ML20024B618
| ML20024B618 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 07/07/1983 |
| From: | Lois L, Requa G, Vissing G Office of Nuclear Reactor Regulation |
| To: | HARTSVILLE GROUP |
| References | |
| NUDOCS 8307110037 | |
| Download: ML20024B618 (97) | |
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s' UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION s.
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of CAROLINA POWER AND LIGHT COMPANY Dacket No. 50-261 (H.B. Robinson Steam Electric (Steam Generator Repair)
Plant, Unit 2)
NRC STAFF RESPONSE TO INTERROGATORIES 0F THE HARTSVILLE GROUP The fhllowing are the responses of the NRC Staff to the Interroga-tories of the Hartsville Group. The Staff and the Intervenor have agreed which of the interrogatories submitted are to be responded to by the s
N Staff. We are not responding to any of the interrogat6 ries related to Contention 2 since the Staff decision to publish an Environmental Impact Statement renders Contention 2 moot.
N Contention 1 Questions No. 21 through 55 INTERROGATORY 21 Do you agree that CP&L has been responsible for a history of repeti-tive noncompliance with N.RC rules and regulations?
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RESPONSE
As detailed below, in$t'snces of repetitive noncompliance have been_
in isolated areas of Plant operation.
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a 2-CP&L as the Licensee for the H. B. Robinson 2, the Brunswick 1 & 2, and the Harris facilities is responsible for any noncompliances with NRC rules and regulations.
Three apparent instances of repetitives noncompliances at the Brunswick facility have occurred in the areas of Q-list identification of instruments, calibration of instruments used to verify technical specification requirements and the inadequate implementation of correc-tive action.
a.
The repetitive Q-list noncompliance is discussed in IE Reports 325/82-10 and 324-82-10, and IE Reports 325/82-45 and 234-85-45 CP&L's responses to these violations are dated May 24, 1982 and March 18, 1983.
b.
Repetitive violations concerning the calibrations of instruments, used to verify technical specification requirements, are discussed IE Reports 325/82-16 and 324/82-16, and IE Reports 325/82-45 and 324/82-45.
CP&L responses to those violations are dated August 3 and 26, 1982, and March 18 and April 22, 1983.
c.
The repetitive violation concerning inadequate implementa-tion of corrective action is discussed in IE Reports 325/81-20 and 324/81-20, and IE Reports 325/82-08. CP&L's responses to these violations are dated October 6, 1981 and May 12, 1982.
Apparent instances of repetitive noncompliances at Robinson 2 are identified below.
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a.
Overexposures in 1981 associated with steam generator eddy current and tube plugging activities. These overexposures were either calculated or as measured by dosimetry.
IE Reports 261/81-10, 81-11, 81-17 and 81-24 apply.
b.
Failure to establish and implement an adequate drawing control program.
IE Reports 261/81-12 and 261/82-20 apply.
Apparent instances of repetitive noncompliances at Harris are identified below, a.
Improper curing of concrete test cylinders, can be iden-tified in IE Reports 400/79-17,400/81-24,400/82-16.
i b.
Failure to identify and correct nonconforming conditions or electrical cable tray supports, identified in IE Reports 400/81-25/ 400-82-24, 400/82-28.
c.
Failure to protect equipment from the environment and adjacent construction activities, identified in IE Reports 4//880-26,400/80-27,400/81-13,400/82-02,400/82-05.
l The above instances of noncompliance in isolated areas does not lead the Staff to a conclusion that CP&L has any wide-spread problem of repetitive noncompliance with NRC regula-tions.
i INTERROGATORY 22 l
If your response to Interrogatory 21 is negative, explain in detail i
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the respects in which you do not agree.
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RESPONSE
Not applicable.
INTERROGATORY 23 Do you agree that CP&L has been responsible for breakdowns in corporate and facility management controls in the areas of corporate oversight, facility management and operations, and problem identifi-cation and correction which'suggest a programmatic failure?
RESPONSE
CP&L as the Licensee is responsible for instituting corporate and facility management controls in the areas ifentified. Weaknesses which can be categorized as programatic in nature, have been identified in certain functional areas at the Brunswick and Robinson 2 facilities.
INTERR0GATORY 24 If your response to Interrogatory 23 is affirmative, describe each such breakdown in detail.
RESPONSE
A programmatic failure in the area of surveillance activities was identified at the Brunswick facility. This breakdown was discussed in IE Reports 325/82-28 and 324/82-28. The surveillance program at Robinson 2 was reviewed in detail subsequent to the identification of the Brunswick problems, but no similar breakdowns were identified. This Robinson 2 review is discussed in IE Reports 261/82-27 and 82-35.
l At Robinson, the NRC has identified breakdowns in management controls in the areas of health physics controls. This documented in IE Reports 261/81-07,81-10,81-17,81-24,81-11.
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While the above indicates breakdown in renagement controls with respect to specific problems, they do not indicate a failure in the overall program of management controls which covers a much broader spectrum of programs.
INTERROGATORY 25 What is the basis for your response to Interrogatory 24? Identify all documents, testimony or oral statements by any person and legal requirements on which you reply in support of your position.
RESPONSE
Basis for response are the IE Reports listed in response to Inter-rogatory No. 24.
INTERROGATORY 26 If ycur response to Interrogatory 23 is negative, explain in detail the respects in which you do not agree.
RESPONSE
Not applicable.
9 INTERR0GATORY 27 Describe in detail each CP&L violation of NRC operating procedures, rules and regulations categorized at Severity Level I pursuant to NRC Enforcement Policy.
RESPONSE
None.
INTERROGATORY 28
Describe in detail each CP&L violation of NRC operating procedures, rules and regulations categorized at Severity Level II pursuant to NRC Enforcement Policy.
RESPONSE
None.
INTERROGATORY 29 Describe in detail each CP&L violation of NRC operating Interrogatory is(SIC).
RESPONSE
Question is incomplete and not understood.
1 INTERROGATORY 30 Describe in detail each CP&L violation of NRC operating procedures rules and regulations categorized at Severity Level III pursuant to NRC Enforcement Policy.
RESPONSE
Severity Level III Violations at Robinson 2 since the establishment of the present severity level system on October 7,1980.
l a.
Personnel Monitoring 1)
IE Report 261/81-10(item 81-10-01)andEA-81-46 2)
CP&L response letters dated June 17 and 30,1981 b.
Calculated Overexposures 1)
IE Report 261/81-10(item 81-10-03)andEA-01-46 2)
CP&L response letters dated June 17 and 30,1981 l
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c.
Inadequate Surveys and Overpressure 1)
IE Report 261/81-24(item 81-24-01)andEA-82-07 2)
CP&L response letters dated December 31, 1981 and January 5,1982 d.
Failure to follow procedures 1)
IE Report 261/81-24(item 81-24-02)andEA-82-07 2)
CP&L response letters dated December 31, 1981 and January 5, 1982 e.
Improper Radioactive Waste Shipment 1)
IER261/82-40(item 82-40-01) 2)
CP&L response letter dated December 31, 1982 Severity Level III violations at the Brunswick facility since the establishment of the present severity level system on October 7,1980.
a.
Inadequate Survey and Overpressure-1)
IE Reports 324/81-16 and 325/81-16 and EA 81-77 2)
CP&L response letter dated October 30, 1981 l
l b.
Inoperable Reactor Water Level Instrument and Failure to take adequate corrective action.
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IE Reports 324/82-02 and 325/82 02 and EA 82-75 2)
CP&L response letter dated August 16, 1982 i
o 8-c.
Surveillance Program Inadequacies and Inadequate Corrective Action 1)
IE Reports 324/82-28 and 325/82-28 and EA 82-106 2)
CP&L response letter dated May 2, 1983 d.
Failure to operate the facility in accordance with the facility operating license and Technical !pecifications 1)
IE Reports 324/83-03 and 325/83-03 2)
Response due June 25, 1983 Security Level III violations at the Harris facility since the establishment of the present severity level systems on October 7,1980 None.
INTERR0GATORY 31 Describe in detail each CP&L violation of NRC operating procedures, rules and regulations categorized at Severity Level IV pursuant to NRC Enforcement Policy.
RESPONSE
Severity Level IV Violations at Robinson 2 since October 7,1980 a.
Inadequate procedure 1)
IE Report 261/81-03(item 03-02) 2)
CP&L Response dated March 23, 1981 b.
Failure to follow procedures 1)
IER 261/81-07 (item 81-07-29) 2)
CP&L response dated July 30, 1981
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c.
Failure to control modification 1)
IER 261/81-08 (item 81-08-01) 2)
CP&L response dated April 30, 1981 d.
Inadequate maintenance program
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1)
IER261/81-15(item 81-15-03) 2)
CP&L response dated June 23, 1981 e.
Failure to perform adequate 50.59 review and failure to report 1)
IER261/81-19(item 81-19-03) 2)
CP&L response dated August 21, 1981 f.
Inexperienced health physics technician 1)
IER261/81-24(Item 81-24-03)andEA-82-07 2)
CP&L response dated January 5,1982 g.
Failure to control post maintenance testing 1)
IER 261/81-27 (item 81-27-33) 2)
CP&L response dated November 18, 1981 h.
Failure to implement modification controls 1)
IER261/81-36(item 81-26-01) 2)
CP&L response dated March 30, 1982 1.
Failure to maintain procedures
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1)
IER261/81-26(item 81-36-02) 2)
CP&L response dated March 10, 1982 1
j.
Failure to perfonn safety review 1)
IER261/81-36(item 81-36-03) 1 2)
CP&L response dated March 10, 1982 k.
Inadequate procedure 1)
IER26/182-08(item 82-08-01) 2)
CP&L response dated April. 23,1982 1.
Failure to perform timely audit 1)
IER271/82-16(item 82-16-04) 2)
CP&L response dated July 14, 1982 m.
Inadequate access control 1)
IER261/82-16(item 82-16-05) 2)
CP&L response dated July 14, 1982 n.
Failure to implement drawing controls 1)
IER 261/82-20 (item 82-20-02) 2)
CP&L response dated August 20 and August 27, 1982 o.
Failure to maintain valve lineups 1)
IER261/82-20(item 82-20-03) 2)
CP&L response dated August 20 and August 27, 1982
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Failure to implement storage requirements 1).
IER 261/2-20 (item 82-20-09) 2)
CP&L response dated August 20 and August 27, 1982 q.
Failure to perform adequate review 1)
IER 261/82-27 (item 82-27-01) 2)
CP&L response dated October 8, 1982 9
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Use of superceded documents 1)
Use 261/82-27 (item 82-27-02) 2)
CP&L response dated October 8,1982 s.
Failure to control high radiation area 1)
IER261/82-31,(item 82-32-04) 2)
CP&L repsonse dsated February 24, 1983 and October 22, 1982 t.
Failure to implement procedures 1)
IER261/82-32(item 82-32-01) 2)
CP&L response dated October 29, 1982 u.
Failure to implement adequate corrective action 1)
IER261/82-32(item 82-32-02) 2)
CP&L response dated October 29, 1982 v.
Failure to establish and implement adequate surveillance
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IER261/82-32(item 82-32-03) 2)
CP&L response dated October 29, 1982 w.
Failure to perform audits 1)
IER261/82-33(items 82-32-01and02) 2)
CP&L response dated December 2,1982, December 31, 1982, and March 14, 1983 x.
Failure to distribute audit in time 1)
IER261/82-33(item 82-33-03) 2)
CP&L response dated December 2, 1982, December 31, 1982 and March 14, 1983 y.
