ML20004F905

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Forwards Response to NRC Questions Re Generic BWR Scram Pipe breaks,post-accident Sampling,Mods to ADS Logic Per NUREG-0737 Item II.K.3.18,fracture Toughness & Head Spray Standby Liquid Control Sys
ML20004F905
Person / Time
Site: Fermi DTE Energy icon.png
Issue date: 06/24/1981
From: Colbert W
DETROIT EDISON CO.
To: Kintner L
Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.18, TASK-TM EF2-53-873, NUDOCS 8106260160
Download: ML20004F905 (11)


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Detroit Eaison Ei?i!!==

June 24, 1981 EF2 - 53,873 t- #

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Mr. L. L. Kintner S f3 Division of Project Management b[fl i - ,

Office of Nuclear Reactor Regulation g, .J-i - D JUN 2 51981* . C U. S. Nuclear Regulatory Commission '

Washington, D. C. 20555 g ;7 V wmot g U ,

Dear Mr. Kintner:

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Reference:

Enrico Fermi Atomic Power Plant, Unit 2  :

NRC Docket No. 50-341 i

Subject:

Position ASB-1 (Generic BWR Scram Pipe Breaks) l Position CEB-2 (Post-Accident Sampling) '

Position RSB-11 (II.K.3.18-ADS Logic Mods)

Fracture Toughness (GDC-51)

Position MTEB-3 (ISI-Head Spray /SLCS) i j

Detroit Edison's responses to these NRC Staff positions are enclosed. This informc. tion will be included in a forthcoming FSAR amendment as appropriate, l

Sincerely, )

fs ;- hkb William F. Colbert Technical Director Enrico Fermi 2 WTC:jl cc: Mr. B. Little 4,

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NRC QUESTION - CRD SYSTEM - ASB - 1 -

By letter dated May 5, 1981 we requested information regard-ing our Office of Analyses and Evaluation of Operational Data (AEOD) report entitled, " Safety Concerns Associated with a Pipe Break in the BWR Scram System." The report describes a potential sequence of events which could result from a break in the BWR scram discharge piping during a scram con-dition concurrent with an inability to reclose the scram out-let valves. Provide generic information requested.

RESPONSE

Detroit Edisons response to the May 5, 1981 letter re-garding a break in the Scram Discharge Volume is being directed to Robert Tedesco, Assistant Director for Licensing. The May 5, 1981 letter requests a generic evaluation of the Office of Analysis and Evaluation of Operational Data (AEOD) report entitled, " Sa f e t,%

Concerns Associated with a Pipe Break in the BWh Scram

. System." Detroit Edison has reviewed the responses in the General Electric NEDO-24342, !' General Electric Evaluation in Response to NRC Request Regarding BWR Scram System Pipe Breaks," and has confirmed that the generic responses are applicable for Fermi 2.

Detroit Edison is aware of the coordinated efforts being made by the NRC and General Electric on behalf of che BWR licnesees.to resolve this issue. A generic evaluation of NEDO-24342 Da being made by the NRR and is scheduled for late July, 1981. The applicant will review the Fermi 2 Scram Discharge Volume design for compliance with the criteria of the generic safety evaluation report in submitting the plant specific information.

J. R. Green

/dk 6-24-81

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DECO RESPONSES TO NRC l 1-II.B.3 - Requirement' for Monitorina Chloride in Reactor Coc tant -

Detroit Edison's exception to the monitoring of chloride as stated in draft Amendment 36 is based on typical BWR operation in which chlorides are closely monitored and controlled. With this philosophy-in mind, one would not expect. chlorides in an accident situation-unless, for example, heat exchanger tube leaks or massive domineralizer performance reductions occurred simultaneously. In either case,

-process instrumentation would. provide an indication immediately of- a possible chloride intrusion and corrective action would-be implemented..

