch Company (ASCO), has confirmed that all three solenoids have an explosion proof, water-tight enclosure, NEMA-7, with a high temperature coil. Thus, the qualification of solenoid #1 applies to solenoid #2 and # 3 as well. It should be further noted that all three solenoids are applied as fail safe devices.
The test results reported in APED-5750A and Supplements 1& 2 combined with the design requirements of the Shoreham MSIV's provide sufficient assurance of MSIV environmental qualification at Shoreham. SIIOREIIAM OUTSTANDING ISSUE NUMBER 11
SilOREHAM CONFIRMATORY iSSU." NUMBER 4 .. SUMMARY REPORT OF TARGET-ROCK
SAFETY / RELIEF VALVE ENVIRONMENTAL TESTING
Test Performed on Specimen Results
1. Aging Simulation Solenoid valve and air operator remained operable 7 Radiation (1.9 x 10 Rads) following exposure described at left. Thermal (28S* F for 480 hrs) Mechanical (8000 cycles)
2. Emergency Ambient Solenoid valve and air operator remained operable following exposure described at left.
Time Temp. Pressure Minutes ~F PSIG 0-1 337 65 2 - 20 334 - 346 48 - 50 20 - 181 342 48 181 - 299 85 - 94 0 300 - 30l* 343 65 301 - 480 344 - 347 46 - 51 482 - 650 322 46 650 - 5521** 290 - 301 25 - 26 3. Final Radiation Aging Solenoid valve and air operator remained operable lx3 x 10 7 Rads following expasure described at left.
4. On conclusion of test, solenoid valve and air operator assembly was still operable and pneumatic system leakage was .115 SCFH. Therefore, this equipment is qualified for its intended service. 5. The environmental qualification requirements for the Shoreham LOCA ' profile are shown in FSAR Table 3.11.1-1, Safety Related Equipment and Components and Associated Accident Environments (Inside Primary Containment). In accordance with normal procedures, the complete test report is
available for audit. . , * Simulated -LOCA- test started at Time;= 300. minutes. ** The final' interval of 5521 minutes at 290 F is'more sev'ree than the FSAR long-term post-LOCA condition of 150 F for 99 days.
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* SHOREHAM CONFIRMATORY ISSUE #12 POSITION INDICATION FOR THE RHR CROSS-TIE VALVE
In accordance with a prior agreement made with the NRC Staff, redundant closed position indication has been provided for RHR cross-tie valve lEll*MOV-050 (Ell-F010). The attached exhibit (7 pages) documents the implementation of this agreement. wo , f j s,y,f.i.p.f'.. ny d;tQ , ' ~
._._ __ - FIELD DEVIAll0H DISPOSITION REQUEST ' JAN 1a m * P4UCLD/.a ENE RGY OlVISIONS y f, g. , ,
FDDR NO. -01 -MO SHEET 1 OF 3 PROJECT ' 'UU 'Y 5 DATE ORIGINATED I,/ I 8 / ~I 9 TELECON CR TWX APPROVAL BY: * *2t .sts EQUIPY ENT (DE SOPIPTION AND/CR P/PL) }{llP@l PA"7L C'.' G. ' P E C. E TC. N O. SH ECT NO. R EV NO. C.*.G. SP E C. E T C. T I T L E
791E 4IBTF S K 13,14 17_ F:.H f; Eterneo rat.Y
DESCRIFT ON OF OEVIATION: IMPACT CLASSIFICATICN CATEGORYl O Valve Ell-F010 does not now have a redundant CATEGORY H b PRIORITY CLASSIFICAT:ON position indicating light. MRC requirca such lightD EVERGENCY O UAGENT O bs added to Ell-S7 circuitry. nou riN r n APPf}OVJ D P.'s ,, J, . f.i. bE SIGN E r.Li'4E E R DAIL
CCB CHAIRY AN
. M AT'LS AFPL E N0iN E ER SUGGESTED DISPOSITION b#- U - 1 K 2.n w ||1 ~ ' ,f,. 't hlf n .:.: %&s [.n ~ Add redundant positica indicating light for Ell-F010 .