Failure to respond to audit in 30 days 1)
IER 261/82-33 (item 82-33-04) 2)
CP&L response dated December 2,1982, December 31, 1982 March 14, 1983 l
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Unauthorized waste disposal l
1)
IER261/82-34(item 82-34-02) 2)
CP&L response dated November 30, 1982, and January 13, 1983 l
aa. High radiation area controls 1)
IER261/82-34(item 82-34-07) l 2)
CP&L response dated November 30, 1982 and January 13, 1983 L
ab. Failure to establish and implement equipment control procedures 1)
IER 82-37 (item 82-37-13) 2)
CP&L response dated January 21, 1983 ac. Failure to implement corrective actiori 1)
IER261/82-41(item 82-41-04) 2)
CP&L response dated February 3, 1983 ad. Failure to establish adequate calibration procedures 1)
IER261/82-42(item 82-42-01) 2)
CP&L response dated March 4, 1983 ae. Overpressure protection outside design basis 1)
IER261/82-42(item 82-42-02) 2)
CP&L response dated March 4, 1983 af.
Failure to implement procedures 1)
IER261/83-02(item 83-02-01) 2)
CP&L response dated March 25, 1983 af. Failure to take adequate corrective action 1)
IER261/83-02(item 83-02-03) 2)
CP&L response dated March 25, 1983 ah.
Improper radioactive waste shipment i
1)
IER261/83-03(item 83-08-01) 2)
CP&L response due June 23, 1983 ai. Failure to maintain procedures 1)
IER 261/83-12 (item 83-12-01) 2)
CP&L response due June 26, 1983 aj.
Failure to implement housekeeping 1)
IER261/83-12(item 83-12-02) 2)
CP&L response due June 26, 1983 ak.
Failure to provide adequate controls 1)
IER 261/83-12 (item 83-12 03) 2)
CP&L response due June 26, 1983 Severity Level IV violations at Brunswick facility since October 7,
- 1980, a.
Service Water Systems Secured Rending Safety Systems Inoperable 1)
IE Reports 324/81-02 and 325/81-02 2)
CP&L response dated April 14, 1981 b.
Late Licensee Report to NRC 1)
IE Reports 324/81-04 and 325/81-04 2)
CP&L response dated June 12, 1981 l.,
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c, Failure to Follow Procedures 1)
IE Reports 324/81-06 and 325/81-06 2)
CP&L response dated June 12, 1981 d.
Inadequate Procedure 1)
IE Reports 324/81-13 and 325/81-13 2)
CP&L response dated August 24, 1981 e.
Failure to Sample Radioactive Liquid Effluent Prior to Release 1)
IE Report 324/81-20 and 325/81-20 2)
CP&L response dated October 12, 1981 f.
Failure to Declare High Pressure Coolant Injection System Inoperable 1)
IE Reports 324/81-24 and 325/81-24 2)
CP&L response dated November 2,1981 g.
Failure to Implement Procedures and Failures to Take Required l
Primary Coolant Samples 1)
IE Reports 325/82-01 and 125/82-01 2)
CP&L response dated Mared il, 1982 l
h.
Failure to Implement Proc?dures 1)
IE Reports 324/82-05 and 325/82-05
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2)
CP&L response dated April 30 and May 10, 1982 i
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1.
Failure to Promptly Implement Corrective Action 1)
IE Reports 324/82-08 and 325/82-08 l
2)
CP&L response dated May 12, 1982 l
j.
Failure to Implement Procedures
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IE Reports 324/82-09 and 325/82-09 2)
CP&L response dated May 21, 1982 and February 25, 1983 k.
Failure to Take Adequate Corrective Action and Failure to Follow Procedures 1)
IE Reports 324/82-i) and 325/82-10 2)
CP&L response dated May 24, 1982 1.
Failure to Establish Procedures and Failure to Establish a Trend and Review Program 1)
IE Reports to 324/82-16 and 325/82-16 2)
CP&L responses dated August 3, August 26 and Novembe 29, 1982 m.
Failure to Establish Measures to Assure Prompt Corrective Action l
1)
IE Reports 324/82-20 and 325/82-20 2)
CP&L responses dated August 16 and September 24, 1982 n.
Failure to Establish New Pump Reference Data 1)
IE Reports 324/82-23 and 325/82-23 2)
CP&L responses dated August 18 and September 15, 1982 1
o.
Operating Procedures Not Conforming to ANSI Requirements and Not Adequately Maintained 1)
IE Reports 324/82-25 and 325/82-25 2)
CP&L response dated September 7, 1982 p.
Failure to Establish Measure to Assure Require Post Maintenance Testing is Performed 1)
IE Reports 324/82-26 and 325/82-26 2)
CP&L response dated September 13, 1982 q.
Improper Radioactive Waste Shipment 1)
IE Reports 324/82-42 and 325/82-42 2)
CP&L response dated January 5, 1983 r.
Failure to Take Adequate Corrective Action 1)
IE Reports 324/82-45 and 325/82-45 2)
CP&L responses dated March 18 and April 27, 1983 s.
Failure to Establish Adequate Written Procedures 1)
IE Reports 324/83-10 and 325/83-10 2)
CP&L response due June 23, 1983 Severity Level IV Violations at the Harris Facility since the estab-lishment of the present severity level system on October 7,1980.
a.
Failure to Identify and Correct Nonconfonning Conditions
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1)
IER Nos. 400/82-05 and 401/82-05 2),
CP&L responses dated April 21, 1982'and July 22, 1982 b.
Failure to Perform Required Audits 1)
IER Nos. 400/82-07 and 401/82-07 2)
CP&L response dated May 19, 1982 c.
Welding Record Discrepancies and Use of Non-certified Welding Inspectors 1)
IER Nos. 400/82-03 and 401/82-03 2)
CP&L response date June 1, 1982 d.
Failure to Follow Procedures 1)
IER Nos. 400/82-21 and 401/82-21 2)
CP&L responses dated August 5, 1982 and August 24, 1982 e.
Failure to Identify Nonconforming Condition 1)
IER Nos. 400/82-24 and 401/82-24 2)
CP&L responses dated September 2,1982 and December 17, 1982 f.
Failure to Follow Procedures 1)
IER Nos. 400/82-26 and 401/82-26 2)
CP&L response dated September 23, 1982
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g.
Failure to Identify Nonconforming Condition 1)
IER Nos. 400/82-28 and 401/82-28 2)
CP&L response dated October 7,1982 h.
Materials Nonconformance with Procurement Document 1)
IER Nos. 400/82-39 and 401/82-39 2)
CP&L response dated February 4,1983 1.
Failure to Document Material Upgrade 1)
IER Nos. 400/83-07 and 401/83-07 2)
CP&L response dated May 12, 1983 j.
Deficient QA Inspection 1)
IER Nos. 400/83-16 and 401/83-16 2)
CP&L response due June 19, 1983 INTERROGATORY 32 Describe in detail the corrective actions and management controls l
instituted by CP&L with respect to each instance of violation of NRC operating procedures, rules and regulations referred to in response to Interogatories 28-31.
RESPONSE
The Licensees planned corrective actions are detailed in response to the enforcement action. The NRC reviews the responses for adequacy and subsequently reviews the implementation of the Licensees corrective action on site.
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INTERR0GATORY 33 What are the bases for your responses to Interrogatories28-327 Identify all documents, testimony or oral statements by any person on which you rely in support of your position.
-RESPONSE The documentation listed in response to Interrogatories 30 and 31.
INTERROGATORY 34 Have any CP&L employees or contractors or subconstractor employees been warned, counseled, disciplined, transferred, demoted, penalized, suspended or terminated as a result of noncompliance with NRC operating and administrative procedures, rules or regulations at any licensed facility or for actions under any NRC license since January 1,1978?
RESPONSE
The NRC does not track this data. This information should be obtained from CP&L.
INTERR0GATORY 35 Has CP&L been the subject of requests for action, notices of proposed action, notices of violation, notices of proposed imposition of civil penalties, orders to show cause, proceedings to modify, suspend or revoke a license or to impose civil penalties pursuant to 10 C.F.R. Part 2, Subpart B, any other provisions of AEC or NRC statutes or regulations, or any civil or criminal proceeding in the courts of the United States or any State, before any agency of the United States or any State with resect to activities under AEC/NRC license? Describe in detaile each such instance, the violation or claim alleged, its date and place, the CP&L response including any evidence offered in answer, remission or mitiga-tion, the proceedings had thereon and the outcome.
RESPONSE
Yes, violations and civil penalties are addressed under previous Interrogatories (27-33)backtoOctober7,1980(dateofthepresent enforcementpolicy).
Known orders for modification of license the H.B. Robinson license:
a.
September 19, 1980 concerning environnental qualification b.
October 24, 1980 concerning environmental qualification c.
April 20, 1980 concerning primary system isolation valve testing d.
July 10, 1981 concerning TMI action items e.
March 14, 1983 concerning TMI action items Known orders for the Bruswick facility NRC orders issued are as follows:
a.
Commission orders confirming commitments for TMI related requirements were issued on July 10, 1981, March 14, 1983 and May 5, 1983.
b.
On December 22, 1982 an NRC Confirmatory Order was issued confirming the commitment to implement a long range Brunswick Station improvenent program submitted by CP&L.
INTERROGATORY 36 What are the bases for your responses to Interrogatories 34 and 35?
Identify all documents, testimony or oral statements by any person and legal requirements on which you rely in support of your position.
RESPONSE
The documentation listed in response to Interrogatories No. 34 and 35.
INTERROGATORY 37 Identify in detail any complaints made to the NRC regarding violations of NRC operating and administrative procedures, rules and regulations with respect to any activities under any AEC/NRC license
1 -
issued to CP&L.
For each complaint, set forth the name, address and telephone number of the persons complaining or involved in the manner complained of an explain fully the manner in which Applicant learned nf the complaint.
RESPONSE
See response to Interrogatory No. 38.
INTERROGATORY 38 Identify in detail any instances in which allegations have been made of pressure, intimidation, harrassment, encouragement, direct orders, suggestions, or inducement of any sort of employees of CP&L or its con-tractors or subcontractors intended to result in the violation of or noncompliance with NRC operating and administrative procedures, rules or regulations.
For each such instance, set forth the name, address or telephone numbers of the person (s) making the allegation or invovled in the NRE Matter alleged, describe fully any investigations made by CP&L or the NRC Staff,'and describe in detail any actions taken.
RESPONSE
The following is a list of all allegations in summary form, that were readily retrievable as a result of computer input. Much earlier allegations regarding CP&L operations may exist in filed documentation.
The retrieval of any earlier allegations (if any exist) would require the manual search of inspection reports and other correspondence that would encompass a massive effort and several hundred manhours.
The allegations as submitted in this list provide the case number, respective dates for opening and closing a case, the subject matter and a summary of actions taken or planned.
A key to acronym and abbreviations is included.
Case No:
83A002
Subject:
Improper Marking of Hi Rad Opened:
1-27-83 Areas - Possible Overexposure
Pending: Additional info to be provided by OI Case No:
83A003
Subject:
Failure to adhere to ALARA Opened:
1-27-83 Policy when working in contain Facility:
Brunswick ment Pending: Awaiting report of technical review Case No:
83A009
Subject:
Use of uncertified welders and Opened:
3-383 Falsification of records in the Facility:
Brunswick construction of spent fuel racks Pending: Status to be provided by 01 Case No.