Chloride monitoring of. reactor coolant will'be performed if process.

instrumentation indicates - a possible chloride intrusion. This situation would initiate the chloride monitoring program within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. The program will involve off-site analysis of a grab sample and/or continuous on-line chloride analysis by specific ion electrode.

ile do take axception to the st 'ff response recommending the control of chlorides during an accident. Chloride control is not possible during an accident situation.

4-II.B.3 - Requirement for Monitoring Boron in Reactor Coolant e

Detroit Edison takes exception to the requirement for the monitoring 4

of boron because boron is not a normal constituent in reactor coolant. Boron monitoring will be perfotaed subsequent to boron injection.

Capability for boron analysis will be provided by of f-site analysis of a grab sample and/or continuous on-line boron analyzer utilizing i' specific ion et.ectrode.

2-II.B.3(4) - Recommendation. for Monitoring Dissolved Oxygen in Reactor Coolant Detroit Edison's exception to r.anitoring dissolved oxygen is because during an accident situation, dissolved oxygen levels cannot be controlled. Verification that oxygen is > 0.1 ppm will not initiate any corrective' action. However, monitoring of the reactor coolant oxygen for chloride stress corrosion cracking susceptibility of associated safety systems will provide input for assessment of safety boundary integrity.

Detroit Edison will monitor dissolved oxygen by providing grab sample capability for off-site - analysis and/or an on-line

. polarographic oxygen analyzer.  ;

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3%s 3.1.97 (Revision'2) Requirement for Monitoring pH in Reactor Coolant Detroit Edison ' takes exception to the recommendation for monitoring pH in the -reactor coolant to verify that pH is jt :7.0 to assure protection against chloride stress corrosion cracking. BWR operating

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philosophy dictates operation in' the neutral pH region with no provisions for~ chemical additions for pH control. In order for pH to be jt'7.0 in an accident condition,some chemical addition would be required which would be contrary to operating philosophy.

The monitoring of chloride, oxygen, and temperature is adequate to assure protaction against chloride stress corrosion cracking in safety'related systems subsequent to an accident.

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B.la Following a DBA event, the reactor water level is not maintained above 2/3 core height. In this scenario, the reactor vessel is not pressurized and as a result the intended sampic source is the discharge of the RHR pumps which are acting as the driving force for the circulation of reactor coolant.

B.lb Suppression pool atmospheric samples are taken from taps on. opposite sides of the pool proper. Each tap location is selected to maximize the distance to either a downcomer or safety / relief valve discharge sparger.

Since both of these steam sources discharge under water, there is no significant affect on the atmospheric sample.

Liquid samples from the suppression pool are obtained from the RHR pump discharge. RHR pump suction strainers are located as far as practical from both the downcomer and S/R valve discharges in order to maximize the temperature effects on pump NPSH. As a result, the sample should be representative of the mixture in the pool.

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B.2 The post accident sampling control panel and all electri- -

cally operated valves within the sample station will be supplied with restorable balance of plant power.

NUREG-0737 does not stipulate that the post-accident sampling system is classified as a safety grade system.

Restorable power provides the necessary reliability for this system.

B.3 The sampling station is located in the Auxiliary Build-ing on the first floor (elevation 583'-0") between columns G12 and H12. From the sample station, the samples are either manually carried (small cask) or wheeled on a dolly (large cask) to the analytical labora-tory or to an exit for off-site shipment. The laboratory and exit are at elevation 583'-6" and are within 150'ft.

of the sample station. The route to be traversed is through a doorway into the Turbine Building with a turn north to the analytical laboratory or south to the exit.

It is estimated that these routes can be conservatively walked in less than 2 minutes. The transit doses (MREM) for a 2 minute walk from the Main Control Room to the sample station has been calculated as 2.1 for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,

. 6.9 for 1 day, 4.0 for 1 week, and .5 for 1 month. The analytical laboratory is in the general area of the control room except that it is at the same elevation as and closer to the sampling station. It can be extra-polated that the transit doses wculd be conservatively equal to or less than the doses identified for the main control room.