# to panel H11-P601. /'d'J'/....,/4 . . 4i -
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F I E L D M /. N /. ' ( A DIS APPROV ED BY;
D Af f SUPPLE P.'E NT INST REQUiHED YES NO NU'.* D r n 0 ECN E I d O ''' [ 7 #b' FlN AL DISPOSITICN D A pp R G p c u D A u f l'0 5 IND LICt+F TD H //-f(, O l O p Fog o STnietetiCN coeE 2 - 7 g Tu/s m /2, PER INEE T5 ,yyny,t gxycns,t , M k 194- _
F'ERFORM O A IN ACCORDANCEWITH - G.E. SITE PRACTICES AND PROCEDURES OTHER PLANTS AF FECTED YES O ICENTIFY THK CHAMGE~ OcEs A>cT DE-GICADE THE S)'STSM NO W SAFETY /REllADILITY REVIEW REQUIRED: YES O NO% " Eto v on< cacca No- $TATEMENT OtSPOSITION CO:*,'LET ED E, AT E
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[ES; ABLCChECK VV 1 500 SriUT0FF VV 1128 2.62 1 512 D E11-F010 M0V050 E 11 -F050B A0V081B * '' A - CONT AlhF.ENT SPRAY A MANUAL 1128 1 1 501 HEADER VV L3\| 513 INJECf10N VV E11-F016B MOV038B _Y E l -F0003 HV0713 y . CONTAINMENT SPRAY RHR PUMP B 1128 1 514 SUCTION yy 1122 2.62 1 502 HEADER VV 2.62 . E11-F0218 l'.0V039B E 11 -F004B M0V0318 _ - I ASC 32B6TF sH 3 1 E . . - - F PpK 45-01- 600 - SH . . 7 C ,. 7 I I [ i 53 ~ =::::. 1 C't13bI6NaV 2 | 3 - - . SHOREHAM OUTSTANDING ISSUE NUMBER 13 SHOREHAM RCIC SYSTEM DESIGN As s'.ated in Shoreham FSAR Section 5.5.6.1, the RCIC system piping and equipment, including support structures, are designed to withstand effects of the DBE/OBE without a failure that could lead to a release of radioactivity in excess of the guideline values in published regulations. FSAR Table 3.2.1-1 XII provides further verification that the RCIC system piping and equipment meets seismic Category I requirements. SHOREHAM OUTSTANDING ISSUE NUMBER 19 PLANT PROCESS COMPUTER In discussions between G. E. and the NRC Reactor Systems and C&I Branches on November 20 and 21, 1978, the failure of the reactor feedwater level sensing subsystem was shown to have resulted in nooCPR effects on the four leading limiting transients. The failure of the L8 instrumentation, including computer input, would result in a negligible a CPR effect (0.02). Therefore, it would seem appropriate to further explore the failure aspects of this subsystem since even given their failure, they do not result in any effect on the previously described (FSAR) transient analysis. A failure of the computer inputs, as discussed above, would have a negligible effect onaCPR. SHOREHAM OUTSTANDING ISSUE NUMBER 20 FIRE PROTECTION Question 6 (Section 1, Position D3) : In Class IE cable trays running vertically and horizontally, solid bottom trays with covers are used. Demonstrate that the reactor building fire detection system is capable of detecting both an internally generated fire and an exposure fire that would ignite the cables in these covered trays, and describe the firefighting techniques that will be used to extinguish the fire in the event that one should occur. Response: Throughout the reactor building, an early warning detection system (smoke and temperature detectors) is installed with alarm and annunciation in the control room, thus locating the fire in its incipient stage. An internally or externally initiated fire, resulting in burning of cables in the cable tray, will be detected by these smoke detectors, since fires in retardant type cables are smoldering and accompanied by considerable smoke. These detectors are UL listed and shall be field tested for their intended function. Cable tray covers ensure that a f. ire within one cable tray will not propogate to the adjoining cable trays. However, in the unlikely event that a fire does originate in the cable tray, the fire shall be contained and limited due to the low inventory cf combustibles and the self extinguishing properties of the cable insulation used. Water hose station and portable extinguishers are provided for fire suppression. In the case of fire in the cable tray, Nater shall be sprayed en che cable tray to keep it cool and thus prevent the reignition of fire. . 6 ' . Pending FSAR figure change to reflect improvement in intake canal slope stability. The bases are discussed in detail in FSAR section 2.5.5 and Appendix 2L. , LEGEND 1 6 , -y- 2 TON 00LOJ , * * IS POUNO STONE g , - h',*' . IIS .*OUND STONE * ~ ,. 230 POUND STONE " - ' - ' 3000 pound STONE * ' , gs ' / *, CA gy pguMG ! [ 4400 POUND STONE ,g. ~ * Y 6000 POUND STONE + E', * ' ,q'.v s . ., o 4 '' ' "* * .- - - I ' f LOCATION OF INTAKE C AN/. - '*# 0, sccTiCNsttlGs. 2.4.e -ss,'e,c s o) ~y > sr.ntLH'*E'*u { - , ', . 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