83A010
Subject:
Failure to adhere to ALARA Policy Opened:
2-19-83 Facility:
Brunswick
1 UTL/FACLTY CASE DATES SUBJECT SUPMARY CPL 20047 Brunswick OPND: 78-10-04 Inadequate Guard Tra Inv. By I, S CLSD: 78-12-14 ing. Conduct Of 0 Of 4 Alleg Subt Guards, Etc.
1 Item On N/C CPL 23064 Reactor Vessels for Inv. by R4 Disclosed Brunswick OPND: 79-08-07 Units 1 & 2 Switched Unfounded CLSD: 79-10-24 During Const Closed W/0 Further Action CPL 20073 Use of Alcoholic Spec. Backshift Insp.
Brunswick OPND: 79-08-24 Beverages by Operat-(2) By Res Insp CLSD: 79-10-01 ing Staff Allege Unsubst No N/C CPL 2B001 Security Locks Inade-Allegations Found To Brunswick OPND: 80-01-02 quate and Personnel Be Outside NRC Juris-CLSD: 80-02-21 Safety Hazards diction Alleger Referred To D0L/0SHA CPL 2E012 Radiation Safety RI Conducted by F Brunswick OPND: 80-03-07 Practices 3 of 3 Alleg Subst CLSD: 80-09-30 6 items of NC CPL 2E017 Failure to Decontam-SI by FFMS Brunswick OPND: 03-28-80 inate Clothing Before 0 of 1 Allege Subst CLSD: 80-08-29 Leaving Site No Related N/C But Rpt Part of CP Pkg CPL 2E024 Inadequate HP SI Conducted by F Brunswick OPND: 80-04-25 Procedures During 27 items of NC CLSD: 80-09-22 Outage Identified in Related Spec Insp Civil Penalty Issued CPL 2E033 Deliberate Decision Inv Conducted by I&O Brunswick OPND: 80-05-19 To Continue 0 of 1 Allegation CLSD: 80-08-06 Unmonitored Release 0 Items of NC l
CPL 2E038 Security Records Case Closed W/0 l
Brunswick OPND: 80-06-17 Falsification Action Due To Lack l
Of Specifics l
No Response To Letters CPL 2E066 Possible Intentional Insp Conducted BY l
Brunswick OPND: 80-08-11 Improper Position Of SRI &INV.
CLSD: 81-01-05 Stop Check Valves 1 of 1 Allegations i
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UTL/FACLTY CASE DATES SUBJECT
SUMMARY
Subst I Item of NC CPL 2E067 Inadequate Correla-RI Conducted By F Brunswick OPND: 80-09-18 tion Between TLD &
Prior To Allegation CLSD: 80-11-08 Dosimeter Which Already Sub-stiated The Inade-The Inadequacies -
1 N/C CPL 2E091 Improper Radiation Inv by I&F Brunswick OPND: 80-11-13 Protection Practices 1 of I alleg subst CLSD: 82-02-12 Unreported Releases 5 violations CPL 2F022 Improper Radwaste Allegations referred Brunswick OPND: 81-03-31 Operation-Inop Valve to F and 0 for action CLSD: 81-05-12 and Leaking pipes deemed appropriate no action by EIS for lack of specifics CPL 2F043 Improper Security RI conducted by EIS Brunswick OPND: 81-10-15 Practices
& S 0 of 2 allega-CLSD: 81-01-21 tions subst I related violation ID CPL 2G001E Failure low level 3 Viol - Each SL3 Brunswick OPND: 82-01-05 Transmitter-6 days R2 Proposed 50K CLSD: 81-11-15 SL3-CP H2' Prop 240K CP 120K Issued CPL 2G007 Improper Security Insp conducted by Brunswick OPND: 82-12-12 Practices-Access R2 T 0 of 2 allega-CLSD: 82-01-17 Control tions subst No Violation identified CPL 2G035 Large numbers of RI by TIB Brunswick OPND: 82-05-1 contaminaed workers 1 of 4 Alleg Subst CLSD: 82-09-2 workers No Violations CPL 2G041 Potential Sabotage Inv Conducted by Brunswick OPND: 82-06-0 of Incore Nuclear Rpt Rvwd by 01 and CLSD: 82-09-2 Instruments PSS No Further R2 Action No Info of Sabotage CPL 2G055 Falsification of Pending:
Inv.
Brunswick OPND: 82-09-03 QC Inspection Conducted By Licensee CLSD:
Records Q.A. Task Force
UTL/FACLTY CASE DATES SUBJECT
SUMMARY
CPL 2G069 Falsification of Inv. Conducted Could Brunswick OPND: 82-10-04 Dosimetry Records Not Substantiate CLSD: 83-02-14 No N/C CPL 20046 Improper Installa-RI By RCES Harris OPND: 78-09-12 tion Of Rebar &
0 of 6 alleg Subst CLSD: 78-10-26 Achor Bold Pads No N/C CPL 20071 Inadequacies Resolu-RI by RCES Harris OPND: 79-08-16 tion of NCRs of Con-2 of 3 Alleg Substi CLSD: 70-01-28 crete placement 2 Items of N/C CPL 20096 Qualification of Lack of Detailed Info Harris OPND: 79-11-29 Craft Personnel A11eger Failed to CLSD: 80-06-23 Respond To Request for Details Closed W/0 Action CPL 2E016 Improper No Concerns Within Harris OPND: 80-03-21 Construction NRC Jurisdiction CLSD: 80-07-15 Practices Identified Insuffi-cient Detail To Proceed Allegation Referred OSHA CPL 2F017A Improper QC Practice Ri conducted By ET1 Harris OPND: 81-3-23 Qualification,&
+ 01 of 4 allega-CLSD: 81 2-28 NCR Handling tions subst 1 violation identi-fied 1 IFI JPL 2F017B Inadequate QC RI conducted by 0 Harrii OPND: 81-03-23 Inspector 0 of 2 allegation CLSD: 81-09-28 Qualifications subst 0 items of NC CPL 2F029 Improper Insp conducted by Harris OPND:
11-06-23 Welding of rebar 0 of 1 allegation CLSD:
11-11-06 in containment subst 2 violations b1dg identified CPL 2F035 Improper Const SI by E Harris OPND: 81-08-10 Materials and 8 of 12 allegations CLSD: 81-12-07 Practices subst i violation 1 URI CPL 2F051 Improper welding Inv by EIS Harris OPND: 8 11 inspection practices 1 of 1 Alleg Subst CLSD: 8_7-09-30 2 Viol - Both SL4
O,
UTL/FACLTY CASE DATES SUBJECT CUW4ARY CPL 2G010 Improper construc Invest by I Harris OPND: 82-02-02 Inspection Practices 1 of 2 allegations CLSD: 82-05-26 subst No violations CPL 2G025 Falsification of RI conducted by E Harris OPND: 82-03-31 Daily Aggregate 1 of I allegation CLSD: 82-06-16 Tests subst 1 insp followup 82-12-01 0 violations CPL 2G027A Forgery of QC Weld Inv by EIS Harris OPND: 82-04-14 Inspection Document 0 of 1 Alleg Subst CLSD: 82-09-22 No Violations 1 Related URI CPL 2G027B Discrepancies Id'd Pending: Review Harris OPND: 82-09-22 In Hanger Insp of Technical CLSD:
Records During Inv Follow Up CPL 2G067 Falsified Hydrotest Inv. Conducted Harris OPND: 82-09-30 Results Insufficient Detail CLSD: 82-11-19 To Proceed - Case Closed CPL 20011 Inadequate Radiation SI By FFMS - Results Robinson OPND: 78-02-17 Protection Program Not Indicated In CLSD: 78-03-01 (Several Examples)
File - FFMS Memo Indicate A11eger Satisfied W NRC
Response
CPL 2D004 Inadequate Training RI by FFMS Robinson OPND: 79-01-29 of HP Personnel None of 3 Alleg Subst CLSD: 79-03-22 No N/C - A11eger Filed i
Complaint W/ DOL: WHA CPL 2E015 Falsification of RI Conducted By S Robinson OPND: 80-03-12 Guard Training 0 of 1 Allegation l
CLSD: 80-06-23 Records 0 Item of liC l
CPL 2E074A Inadequate Security Insp Conducted By S Robinson OPND: 80-10-07 Practices 0 Of 7 Allegations CLSD: 80-11-21 Subst 0 Items of NC
.~
UTL/FACLTY CASE DATES SUBJECT
SUMMARY
CPL 2E074B Inadequate Control Backshift 1 sp By Robinson OPND: 80-10-07 Room Practices RONS On 2 Occasions CLSD: 80-11-08 CPS Personnel Sleep Not Identify Any Problem 1 Alleg Unsubst - NC CPL 2E078 Inadequate Operation RI Conducted By 0 Robinson OPND: 80-10-16 of Contair. ment Hatch And Telecon To CP&L CLSD: 80-11-08 By F 1 Of 2 A11ega-tions Subst. O Items of NC CPL 2E085A Inadequate HP SI Conducted by ETI Robinson OPND: 80-10-27 Practices 0 of 7 allegations CLSD: 81-07-20 (Inhalation) sbust 6 related violations identified CPL 2E085B Inadequate Security Inspection conducted Robinson OPND: 80-10-27 (Escort) Practices by 5 0 of 1 allega-CLSD: 81-03-26 tion subst No items of NC No related items CPL 2E086 Inadequate Neutron SI conducted by ETI Robinson OPND: 80-10-30 Exposure Records 0 of 3 allegations CLSD: 81-07-20 And Surveys subst 6 related violations identified CPL 2E097 Improper Contamina-SI Conducted by ETI Robinson OPND: 80-11-21 tion (Control)(Tools) 0 of 2 allegations CLSD: 81-07-21 and Suspected Drug subst 6 related Use identified CPL 2F010 HP Practices 1 Rvw Licensee Inv
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Robinson OPND: 81-02-26 RWPs Trng 1 of 12 allegations '
CLSD: 81-12-04 contam contr subst 4 referred to OSHA No violations CPL 2F054 Possible Materials Inv by EIS disclosed Robinson OPND: 81-12-30 False Statement NFS - Sev Lev 3 CLSD: 82-05-26 violation Referred to Enf CPL 2F054E Material False State-Issued as an Infrac-Robinson OPND: 82-05-19 ment - SL3 - No CP tion under old policy l
CLSD: 82-07-07
- Failure to jeport I:
j'N UTL/FACLTY CASE DATES SUBJECT
SUMMARY
CPL 2F055E Failure to evaluate Adequate Response Robinson OPND: 81-12-31 radiation risk to Acknowledged by ETI E
CLSD: 82-05-07 workerSL3-CP(5K)
DPC 20014 Inadequate Seismic SI by I, C & NRR Catewba OPND: 78-03-09 Analyses - Pipe Unresolved Item IDd EDS Nuclear CLSD: 78-03-23 Restraints
- NRR Determined 1
s Method Acceptable W/ Conditions r
M =,.
s 4
4 4
'A
Pending: Awaiting report of technical review Case No:
83A024
Subject:
Onsite marijuana use Opened:
4-5-83 Facility Brunswick Pending: Awating report of Licensee internal investigation Case No:
83A026
Subject:
Improper QA practices on safety Opended:
4-12-83 related installations Facility:
Brunswick Pending: Submitted for technical review Case No:
83A035
Subject:
Anon allegation on HP Matters
.0pened:
5-31-83 Facility:
Brunswick Pending: Submitted for technical review Case No:
83A041
Subject:
Anon caller: weed killer on f
Opened:
5-12-83 stainless pipi Facility:
Catawba Closed:
5-12-83 Pending: Closed after review of Technical Evaluation 6/7/83 Case No:
83A45
Subject:
Alleged over-exposure ALARA Opened:
6-10-83 violation Facility:
Robinson l
31 -
Closed:
6-10-83 Inspection conducted case closed subject submitted erroneous info to licensee Case Key: Acronyms for Sunnary Resolution of Allegations ANON Anonymous
=
INV Investigation
=
I, or IS Investigators of Investigative Staff
=
R4 Region IV
=
N/C or Noncompliance
=
RES Resident Inspector
=
D0L Department of Labor
=
OSHA Occupational Safety & Health Agency
=
=
SI Special Investigation or Inspection
=
i F
FFMS or Fuel Facilities & Materials Safety
=
Civil Penalty CP
=
SRI Senior Resident Inspector
=
NC Noncompliance
=
I&O Office of Investigator or Investigators & Operations
=
0 Operations
=
i EIS Enforcement & Investigators Staff
=
Security S
=
ID Identified
=
SL Severity Level
=
i l
p.e m
S
Techical Staff T
=
Office of Investigations OI
=
PSS Program Support Staff
=
Quality Assurance QA
=
RCES Reactor Construction & Engineering Support
=
Inspector Follow Up Item IFI
=
Unresolved Item URI
=
Reactor Operations & Nuclear Support RONS
=
ETI Engineering & Technical Inspection
=
MFS Material False Statement
=
Nuclear Reactor Regulation NRR
=
INTERR0GATORY 38 Identify in detail all documents reflecting disagreements, disputes or differences of opinion between employees of CP&L and their supervisors or CP&L management regarding compliance or sufficiency of compliance with NRC operating and administrative procedures, rules or regulations.
