ML20203H806

Forwards Request for Addl Info Re Prairie Island,Unit 2 Control Rod Drive Mechanism Leak

ML20203H806
Person / Time
Site:	Prairie Island
Issue date:	02/18/1998
From:	Wetzel B NRC (Affiliation Not Assigned)
To:	Richard Anderson NORTHERN STATES POWER CO.
References
TAC-MA0751, TAC-MA751, NUDOCS 9803030331
Download: ML20203H806 (5)

Text

1. I Mr. Roger _O. Anderson, Director February 18,'1998 Licensing and Management lasues Northem States Power Company 414 Nicollet Mal; Minneapolis, MN 55401 SUBJECTr REQUEST FOR ADDITIONAL INFORMATION CONCERNING THE PRAIRIE ISLAND UNIT 2 CONTROL ROD DRIVE MECHANISM LEAK (TAC NO. MA0751)

Dear Mr. Anderson:

On February 10,1998, members of the NRC staff (both the Office of Nuclear Reactor Regulat:on and Region Ill) conducted a telephone conference with personnel from the Prairie Island NJelear Generating Plant concerning the leak on the G9 control rod drive mechanism (CRDM) on Unit 2. During the telephone call, the staff asked Northern States Power's plans for addressing both Unit 2 and operability for similar components on Unit 1 which was operating at 100 percent. Enclosed is a list of questions that were discussed during the telephone call.

Please provide a response to these questions or a schedule for addressing the unanswered questions by February 23,1998, so that the staff can promptly assess the generic aspects of this issue.

If you have any questions regarding this matter, please contact Beth Wetzel at 301-415-1355.

Sincerely, f

ORIGINAL SIGNED BY Deth A. Wetzel, Senior Project Manager Project Directorate lil-1

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i>1 Mr Roge. O. Anderson, Director Prairie Island Nuclear Generating Northem States Power Company Plant cc:

J. E. Silberg, Esquire Site Licensing -

- Shaw, Pittman, Potts and Trowbridge Prairie Island Nuclear Generating

- 2300 N Street, N. W.

Plant -

- Washington DC 20037 Northem States Power Company 1717 Wakonade Drive East Plant Manager -

Welch, Minnesota 55089

' Prairie Island Nuclear Generating Plant Triba! Council Northem States Power Company Prairie Island Indian Community 1717 Wakonsde Drive East ATTN; EnvironmentalDepartment Welch, Minnesota 55089 5636 Sturgeon Lake Road Welch, Minnesota 55089 Adonis A. Noblett Assistant Attomey General Office of the Attorney General 455 Minnesota Street Suite 900 St. Paul, Minnesota 55101-2127 U.S. Nuclear Regulatory Commission Resident inspector's Office 1719 Wakonade Drive East Welch, Minnesota 55089 9642 Regional Administrator, Region 111 U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Mr. Jeff Cole, Auditor / Treasurer

'Goodhue County Carthouse Box 408 Red Wing, Minnesota 55066-0408 Kris Sanda, Commissioner Department of Public Service 121 Seventh Place East

- Suite 200.

St. Paul, Minnesota 55101-2145 November 1996

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• a Request for AdditionalInformation Prairie triand Unit 2 Corrective Ardi~m

.1.

Please, provide the NRC with a copy of your root cause' analysis of the flaw indications in the Unit 2 control rod drive mechanism (CRDM) pertial length housing. The following are points that you may want to consider in your root cause analysis:

A. A description of the design of the partiallength CRDM housings including description of

. the materials of construction, fabrication methods, and design aspects for the housings -

and what the design code of record is for the part-length CRD" housings.

B. The A ferrite requirements, if any, specific for the stainless steel used in the fabncation of the full length CRDM housings in the Prairie Island reactor vessel designs and any records, if available, with respect to the A-ferrite requirements.-

1 C. A discussion'of the impact of corrosive environmental conditions and thermal-hydraulic loads on the detected flaw indications in the No. G9 penetration.