B.6 The post-accident sample collection system will be operated on a semiannual basis to insure operability and provide the necessary operator training. Specific procedures will be generated to include procurement, transporation and analysis of samples. Diluted and undiluted liquid and gaseous samples will,be obtained from the reactor coolant, drywell containment, and suppression pool.

This will allow verification of the operation of all com-ponents on the sample station and permit the operator to familiarize himself with the operation of sample station.

All analytical instrumentation will be calibrated using approved Radchem procedures and traceable to the National Bureau of Standards.

B.4 The approved Radchem procedures used for the monitoring of reactor coolant are traceable to the National Bureau of Standards (NBS) or the American Society for Testing and Materials (ASTM) and qualified for accuracy and pre-cision in an accident environment. This instrumentation will meet or exceed the accuracy ranges specified in Regulatory Guide 1.97 (Revision 2) .

B.5 When an accepta.ble technique for estimating fuel damege is prepared, Detroit. Edison will review :it for inclusion in Enrico Fermi 2 procedures .

RSB - 11 (II.K.3.18 - MODIFICATIONS TO ADS LOGIC)

Detroit Edison commits to complete and test the modifications to ADS logic required by NUREG - 0737 Item II.K.3.18 prior to fuel loading. , FSAR Page H.II.K.3.18-2 will be revised to so indicate this schedule committmeat in a forthcoming FSAR amendment.

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I GDC 51 - MSIV Ductility In response to a concern by NRC staff, Detroit Edison submits pertinent data on the Quaker Alloy Casting Co. Main Steam Ifel*hb

' Valves as submitted for the TVA X20 and Grand Gulf 1 plants. -

This data was taken from the SSES response to NRC questions 251.1 - 251.5. This data indicates that additional heat treatment increases the ductility of the MSIV.

We believe this data vill resolve the NRC staff concern on this iten.

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Table 11 Sh. No 9

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Project TVA X20

  • Valve MSIV Component Body Applicable Code ASME Sect. III, 1974 with summer 1975 addenda.

Valve Vendor Atwood & Morrill Co. , Material Vendor: Quaker Alloy Casting Co.

Material Specification - ASME SA216 Grade WCB H;at No. F3547 Chemical Composition (wt. :) - C/.23 Mn/.88 Si/.38 p/.016 S/.015 Al/NA Grain Size (ASTM lio.) - NA Heat Treatment 1700/172f F (6 hr. 20 min.) air cool + temper 1345CF (6,hr. 45 min.) air cool + post weld 1200/1225 F (6 hrs. 30 min) air cool Charpy V - Notch Impact Toughness Test Temperature : +60'F Ft-lb : 66, 56, 54 Mils: 53, 50, 53

% Shear : 40, 40, 40

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Table 11 Sh. N9 )__

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Project Grand Guit i Valve MSIV Component Body Applicable Code ASME lect. !!!, 1974 Valve Vendor Atwood 5 feorrill, Co.

Material Yendor Quaker Alloy Casting Co.

Material Specification - ASME SA 216 Grade WCB Heat No. - F6406 Chemical Composition (wt ".) - C/.23 Mn/.89 51/.53 P/.019 S/.012 A1/f4A Grain Size (ASTM No.) - NA

+ terper Heat Treatment 1680/1710*F (5 hrs. 30 min) air cool 1350*F (5nr. 30 min) air cool

+ Post weld 1200 F (6 hr.) air cool Charpy V - Notch impact Toughness Test Ter perature +60'F Ft-lb 32,31,34 Mils 33,32,31 1 Shear 40',40,40 l

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43 NA - Not Available

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, , ,., . ' M-$g ,3 rejyf c 4 N4C chy geSi4L The Detroit Edison Company agrees to incorporate the Head Spray and Standby Liquid Control Systems into the Class'2 in-service inspection program. The selection criterla-for welds

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in these systems will be the same as described in Class 2 Sys-tem Weld Program, Enrico Fermi Atomic Power Plant Unit 2, 30, April 1981.

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