Include the subject, data, names of persons involved and resolution for each such instance.
RESPONSE
The NRC does not track this information.
INTERROGATORY 40 What evaluations of CP&L or its nuclear facilities have been carried out by the NRC Systematic Assessment of Licensee Performance Review Group?
Identify each such study or assessment and describe in detail its results and conclusions.
RESPONSE
Evaluations of the CPL nuclear facilities by the Systematic Assessment of Licensee Performance Review Group and the results and conclusions of such studies are contained in;
- IE Inspection Report 50-261/83-07 IE Inspection Report 50-261/82-17
- IE Inspection Report 50-324/83-09
- IE Inspection Report 50-325/83-09 IE Inspection Report 50-324/83-15 IE Inspection Report 50-325/82-15
- IE Inspection Report 50-400/82-14 IE Inspection Report 50-400/82-14 IE Inspection Report 50-401/82-14
- Not issued as of 6/9/83.
INTERROGATORY 41
. Describe in detail the basis for any rating of CP&L or any of its facilities by the NRC Systematic Assessment of Licensee Performance Review Group.
RESPONSE
The basis for rating a utility by the NRC SALP Review Group is contained in:
NRC Manual Chapter 0516, dated March 23, 1982 entitled -
Systematic Assessment of Licensee Performance" NRC Region II, Regional Office Instruction No. 1411 Revision 1, I
dated February 28, 1983 entitled
" Systematic Assessment of l
Licensee Perfonnance" l
Detailed bases on a specific rating are contained in the individual SALP reports when issued.
INTERROGATORY 42 I
What are the bases for your responses to Interrogatories 40 and 417 L
RESPONSE
The documentation listed in response to Interrogatories 40 and 41.
INTERROGATORY 43 Have any audits.or reviews conducted by NRC Staff or consultants to NRC Staff resulted in recomendation by one or more Staff members that sanctions be imposed upon CP&L for violation of or non-compliance with NRC operating and administrative procedures, rules or regulations where no sanctions were in the end imposed? If so, identify each such incident, describe in detail the violation or non-compliance, identify the staff member recomending imposition of sanctions, including that person's title and address, and the reason that no sanctions were imposed.
RESPONSE
No.
INTERROGATORY 44 What is the basis for your response to Interrogatory 43? Identify all documents, testimony or oral statements by any person on which you rely in support of your position.
RESPONSE
NRC Policy provides for the filing of differing professional opinions. There have been no differing professional opinions filed in Region II regarding sanctions to be imposed on CPL for noncompliance to NRC regulations.
IN addition, the principal inspectors involved with each plant have been questioned as to their agreement or disagreement on l
each sanction to be imposed. There has been only one differing I
professional opinion regarding a proposed sanction. The disagreement was solely on the amount of the civil penalty, not on whether to impose the civil penalty. This was resolved to the satisfaction of r
Headquarters and Region II.
l
INTERROGATORY 45 Do any NRC Staff members differ in any way from the Staff position on Contention la or Contention Ib in this proceeding?
RESPONSE
Not to our knowledge. Also see answer to Interrogatory No. 44.
INTERR0GATORY 46 If the answer to Interrogatory 45 is affirmative, identify each such NRC staff member, including that person's title, address and telephone number.
RESPONSE
Not applicable.
INTERROGATORY 47 If the answer to Interrogatory 45 is affirmative, identify in detail the diff'rences of each such identified staff period with the NRC Staff position and the bases for that difference.
RESPONSE
Not applicable.
INTERROGATORY 48 What are the bases for your responses to Interrogatories 45-47?
Identify all documents, testimony or oral statements by any person on which you rely in support of your pcsition.
RESPONSE
NRC Region II has no record of differing opinions regarding Contention la or Contention Ib, nor is the Staff aware of any disagreement communicated outside of the differing professional opinion method.
INTERROGATORY 49 Is the NRC Staff currently considering the imposition of any fines or sanctions on CP&L for violations of any NRC operating and administrative procedures, rules or regulations? If so, describe in detail the incident involved?
RESPONSE
Yes, the NRC has under consideration enforcement actions which may be. indicated as a result of recent inspections. At the present time the actions to be taken are not firmly established pending completion of NRC reviews. Any further discussions of the incidents may jeopardize further investigation and would have to await conclusion of the decision making process.
INTERR0GATORY 50 What is the basis for your response to Interrogatory 49? Identify any documents, testimony or oral statements by any person upon which you rely for support for your position.
RESPONSE
See response to Interrogatory No. 49.
INTERR0GATORY 51 Describe in detail how the procedures followed by the NRC Staff in conducting an investigation of alleged non-compliances.
RESPONSE
When the NRC Staff receives an allegation of non-compliances with NRC regulations, the allegation is referred to the appropriate technical personnel at the region for evaluation.
If the allegation is of potential immediate safety significance, the licensee may immediately be informed of the allegation in order to allow immediate corrective
actions, if the safety significance is such that the licensee need not be immediately informed, but inspection into the concern can not await the next scheduled inspection, the Region would conduct a special inspection into the allegation. All other allegations would be
~
inspected in the course of scheduled inspection and enforcement activities. All allegations are tracked with an assigned tracking number both within the Region and through an NRC-wide systems maintained by I&E. Evaluation of an allegation may also involve the Office of Investigations (01). OI is independent of the Office of Inspection and Enforcement and reports to the Commission. OI's function is to investigate allegations which may involve wrongdoing or which involve extensive interviewing on both technical and non-technical issues. When the inspections are completed and any 01 investigation input has been provided to the inspection enforcement staff, the staff, if the allegation is substantiated, will take appropriate enforcement action and will issue an inspection report which provides the basis for the enforcement action.
For those allegers who wish to be recontacted, i
contact is made by letter or by telephone. Cases, whether closed or not, remain on file indefinitely.
l INTERR0GATORY 52 What standards does the NRC Staff employ in determining which level of enforcement severity shall be assigned to each instance of violation or non-compliance?
RESPONSE
10 CFR Part 2, Appendix C entitled " General Policy and Procedure i
for NRC Enforcement Actions"
- .a. -,
INTERROGATORY 53 Describe in detail the basis for Region II determinations which result in the notification of Washington NRC officials of items of non-compliance or violation.
RESPONSE
Region II deteminations of which items of non-compliance or violations are submitted for concurrence to I&E Headquarters, is currently based on the " case law" concept supplemented by guidance memorandums issued by the Director of Enforcement, IE. All enforcement actions involving Severity Level III or higher violations must be submitted to IE Headquarters for detailed review. Level IV violations occuring after an enforcement conference are also forwarded to IE Headquarters for review and concurrence with planned Regional action.
Also an audit program by IE Headquarters is in effect to assure Regional uniformity.
INTERROGATORY 54 Is Region II currently under NRC internal investigation or review I
for failure to adequately conduct inspections or audits or to apply sufficiently stringent severity levels to non-compliances or violations?
RESPONSE
A formal request was made by a member of the public involved in the Catawba hearings to 0I to review the handling of allegations by Region II relative to welding allegations relative to Catawba. 01 transferred the request to OIA who is presently reviewing the matter in accordance with the request. No other Region II cases are being reviewed.
l
_. ~.... _ _ _,.. _. - -.
u.
INTERROGATORY 55 If the answer to Interrogatory 54 is affirmative, describe those investigations in detail and identify all documents, testimony or oral statements by any person upon which you rely.
RESPONSE
NRC Region II is not conducting the evaluation. The system that NRC employed in any matter of this type is to conduct separate and independent review by OIA.
Contention 3 INTERROGATORY 1 When do you maintain that Robinson Unit 2 will exceed Pressurized Thennal Shock (PTS) screening criteria based upon current operation, procedures and practices?
RESPONSE
The Staff has found acceptable the CP&L estimated dated of 1993 for HBR-2 to reach the PTS screening criteria under current conditions.
INTERROGATORY 2 Describe in detail any proposed changes to operation of Robinson 2 which are designed to extend the period before Robinson 2 would exceed PTS screening criteria.
RESPONSE
Many changes could effectively be used to extend the period of operation before HBR-2 reaches the PTS screening criteria. With respect l
i to the peak weld it has been estimated that a fast flux reduction factor ofabout9.2(asoftheend of 1981) would be adequate for the HBR-2 pressure vessel to reach the PTS screening criteria at 32 EFPYs. Flux reductions could be acconiplished with a variety of core modifications.
i l
- -, _ _ _. ~ _ _ _. _ _.
However, CP&L has not yet proposed the exact scheme for flux reduction.
The Staff is aware that CP&L is studying flux reduction to be effected with and after reload Cycle 10. The Staff will review and evaluate any flux reduction schenes when submitted.
INTERROGATORY 3 For each of the proposed changes identified in response to Interro-gatory 2, specify the reason that the change would extend the period before exceedance of PTS screening criteria.
RESPONSE
To reach the PTS screening criteria the peak weld fast neutron fluence at HBR-2 must reach the value of 19.5X10 n/cm, (E 1.0MeV).
Reduction in the fast flux will extend the period for fluence accumulation to the indicated level.
INTERR0GATORY 4 For each of the proposed changes identified in response to Interrogatory 2, specify the length of time which the change would add to the period before Robinson 2 would exceed PTS screening criteria.
RESPONSE
As mentioned in the response to Interrogatory 2 the staff does not yet have the specific flux reduction scheme or other measures from CP&L for review and evaluation, hence, no specific response can be given.
INTERR0GATORY 5 l
What are the bases for your response to Interrogatories I-5?
j-Identify all documents, testimony or oral statements by any person upon which you rely in support of your position.
~
RESPONSE
The Staff relies upon its own review and evaluation, the work of its consultants at BNL, sumittals, meetings and discussions with the Applicant and his consultants. The major documents for the PTS work for HBR-2 are:
1.
NRC Staff evaluation of pressurized thermal shock, (and Enclosure A)
SECY-82-465), November 23, 1982.
2.
Aronson, A. L., et al. " Evaluation of Methods for Reducing Pressure Vessel Fluence" BNL-NUREG-32876, BNL, March 1983.