D. Tr.e failure mechanism for the flaw indications in the No. Gg partial length CRDM housing nozzle, and the basis for concluding that the mechanism is considered to be the predominant mode of failure for the housing,

2. Summarize your intended corrective actions for Prairie Islar.d Unit 2 and the respective -

schedule for these actions.

Prairie Island Unit 1 A===== ment

1. Provide a safety evaluation of the flaw indications in the CRDM housing and a jus *r ation for concluding the Prairie Island Unit 1 plant can be safely operated for the remainder of the current operating cycle. The following are items that should be considered in this -

assessment:

A. A description of the NDE [ nondestructive examination) methods currently used to analyze the flaws in the CRDM housing in Unit 2,' and a description of the techniques used to qualify the NDE techniques for use.

B. An evaluation of the orientation and size of the flaws in Unit 2 detected by the NDE

- personnel. Include in this assessment the following analysis elements:

(1) A description of the geometric orientation of the flaws and the orientation of the flaws with respect to each other.

(2) A conservative determination of the flaw sizes in the CRDM housing.

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ENCLOSURE

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2-(3) A best estimate determination of the relative proximity of the flaw indications in the No. Gg CRDM housing to one another, and an assessment of whether or not the 4

flew indications need to be considered as one indication in accordance with the

. ASME [American Society of Mechanical Engineers) Code proximity rules.

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- C What are the worst-case sizes of fiaws that u;e postulated to be present in partiallength CRDM housings in the Prairie Island Unit 1 RPV [ reactor pressure vesselj head and what is your assessment of these flaws? Based on the results of the Unit 2 flaw evaluation; justify why any flaws postulated in the Prairie Island Unit 1 partial length s

CRDM housings would be considered acceptable to remain in service without necessitating a repair of the flaws over the remainder of the Unit 1 operating cycle.

Include in your flaw evaluation allowances for crack growth of the flaws that are postulated to occur from the start of the Unit 1 operating cycle to the currer t tire-1 for the remainder of the current operating cycle. What are the margins aps*

oility.

- for both normal operating and design-basis loads?

D. Provide a description of any enhanced leakage monitoring being conducted for the -

Prairie Island Unit i reactor coolant system. Provide a discussion of any additional compensatory measures taken on Prairie Island Unit 1.

E. Provide an analysis of the limiting postulated event (s) for the partial length CRDM housings. Include in your assessment the f611owing elements:

(1) What is the limiting loss of-coolant accident (LOCA) associated with a failure of a -

partiallength CRDM housing?

(2) What are ine estimated consequences of a postulated LOCA resulting from a complete failure of a partiallength CRDM housing? Please address the following questions regarding consequences of the postulated event:

a. What is the potential for missiles during the postulated event?

b. What would the consequences be from hydraulic loads, jet impingement loads, and postulated missiles resulting from the, postulated event? In particular, address what the consequences would be to the full-length control rods, reactor core, and linec penetrating the Prairie Island reactor vessels during the postulated event as a result of postulated hydraulic loads, jet impingements, and

missiles,

c. What are the dynamic effects as a result of the postulated full CRDM housing break at the indicated crack location (including a description of the dynamic analysis models used to assess the jet thrust force caused by jet flow from the postulated break, and a discussion stating whether or not the models are consistent with the Guidelines provided in NRC Standard Review Plan Section 3.6.2.)?

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3-(3) What is the minimum number of ECCS [ emergency core cooling systerni pumps (e.g., Icw pressure safety injection (RHR [ residual heat removal]) pumps, and high pressure safety injection (charging) pumps) that woulC, be required to mitigate the consequences of LOCAs up to and inclusive of the size of the limiting LOCA associated with a co:'plete failure of a partiellength CRDM housing in the Prairie Island units?

(4) What la the minimum number of control rods required to bring the reactor core to a subcriticallevel for postulated LOCAs up to and inclusive of the size of the limiting LOCA associated with a complete failure of a partial length CRDM housing in the Prairle !sland units?

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