3.
Meeting NRC, CP&L, January 25, 1983.
4.
Menorandum from L. Rubenstein to F. Schroeder, " Carolina Power &
Light's Estimate of 1993 for H. B. Robinson-2 Reaching the PTS Screening Criteria," March 15, 1983.
5.
Memorandum from L. Rubenstein to G. Lainas, "H. B. Robinson-2 Pressure Vessel Flux Reduction Plan," May 6,1983.
6.
Summary of Meetings with Carolina Power & Light Company on November 4, 8, and 12, 1982 concerning the pressurized thennal shock issue relating to the H. B. Robinson-2 plant, G. Requa, Project Manager, January 6, 1983.
7.
Letter from S. Varga (NRC) to E. Utley (CP&L), February 1,1983.
8.
Letter from Zimmerman (CP&L) to S. Varga (NRC), " Fast Neutron Fluence Calculations for Reactor Vessel," September 24, 1982.
9.
Letter from Zimmerman (CP&L) to S. Varga (NRC), " Fast Neutron Fluence Calculations for Reactor Vessel," September 24, 1982.
- 10. " Summary of Meeting with CP&L on January 25, 1982 Concerning the i
Pressurized Thermal Shock Issue Relating to the H. B. Robinson-2 l
Plant" G. Requal, Project Manager, February 11, 1983.
- 11. 'Chexal, B. et al.
"EPRI PTS R&D Efforts and Robinson-2 Plant Specific Analysis" Presentation to the NRC, November 12, 1982.
- 12. Trip Report: " Discussion of H. B. Robinson Fast Neutron Fluence Discrepancy to the Pressure Vessel, with W, CP&L and BNL", L. Lois, August 30, 1982.
- 13. Letter Report:
E. Utley (CP&L) to D. Eisenhut (NRC), " Thermal Shock to Reactor Pressure Vessels," January 25, 1982.
- 14. Letter Report:
L. Eury (CP&L) to T. Novak (NRC), " Pressurized Thermal Shock," May 4, 1982.
- 15. Memorandum from L. Lois to C. Berlinger, "H. B. Robinson, Fast Neutron Fluence to the Pressure Vessel," September 8,1982.
INTERR0GATORY 6 Which of the proposed changes identified in response to Interrogatory 2 have been approved by the NRC Staff?
RESPONSE
As mentioned in the respone to Interrogatory 2 above, the Staff has not yet approved any changes related to H. B. Robinson flux reduction.
INTERROGATORY 7 What is the basis for your response to Interrogatory 6? Identify all documents, testimony or oral statements by any person upon which you rely in support of your position.
RESPONSE
See the document list supplied in the response to Interrogatory 5 above.
INTERROGATORY 8 If all currently planned and approved changes in operation of Robinson 2 ar implemented, when do you maintain that Robinson 2 will exceed PTS screening criteria?
l
RESPONSE
l See response to Interrogatory 6 above.
l INTERROGATORY 9 What is the basis for. your response to Interrogatory 8? Identify all documents, testimony or oral statements by any person upon which you rely in support of your position.
4
RESPONSE
See response to Interrogatory 6 above.
INTERROGATORY 10 Has CP&L been issued a 10 CFR 50.54(f) letter with regard to PTS screening criteria or PTS at Robinson 2?
RESPONSE
Yes.- By letter dated August 21, 1981 CP&L was issued a 10 C.F.R. 50.54(f) letter concerning PTS for Robinson 2.
f INTERR0GATORY 11 Is the NRC Staff considering issuing a 10 CFR 50.54(f) letter to CP&L with regard to PTS?
RESPONSE
Notasofthisdate(July 5,1983).
INTERROGATORY 12 i
What is the basis for your response to Interrogatory 11?
Identify all documents, testimony, or oral statements by any person upon which you rely in support of your position?
REPONSE CP&L has provided responses to the NRC letter dated August 21, 1981.
' These responses along with responses of other Licensees who received the August 21, 1981 letter were reviewed by the Staff and contributed to the Staff's present. position concerning PTS. This position is provided in SECY 82-465. SECY 82-465 provides a list of references which related to this issue.
As a result of SECY 82-465 and a subsequent Comission Meeting of December 9,1982, the Comission directed the Staff (See Memorandum for W. J. Dircks from S. J. Chilk dated December 23,1983) to meet with CP&L to determine CP&L's plans for flux reduction programs for Robinson 2 and issue a 10 CFR 50.54(f) letter if appropriate following the meeting.
The Licensee has met with the Staff concernisg their plans of flux reduction programs for Robinson 2 and has provided substantial data which appear to support their view that such programs would prevent reaching the PTS screening criterion before the expiration of the operaing license for Robinson 2.
The material is under review by the Staff for a determination. Documents which support the above in addition to those identified above and referenced above are:
1.
Letter dated February 1,1983 to E. E. Utly, CP&L from S. A. Varga, NRC.
2.
Sumary of Meetings with CP&L on November 4, 8, and 12,1982 concerning PTS issue related to the Robinson 2 plant dated January 6, 1983.
3.
Sumary of Meeting with CP&L on January 25, 1983 concerning PTS issue relating to the Robinson 2 plant dated February 11, 1983.
4.
Letter dated February 9,1983 from E. E. Utly, CP&L to H. R. Denton, NRC.
j 5.
Letter dated February 24, 1983 frcm E. R. Zimerman, CP&L to H. R. Denton, NRC.
6.
SECY 83-79 dated February 25, 1983.
INTERROGATORY 14 What is your assessment of the fluence experienced to date by the welds and plates in the Robinson 2 pressure vessel and the rate of increase expected assuming that future fuel cycles to which CP&L has comitted to the NRC.
RESPONSE
The results of the most recent Staff estimate of the vessel peak 4
wall fluence vs EFPYs including the low leakage core now in operation can be found in Enclosure 1 of document No. 6 of the response to
~
Interrogatory 5 above. However, the projections shown do not represent NRC approved nor CP&L committed loading schemes.
INTERROGATORY 15 What is the basis for your response to Interrogatory 14? Identify all documents, testimony or oral statements by any person upon which you rely for support for your position.
RESPONSE
See Enclosure 1 of document No. 6 in the response to Interrogatory 5.
The calculations for the fluence estimate have been performed by BNL.
INTERR0GATORY 16 Using the fluence information set out in response to Interrogatory 14, what is your assessment of the RT in the Robinson 2 pressure vessel welds utilizing thT presently existing i
methodology outlined in Appendix E to Enclosure A of SECY 82-465, the expected future rates of increase, and the expected dates when the applicable proposed screening criteria [RT for circumferential weNI]OF 270 F for plates and axial welds and 300 F will be exceeded?
l
RESPONSE
Projected fluence values for several loading schemes are shown in of document No. 6,in the response to Interrogatory 5.
The corresponding values of a Best Estimate (BE) and of a Conservative Estimate (CONS) of the peripheral weld RT are also shown. However, NDT I
these projections do not correspond to any CP&L committed or NRC l
l l
l
. _ n_. m i
s.
4 approved leading scheme. See also the responses to Interrogatories 2 and 14 above.
INTERROGATORY 17 What are the. bases for your response to Interrogatory 16? Identify all documents, testimony or oral statements by any person upon which ycu rely for support of your position.
RESPONSE
See Enclosure 1 of document No. 6 in the response to Interrogatory 5.
The calculations for the fluence estimate have been performed by BNL.
INTERR0GATORY 18 Does the NRC Staff agree that the H. B.~ Robinson plant will not exceed the NRC Generic Screening Criteria until 1993.
RESPONSE
Yes. See response to Interrogatory 1 above.
INTERR0GATORY 19A t
Identify all letters, memoranda, notes of telephone conversations, minutes of meetings, correspondence, or other communications between CP&L, its contractors, suppliers or agents with the NRC Staff, its employees, or consultants with regard to PTS at the Robinson 2 facility.
RESPONSE
l See documents listed in response to Interrogatory 5 above.
j INTERR0GATORY 19B
l Identify all reports, memoranda, studies or other documents prepared by or on behalf of the Office for Analysis and Evaluation of Operational Data of the NRC relating to PTS.
RESPONSE
1.
Some general information was described in Appendix C of NUREG-0900,
~
Vol. 4, No. 4 and Appendix C of NUREG-0900-5.
2.
Reports issued in Power Reactor Events / Current Events - Power Reac-tors a.
Power Reactor Events, May-June 1982/Vol. 4, No. 4, " Inoperable Overpressure Protection System," pp.10-12.
(North Anna 1) b.
Current Events - Power Reactors, May-June 1978, "Overpressuri-zation," pp. 8-10.
(Robinson) c.
Current Events - Power Reactors, March-April 1978, " Loss of Non-Nuclear Instrumentation," pp.1-3.
(RanchoSeco)
INTERR0GATORY 20 Identify all memoranda or other correspondence between the Generic Issues Branch of the NRC to the Nuclear Reactor Regulation branch and all internal memoranda within the Generic Issues Branch relating to PTS.
RESPONSE
The Generic Issues Branch has not authored memoranda or other correspondence on the subject of PTS addressed to the Nuclear Reactor Regulation " branch". The same answer applies if we assume the petitioner actually meant to properly identify the Office of Nuclear Reactor Regulation, which is not a " branch".
The Genecic Issues Branch has not issued any internal memoranda on the subject of PTS.
- INTERROGATORY 21 Do any NRC Staff members differ'in any way from the Staff positions set forth in response to Interrogatories 1-18?
RESPONSE
Respondent is not aware of any NRC Staff members who differ in opinion as far as neutron fluence levels or neutron flux reduction measures are concerned.
INTERR0GATORY 22 If the answer to Interrogatory 21 is affirmative, identify each such NRC Staff member, including the person's title, address and telephone number.
RESPONSE
See response to Interrogatory 21 above.
INTERR0GATORY 23 If the answer to Interrogatory 21 if affirmative, identify in detail the differences of each such identified Staff member with the NRC Staff position and the bases for that difference.
RESPONSE
See response to Interratory 21 above.
l l
INTERROGATORY 24 Identify in detail all regulatory guides or other formal or informal guides, standards, rules of thumb or screening criteria employed by the Staff in reviewing the adequacy of proposed actions to reduce neutron flux in the reactor vessel or the safety margins in reactor neutron bombardment.
l i
~ ~ -.
' RESPONSE With respect to fast neutron flux reduction there do not exist any applicable regulatory guides, standards, or other rules.
INTERROGATORY 25 Excluding PTS and steam generator tube degradation, has CP&L or the Staff identified other major reactor components utilized at the Robinson 2 facility which have demonstrated a tendency to degrade with age?
' RESPONSE-All components have a normal degradation with age. Normal degradation is handled by surveillance testing and routine maintenance programs. Abnormal aging, accelerated by corrosion or irradiation, as observed in the steam generators and the pressure vessel has not been observed on other major reactor components utilized in Robinson 2.
INTERROGATORY 26 If the response to Interrogatory 25 if affirmative, identify each such component.
RESPONSE
Not Applicable.
INTERROGATORY 27 What are the bases for your responses to Interrogatories 25 and 26? Identify all documents, testimony or oral statements by any person upon which you rely for support for your position.
RESPONSE
Response to Interrogatory 25 is based on discussions with previous project manager, the resident I&E inspector and members of the NRC technical staff.
INTERR0GATORY 28
- For each component identified in response to Interrogatory 27, what is your best estimate of:
a) the useful life of the component b) when CP&L will be required to undertake replacement of the component; c) what the estimated costs of repair and/or replacement will be.
^
RESPONSE
Not applicable.
INTERROGATORY 29 What is the basis for your response to Interrogatory 28? Identify all documents, testimony or oral statements of any person upon which you rely.
RESPONSE
Not applicable.
INTERROGATORY 30 Which operating reactors utilize the Westinghouse Model 44 steam generators?
RESPONSE
H. B. Robinson 2 Indian Point 2 Indian Point 3 Turkey Point 4 Ginna 1 Point Beach 1 Point Beach 2 l
INTERROGATORY 31 How does the Model 44F steam generator differ in design from other Model 44 steam generators?
RESPONSE
To minimize the potential for several modes of steam generator tube degradation which have been identified to date, the Model 44F generators include the following improvements:
1.
Type 405 ferritic stainless steel quatrefoil tube support plate.
2.
Thermally treated Inconel 600 tubing and stress relief of the innermost eight rows of the tube bundle.
3.
Expansion of the tubes to the full depth of the tubesheet to eliminate crevices.
4.
A flow baffle plate above the tubesheet to direct lateral flow across the tubesheet surface and thus minimize the number of tubes exposed to sludge.
5.
An improved blowdown system to increase blowdown capacity.
l INTERROGATORY 32 Which operating reactors utilize the Westinghouse Model 44F steam l
generator?
l
RESPONSE
Turkey Point 3.
INTERROGATORY 33 What are the bases for your responses to Interrogatories 30-32?
Identify all documents, testimony or oral statements by any person upon which you rely for support of your position.
\\
\\ '
RESPONSE
USNRC Report NUREG-0886, " Steam Generator Tube Experience,"
published February 1982.
INTERR0GATORY 34 How many Westinghouse Model 44F steam generators have experienced significant degradation of tubes resulting in tube leaks?
RESPONSE
There have been no reports of tube leaks in Westinghouse Model 44F steam generators.
INTERR0GATORY 35 Identify each reactor utilizing Westinghouse Model 44F steam generators which has experienced tube leaks.
RESPONSE
There have been no reports of any reactors utilizing Westinghouse Model 44F steam generators having experienced tube leaks.
INTERR0GATORY 36 What data do you possess on the frequency and severity of tube leaks in reactors equipped with Westinghouse Model 44F steam generators? Identify the sources and bases for that data.
RESPONSE
There are no data indicating tube leaks in reactors equipped with Westinghouse Model 44F steam generators.
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INTERROGATORY 37 What are the bases for your responses to Interrogatories 34-36?
Identify all documents, testimony or oral statements by any person upon which you rely for support of your position.
RESPONSE
The docket file of Turkey Point 3, the only operating reactor with Westinghouse Model 44F steam generators, contains no information indicating steam generator tube leakage as of this date.
INTERROGATORY 38 How many tube ruptures have occurred at reactors employing Westinghouse Model 44F steam generators?
RESPONSE
There have been no tube ruptures reported at reactors employing Westinghouse Model 44F steam generators.
INTERROGATORY 39 At which reactors employing Westinghouse Model 44F steam generators, have:
a) steam generator tubes been plugged; b) steam generator tubes been sleeved; or, c) lower steam generator assemblies been replaced?
RESPONSE
There have been no reported replacements, sleeving or plugging of tubes employing Westinghouse Model 44F steam generators.
INTERR0GATORY 40 Identify any additional reactors employing Model 44F steam generators where the operators or owners anticipate:
a) plugging steam generator tubes;
b) sleeving steam generator tubes; c) replacing the lower steam generator assemblies.
RESPONSE
There are no additional reactors employing Westinghouse Model 44F steam generators where the operators or owners anticipate plugging, sleeving or replacement.
INTERR0GATORY 41
.What are the bases for your responses to Interrogatories 39-40?
Identify all documents, testimony or oral statements by any person upon which you rely for support for your position.
RESPONSE
There is only one operating reactor, Turkey Point 3, employing Westinghouse Model 44F steam generators and the docket file of this reactor does not show any instances of tube leaks, plugging, or anticipated sleeving or replacement.
INTERR0GATORY 45
" Describe in detail the bases for the numbe, of plugged tubes allowed at Robinson 2".
l i
RESPONSE
The bases for the maximum number of plugged tubes allowed at H. B.
Rcbinson Unit 2 (HBR-2), are found in the Exxon Nuclear Company (ENC) report XN-NF-82-18 "ECCS Plant Transient Analyses for HBR-2 Reactor Operating at Reduced Primary Temperature", dated March 1982. The licensee's program of reduced temperature, flow and power was proposed to allow up to 20% tube plugging and to improve the operating conditions j
l of the steam generators. Detailed transient and accident analyses were l
l
submitted in the above ENC document. The NRC safety evaluation concluded that the licensee's analyses provide an adequate safety margin for the events analyzed, and were acceptable subject to provision of certain confirmatory infonnation.
INTERROGATORY 48 "What are the bases for your responses to Interrogatories42-477 Identify all document, testimony or oral statement by any person upon which you rely in support of your position".
RESPONSE
With regard to Interrogatory 45, the major document in support of our position is listed in the response. The following additional documents were utilized in the NRC review:
o ENC report XN-NF-80-43 "ECCS and PTS Analyses for HBR-2 Reactor with 6%,10% and 15% Steam Generator Tube Plugging",
September 8,1980.
o ENC Report XN-NF-81-54 "LOCA ECCS analysis for HBR-2 Reactor For Revised Safety Injection Location", August 6, 1981.
o ENC Report XN-NF-79-42 " Review of Plant Transient Analysis for l
Positive Moderator Temperature Reactivity Feedback for HBR-2",
l June 22, 1979.
o ENC Report XN-76-54 "LOCA Analyses for HBR-2 Using WREM Based PWR ECCS Evaluation Model with Reduced LPSI Flow, Steam Generator Plugging and Increased Upper Head Temperature",
December 1976.
o Amendment 61 to HBR-2 Facility Operating License, November 1981 INTERR0GATORY 54 "Was the current derating required by the NRC7" l
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RESPONSE
The current derating was not required by the NRC, but was proposed by the licensee for the reasons given in the response to Interrogatory 45. Based on the favorable conclusions of the staff's safety evaluation, the NRC, by letter to Carolina Power & Light of July v.
23, 1982, issued Amendment 71 to the HBR-2 facility operating license, which authorized Cycle 9 operation at reduced power level and revised the HBR-2 technical specifications for operation at reduced Taverage..,
s At reduced Taverage condition, rated power is defined as 1955 MWt (85%'
of normal rated power).
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INTERROGATORY 58 Has AVT eliminated tube cracking, thinning, and denting?
RESPONSE
l Thinning and cracking of steam generator tubes resulting from t-l phosphate secondary water chemistry ocntrol have been reduced or eliminated by implementing AVT. However, denting was first discovered '
after' implementing AVT. Since 1977 there has been a significant "
decrease in denting initiation and progression while using AVT.
INTERR0GATORY 60 Has the employment of AVT treatment resulted in the occurrence of other problems in steam generator tubes?
RESPONSE
Yes..
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INTERR0GATORY 61 If the response to Interrogatory 60 is affirmative, identify in detail what those problems are, where they have occurred, and the extent to which they have resulted in the need to plug or sleeve steam generator tubes.
RESPONSE
AVT provides no buffering capacity to mitigate the effect of impurities in the cooling water. The use of AVT has resulted in denting of steam generator tubes at region s between the tubes and the tube supports.
In addition to denting, cracking has cccurred at the inner-row tube apex in one plant as a consequence of excessive tube denting. Two plants have had pitting of the Inconel 600 tubes near the tubesheet. Tube sleeving or plugging have been used as repair techniques for most degraded tubes.
INTERR0GATORY 62 What are the bases for your responses to Interrogatories 58-60?
Identify all documents, testimony, or oral statements by any person upon which you rely in support of your position.
RESPONSE
The responses to Interrogatories 58-60 are based on direct experience in the area of steam generator corrosion. A large number of papers and reports in the open literature which are too numerous to document provide supporting information. For example:
1.
J. A. Armantano and V. P. Murphy, " Standby Protection of High Pressure Boilers," Proceedings of the 25th Annual Water Conference of the Engineers' Society of Western Pennsylvania, Pittsburgh, PA, September 28-30, 1964, pp. 111-124.
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2.
H. H. Uhlig, Corrosion Handbook, J. Wiley & Son, Inc. N.Y., 1971, pp. 98-99.
3.
S. L. Goodstine and J.'J. Kurpen, " Corrosion and Corrosion Product Control in the Utility Boiler - Turbine Cycle, combustion, May 1973.
4.
F. Gabrielli and J. J. Kurpen, " Secondary Cycle Chemistry Control for a Pressurized Water Reactor," Proceedings of the American Power Conference, 34 (1972).
5.
M. C. Bloom. "A Survey of Steel Corrosion Mechanisms Pertinent to Steam Power Generation," Proceedings of the 21st Annual Water Conference of the Engineers' Society of Western Pennsylvania, Pittsburgh, PA, October 24-26, 1960, pp. 1-21.
6.
NRC Report NUREG-0886, " Steam Generator Tube Experience".
7.
E. C. Potter and G. M. W. Mann, "The Fast Linear Growth of Magnetite on Mild Steel in High-Temperature Aqueous Condition,"
British Corrosion Jornal, 1, pg. 26 (1965).
~8.. - EPRI Report NP-2541, "PWR Steam Side Chemistry Follow Program, Research Project RP 699-1 Final Report," Electric Power Research Institute, August 1982.
'9.
G. E. Von Nieda, G. Economy, and M. J. Wootten, " Denting in Nuclear Steam Generators--Laboratory Evaluation of Carbon Steel Corrosion Under Heat Transfer Conditions," presented at the NACE Annual Meeting, March 1980.
10.
G. Economy, W. M. Connor, and G. E. Von Nieda, " Laboratory Studies of the Effect of Chemistry on Denting in Nuclear Steam Generators,"
presented at the NACE Annual Meeting, March 1982.
- 11. EPRI Report NP-xxxx, " Rationale for Chemical Control of Feed and Boiler Water, Research-Project RP1171-1 Final Report," Electric Power Research Institute,1982.
- 12. EPRI Report NP-516, "PWR Secondary Water Chemistry Study, Research Project RP401-1 Interim Report," Electric Power Research Institute, February 1977.
13.
W. L. Pearl and S. G. Sawochka, "PWR Secondary Water Chemistry Study - Progress Report," Proceedings of the American Power l
Conference, 39,840(1977).
l l
14.
S. G. Sawochka and W. L. Pearl, "PWR Secondary Water Chemistry l
Study - Progress Report," Proceedings of the American Power Conference, 40, 918 (1968).
Systems, Research Project RP 404-1 Topical Report," Electric Power Research Institute, December 1981.
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INTERROGATORY 64 Do you restimate that the design changes outlined in the FSGRR will eliminate tube leaks?
RESPONSE
Design changes, water chemistry changes, and other changes outlined in the FSGRR, particularly the thermal treatment of the Inconel 600 tubes, should greatly minimize the potential for tube leaks in the replacement Model 44F steam generator tubes. The use of stainless steel support plates should eliminate support plate corrosion buildup as a cause for denting the tubes at that location, and subsequent leakage due to excessive denting of the tubes.
INTERR0GATORY 65 If the response to Interrogatory 64 is affirmative, what is the ba.iis for your respone? Identify all documents, testimony or oral statements by any person upon which you rely for support of your position.
RESPONSE
There is no record of tube leaks in operating Westinghouse Model 44F steam generators indicating an improvement over past performance of
(
l operating Model 44 steam generators.
l The technical literature also indicates that thermally treated Inconel 600 is not susceptible to intergranular stress corrosion cracking and intergranular attack which are two of the principal mechanisms of tube degradation that have been experienced in the past.
l
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INTERROGATORY 66 If the response to Interrogatory 64 is negative, describe in detail your estimates of the number of tubes which will leak during each year of operation from 1984 until decommissioning of Robinson 2.
RESPONSE
There is no data base on which to estimate the number of tubes that will leak every year in the Robinson 2 nuclear plant replacement steam generators.
INTERROGATORY 67 What is the basis for the response to Interrogatory 66? Identify all documents, testimony or oral statements by any person upon which you rely in support of your position.
RESPONSE
The only operating plant with Westinghouse Model 44F steam generators has not reported any steam generator tube leaks as of this date.
INTERROGATORY 74 Is the weld which will be made to rejoin the lower steam generator assemply and the upper steam generator assembly the same weld as the girth weld which has cracked at Indian Point 37 [See Report to Congress on Abnormal Occurrences, April-June 1982, NUREG-0900, Vol. 5, No. 2, pp. 18-19.J
RESPONSE
Yes, as to location. However, the welds at Robinson will be made in the field using manual shielded metal arc welding. The Indian Point 3 welds were shop welded, with the majority of the volume of weld metal 1
being deposited by the submerged arc welding process. Accordingly, these welds are not the same as those in the Indian Point 3 steam
generators. There have been no failures in welds which were field fabricated.
INTERROGATORY 75 If the response to Interrogatory 74 is affirmative, is there any basis for asserting that the same kind of crack is incredible in the repaired steam generators at Robinson 2?
RESPONSE
With (1) the improvements in water chemistry, (2) the efforts to improveotherplantequipment,and(3)thatthechloridecontentofits condenser cooling water is low compared to that of Indian Point 3, we d
would describe the finding of cracks in Robinson Unit 2 steam generators as occurred in Indian Point 3 as very unlikely, Although the mechanism (corrosion fatigue) was identified, the actual cause is not known, and is regarded as inconclusive based on the available data.
INTERROGATORY 76 If the response to Interrogatory 75 is negative, what is the likelihood of such a crack occurring in the weld at Robinson expressed in probabilistic terms?
RESPONSE
Because there has been only one event (thr four steam generctors at Indian Point 3), there is not enough data to perform a meaningful probabilistic evaluation.
INTERR0GATORY 77 What are the bases for your responses to Interrogatories 74-76?
Identify all documents, testimony or oral statements upon which you rely in support of your position.
RESPONSE
1.
Safety Evaluation by the Office of Nuclear Reactor Regulation related to Amendment 47 to Facility Operating License No. OPR-64, May 7, 1983.
2.
Lucius Pitkin Inc. Technical Report No. 7164.
3.
Brookhaven National Laboratories Report, NUREG/CR-3281, BNL/NUREG-51670.
INTERROGATORY 7,8_
What studies are you aware of which have been conducted by CP&L, Westinghouse, the NRC or any other entity which examine the likelihood of steam generator degradation and tube leakt in circumstances involving Model 44F steam generators?
RESPONSE
One of the most significant changes made in the Model 44F steam generators which will directly impact on steam generator tube degradation and leakage is the thermal treatment of the Inconel 600 tubes. There is extensive literature coverage of the fact that thermally treated Inconel 600 has improved resistance to intergranular l
corrosion over mill annealled Inconel 600. Three specific references, 1
most appropriate to the Westinghouse thermal treatment are as follows:
1.
G. P. Airey, " Optimization of Metallurgical Variables to Improve the Stress Corrosion Resistance of Inconel 600".
EPRI Report NP-1354, March 1980.
2.
G. P. Airey, A. R. Vaia, "A Caustic SCC Evaluation of Thema11y Treated Inconel Alloy 600 Steam Generator Tubing," Presented at WICON 82 Symposium, Houston, Texas.
3.
G. P. Airey, " Carbide Dissolution and Precipitation Kinetics of l
Inconel 600".
EPRI Report NP-2093, October, 1981.
l l
9 I
63 -
INTERROGATORY 79 Identify all reports, memoranda, studies or other documents produced by or on behalf of the Office of Analysis and Evaluation of Operational Data relating to steam generator tube degradation in Westinghouse Model 44 steam generators.
RESPONSE
1.
Reports issued in the NUREG-0900 series l
a.
A general update of steam generator tube experience for all 1
PWR vendors as of November 1981 was described in NUREG-0886 issues in February 1982.
b.
A sumary of NUREG-0886 and a description of experience from November 1981 through August 1982 described in Appendix B (A0 76-11) of NUREG-0900, Vol. 5, No. 2.
c.
Steam Generator Tube Failure (Point Beach) NUREG 75/090 (A0 75-1).
d.
Steam Generator Tube Rupture at RE Ginna Nuclear Power Plant NUREG-0900, vol. 5, No. 1 (A0 82-4).
2.
Reports issued in Power Reactor Events / Current Events - Power Reactors a.
Westinghouse NSSS:
Power Reactor Events, January-February 1982/Vol. 4, No. 2 " Steam Generator Tube Rupture," pp.1-7.
(Ginna)
Power Reactor Events, Vol. 2, No. 2/ March 1980, " Steam Generator Problems," pp. 3-5.
(PointBeach1)
Current Events - Power Reactors, August-September 1975,
" Steam Generator Tube Leak," pp. 2-3.
(PointBeach2)
Current Events - Power Reactors, March 1975, " Steam l-Generator Tube Failure," p. 7.
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Current Events - Power Reactors, August 1974, " Tube Degradation in Steam Generators," p. 2.
(Robinson 2and Slurry 2)
INTERROGATORY 80 Identify all memoranda or other correspondence from the Generic Issues Branch of the NRC to the Nuclear Reactor Regulation Branch regarding tube degradation in Westinghouse Model 44 steam generators.
RESPONSE
The Generic Issues Branch has no memoranda regarding this subject.
INTERROGATORY 81 Identify all internal memoranda of the Generic Issues Branch of the NRC relating to steam generator tube degradation in the Westinghouse Model 44 steam generators.
RESPONSE
This interrogatory is the same as number 80.
INTERROGATORY 82 Do an NRC staff members differ in any way from the staff positions set forth in response to Interrogatories 30-81 relating to tube degradation in Westinghouse Model 44F steam generators?
RESPONSE
The Generic Issues Branch is not involved with the issues associated with Interrogatories 30-81, and has no position on this matter. Consequently, we have no response for Interrogatories 83 and 84.
e o INTERROGATORY 85 Identify in detail all regulatory guides or other informa? or fonnal guides, standards, rules of thumb or screening criteria employed by the staff in reviewing the adequacy of steam generator desig1 and performance.
i
RESPONSE
The following are some of the criteria used by the staff in reviewing steam generator design and performance:
o Regulatory Guide 1.83 " Inservice' Inspection of PWR Steam Generator Tubes";
o Regulatory guide 1.21 " Bases for Plugging Degraded PWR Steam Generator Tubes";
o SRP Section 5.4.2.1 " Steam Generator Materials";
o BTP CMEB MTEB 5-3 " Monitoring of Secondary Side Water Chemistry in PWR Steam Generators";
o Plant Technical Specifications.
INTERR0GATORY 86 Has the NRC staff published or is it preparing any reports on steam generators subsequent to the " Steam Generator Status Report" of i-February 1982?
RESPONSE
l The NRC staff has prepared a draft report to be designated NUREG-08441, "NRC Integrated Program for the Resolution of Unresolved Safety Issues (USI's) A-3, A-4, and A-5 regarding Steam Generator Tube j
Integrity." The only other generic report on steam generators is an update of the February 1982 NUREG-0886 " Steam Generator Tube Experience" which the staff is planning to prepare.
t INTERROGATORY 87 If the answer to Interrogatory 86 is affirmative, identify each such document or draft document.
RESPONSE
The draft report is designated an NUREG-08441 as identified in response 86. This draft is currently under intornal staff review.
There are presently no drafts in existence for updating the February 1982 NUREG-0886 " Steam Generator Tube Experience."
Contention 8 INTERROGATORY 8 How will the design of the SGLA vault differ from that employed by Florida Power & Light company at Turkey Point?
RESPONSE
The NRC staff cannot respond to this interrogatory since, except for the brief description in the FSAR, CP&L has not yet submitted the SGLA vault design to the NRC.
INTERROGATORY 9 l
l How will the design of the SGLA vault differ from that employed by Virginia Electric Power company at Surrey?
j
RESPONSE
l Same response as response 8.
i l
INTERR0GATORY 10 What is the basis for your respor.ses to Interrogatories 8 and 9.
l Identify all documents, testimony or oral statements by any person on which you rely for support of your position.
i
,. RESPONSE No SGLA vault design has been submitted or received by the NRC staff as of June 27, 1983.
INTERROGATORY 11-What is the seismic design basis for the SGLA vault?
RESPONSE
Same response as 8.
INTERROGATORY 12 For the tectonic region in which the Robinson facility is located, what is the maximum historical earthquake?
RESPONSE
In the Staff review of the updated FSAR and AEC Safety Analysis Reports, we found the only reference significant to the interrogatory question contained in Appendix D to the February 27, 1967 AEC Safety Analysis Report.
In Appendix D the Staff seismology consultant, the U.S. Coast and Geodetic Survey stated:
"Our estimate, based on the seismic history of the site, the adjacent seismir, areas near Summerville and to the west of the site, and the geology of the site, is that during the lifetime of the facility, we believe that an MM intensity VII l
l earthquake with accelerations of 0.2g (on dense underlying stratum), might occur and should be considered as the maximum l
potential earthquake."
l l
INTERROGATORY 13 What would be the effect on the SGLA vault of the nearsite occurrence of an earthquake of Modified Mercalli Intensity X and Magnitude 7?
i 1
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RESPONSE
Same response as 8.
INTERROGATORY 52 In evaluating the safety of disposal of the SGLAs, what standard will the NRC Staff employ?
RESPONSE
The staff will use the requirements listed in 10 CFR 20.1(c),
20.207, 20.301, 20.302, 20.311, 10 CFR Part 71 as well as the recommendations listed in Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As low As Is Reasonably Achievable," as the principal standards. Other applicable Federal and State regulations will be used as appropriate.
INTERROGATORY 53 Identify any studies, reports, or other documents upon which the NRC Staff will rely ;in making its determinations and reaching its conclusions regarding the safety of the proposed method for disposing of the SGLAs.
RESPONSE
(1) 10 CFR Part 20 (2) 10 CFR Part 71 (3) Regulatory Guide 8.8 (4) NUREG/CR-0199 (5) NUREG/CR-1595
INTERROGATORY 54 Do any NRC Staff members differ in any way from the Staff position on contention 8 in this proceeding?
RESPONSE
There are no staff members known whose positions regarding Contention 8 differ from the staff positions presented regarding radiation protection.
INTERR0GATORY 55 If the answer to Interrogatory 54 is affirmative, identify each such NRC staff member, including the person's title, address and telephone number.
RESPONSE
See response to Interrogatory No. 54.
INTERR0GATORY 56 If the answer to Interrogatory 54 is affirmative, identify in detail the differences of each such identified Staff member with the NRC Staff position and the bases for that difference.
RESPONSE
See response to Interrogatory No. 54.
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INTERROGATORY 57 l
What are the bases for your responses to Interrogatories 54-56?
Identify any documents, testimony or oral statements by any person upon which you rely for support for your response.
RESPONSE
Communication with NRC staff members.
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o INTERROGATORY 58 Identify any reports, memoranda, draft reports, studies, comments or other documents prepared by or on behalf of the Office for Analysis and Evaluation of Operation Data (AEOD regarding the disposal of SGLAs at Robinson, Unit 2, or any other reactor, including, but not limited to, noterial related to the design and construction of long-term storage vaults for the SGLAs or similar large contaminated components removed from reactor buildings.
RESPONSE
We have not issued any reports on this subject.
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PROFESSIONAL QUALIFICATION OF PAUL R. BEMIS
- 3.. USNRC A.
Section.Ch,1,ef - As hief Reactor Projegts Section, supervises the implementation of a program for the routine and reactive inspections of assigned power and research reactors during all phases ~of construc-tions,. testing, operation, or decommissioning to assure the safety of NRC licensed facilities and activities, compliance with,NRC
, requirements, and to enforce the provisions of NRC permits, licenses, rules, regulations, orders and other directives pertinent to the
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protection of the public and' safety and to the common defense and security.
B.
Technical Ass,istant - Assists the Director in establishing policies and i
guidance govering the mission of the Region 11 Division of Engineering
, and. Technical Porgrams. Conducts and/or supervises assigned special projects, inspections, safety analyses or investigations.
Provides the Director with appraisals of and recommendations for improving the effectiveness and efficiency of Regional inspection and licensing programs.
Services in an advisory capacity ta the Director on tecnnical, policy, and administrative matters coming to the Director's office for resolution in the areas of operator licensing, Health physics and Radiation Protection, management programs, and all facets of engineering.
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2 Senior pesident Inspector - In charge of the onsite inspection and C.
coordination of regional inspection at a large two unit facility with one unit in the prestartup thru commercial phases and the other unit in the construction and preop phases.
, Management programs Team - Assist a team of inspectors in looking at D.
all facets of management programs to include:
operations, surveil-lance.. maintenance, Quality Assurance / Quality Control, training, procedures, procurement, and regulatory 2dherence.
Operator _Jicens'ing Examiner /Re,ac,t,or Engi,n_ee,r - Develope and administers E.
~
examinations to reactor operators and Senior reacter operators at PWR's, BWR's, and research reactors.
Served on numerous task forces in licensing po'st TMI.
Nuclear Enoineer/ Operating Reactor Technol,ogy Specialist - Developed F.
and implemented a " modified" SRO program for future resident Developed prograos for the technical staf f at NRC in the inspectors.
areas of systems, security, Radwaste, Health Physics, and Statistics (PRA).
Served as technical assistant to the Korean and Tiwan Governments in the area of operations and Heatth Physics.
Served as operations " specialist" for the cercission during the TMI incident.
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Director of.Trainin,g - At Electrical Utility Developed training programs in l
2.
all areas of a nuclear plant of a utility.
Implemented the programs for the' cold license and first. hot license classes.
Developed program for engineers to take prior to P. E. exae. Developed a structured, self paced program for 4
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A.0's, CRO's, and SF at a fossil power plant, Received SRO license certification.
3.
Manager. Compliance Assistance and Technical' services (Private Irdustry)
Supplied all phases of cpepliance assistance, training, operator licensing exams, specialized operational programs and Rad waste assistance to the nuclear industry.
4.
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Health Physicis't - In charge of the Health Physics program at the
' Walter Reed Institute of Research, ' Armed Forces Institute of Pathology, and Walter Reed Hospital.
B.
Shift Supervisor and Healt.h Ph,ysic,ig - Functioned as shift supervisor and acting operations manager at a land based nuclear power plants, j
. Served as Health Physicist supervisor at one power plant.
l C.
Stud,e,n,t - Attended one year academic program in operations with specialty in Health Physics. Graduate #1 in class.
l D.
Numerous positions in the army to include the Security Agency, and l
taught at a survey school.
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Unoergraduate Majors and Coonates:
Mat,h, Physics, Chemistry, Computer Science, and Accounting a
Graduate Studies:
Nuclear Engineering, Statistics, Busi,ress Administration Numerous Technical ar.d Business short courses
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, Memberships American Nuclear Society, Health Physics Society, Honorary Plath and Physics Societies.
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UN11ED 51 A7 ES Of AMERICA NUEL EAR REGULA10RY COMMISSION BEFORE THE ATOMIC 5AFETY AND LICENSING BDARD A
RICHARD JOHN SERBU PROFESSIONAL QUALIFICATIONS health physicist with the Radiation 1 am presently assigned as aProtection Section of the Radiole;: cal Assessme Integration, Of fice of Nuclear Reactor Regulation, U. S. Nuclear Ps gulatory e
Commission.
I graduated from the State University College of New York at Potsdam w Bachelor of Arts Degree,in Chemistry.
the field of ' radiation protection / health physics in association with nuclear power reactors since June 1973.
From June 1973 to April 1980, I held positions as Project Engineer Dosimetr Health Physics; Manager, Radiological Monitoring; Project Engineer, Radiolo Training; Radiological Controls Supervisor; and Instructor, Chemistry an logical controls at Knolls Atomic Power Laboratory. included devel This programs, operational health physics /ALARA programs, a tection; familiarity with reactor systems; radiation protection aspects of reactor startup; radiation protection for maintenance and refueling / overhaul; chemistry l
Since April of control programs; and compliance with established requireme In this capacity I am responsible for the review and evaluation of radiation pro-tection/ALARA (As low As Reasonably Achievable) as practices which are employed by nuclear reactor licensees and license ap in meeting the standards for_ protection against radiation of 10 CFR Part 20.
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PAUL E. NORIAN PROFESSIONAL QUALIFICATIONS I am a Section Leader in the Generic. Issues Branch, Division of Safety Technology.
I have held this position since 1980 and am responsible for supervising staff activities related to the techn'ical resolution of various Unresolved Safety) Issues.These activities include the development of task action plans (TAP which describe the actions to be followed to resolve each issue, the performance of various technical studies described in the TAP, and development of the proposed regulatory position and value-impact analyses for resolution of these issues.
I graduated from Lehigh University in June 1955 with a Bachelor of Science Degree in Engineering Physics.
I also attended Drexel Institute of Technology, Catholic University of America, and the University of Maryland where I have taken various graduate courses in mathematics, physics, and electrical engineering.-
In July 1955, I began work as a' physicist with the duPont Company at the Savannah River Plant in Aiken, South Carolina.
From that time until March 1962, I worked in the Works Technical Department on operational physics problems associated with the heavy water production reactors of Savannah River. This work included such assignments as the devlopment of monitoring systems, performance of physics calculations required in reactor operation and in the development of new fuel elements, the review of operating procedures, and the analysis of various operating problems.
In March 1962 f
I was transferred to the duPont Company's Chestnut Run Laboratories in Wilmington, Delaware, and worked for its Film Department on the development of industrial applications for plastic films.
In December 1963, I accepted a position with the Division of Reactor i
- Licensing of the U. S. Atomic Energy Commission, and was a project leader in the construction permit review of Consolidated Edison's Indian Point No. 2 reactor and Wisconsin-Michigan's Point, Beach No. I reactor.
I was assigned as a nuclear engineer in the Systems Performance Branch of the Division of Reactor Standards in March 1967.. Ny responsibilities included analyzing and evaluating the performance of engineered safety systems and performing computer calculations for the evaluation of contairenent response and loss-of-c coolant accidents.
In March 1971, I participated in the Regulatory Task
. Force reappraisal of emergency core cooling systems for light water reactors. My main responsibility for the task force was the review of computer codes.and input assumptions for LOCA analyses.
In May 1973, I was assigned to the: Core Performance Branch in the Directorate of Licensing.
I served as Section Leader in the. Thermal Hydraulics Section and supervised s
the review of portions of reactor vendoF model changes to confora with the new requirements for LOCA models specified in Appendix K to 10 CFR Part 50.
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I was responsible for supervising the review of reactor vendor transient and LOCA analysis methods, the improvement of NRC analysis methods used in related accident analyses, and the performance of staff audit calculations for transients and LOCAs.
From June through December 1979, I was assigned to the Bulletins and Orders Task Force as a member of the Analysis Group, I served as Alternate Group Leader and coordinated the reviews of small break loss-of-coolant accidents (LOCA) and transient analyses submitted by the vendor owner's groups since the Three Mile Island accident.
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7 PROFESSIONAL QUALIFICATIONS Bernard Mann Reactor Systems Branch Division of Systems Integration Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission t
t-4 I am employed as a Nuclear Engineer with the Reactor Systems Branch, c
Division of Systems Integration, Office of Nuclear Reactor Regulation, O. S.. Nuclear Regulatory Commission, Washington, D.C.
My duties include evaluation of the design and safety analysis of reactor systems of nuclear power plants with respect to nuclear saf3ty. As part of my duties, I have been responsible for reviewing the safety analyses for steam generator repair for several facilities, including H. B. Robinson
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Unit 2.
1 I have been associated with nuclea.r energy licensing, design, systems analysis, project and test engineering.
From 1955 to 1960 I was employed by the Westinghouse Electric Corporation, Bettis Atomic Power Laboratory, l
where I performed systems design, analysis and process engineering work l
on pressurized water systems for naval reactors.
From 1960 to 1968 I j
was a senior engineer with Aerojet-General Corporation, performing project, l-systems and test engineering work connected with space nuclear power pro-grams.
From 1968 to 1969 I was employed by Battelle-Northwest on the Fast i
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Flux Test Facility (FFTF) program as resident engineer in their Atomics International Office.
From 1970 to 1972 I was a senior engineer with C. F. Braun & Co., where I performed systems design work on nuclear power and process projects, including the fast breedt.r reactor.
From 1972 to 1977 I was employed by the Atomic Energy Commission (subsequently NRC) in the Auxiliary and Power Conversion Systems Branch and Effluent Treatment Systems Branch.
From 1977 to 1980 I was a Nuclear Engineer with Energy Research and Development Administration (subsequently Department of Energy) in the Division of Nuclear Research and Application and subsequently in the Division of Nuclear Waste Management.
In 1980 I rejoined NRC as a senior systems engineer with the Auxiliary Systems Branch.
In 1982 I commenced working for the Reactor Systems Branch.
I attended the University of Louisville where I received a Bachelor of Chemical Engineering degree in 1948.
I received a Master of Science degree in chemical engineering from the University of Cincinnati in 1949.
I also attended specialized courses in nuclear technology offered by the f4RC, Westinghouse, Aerojet-General Corporation, and University of California-Los Angeles.
I am a licensed professional engineer, registered in Pennsylvania (chemical engineering) and California (nuclear engineering).
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GLODE REQUA DIVISION OF LICENSING OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COMMISSION PROFESSIONAL QUALIFICATIONS My name is Glode Requa.
I am currently employed by the U.S. Nuclear Regulatory Ccamission as a Project Manager, Operating Reactors Branch No.1, Division of Licensing, Office of Nuclear Reactor Regulation. My duties include scheduling, managing, and participating in the review and evaluation of applications for license amendments for assigned power reactors.
I received a B.S. degree in Civil Engineering from New Jersey Institute of Technology in 1953 and did graduate studies at Carnegie Tech. in 1956 and 1957.
I have been associated with nuclear energy design, test, fabrication, and con-struction since 1956.
From 1956 to 1966 I was employed by the Westinghouse Electric Corporation, Bettis Atomic Power Laboratory, as a senior engineer designing and managing hydraulic and mechanical testing of Navy reactor compo-nents and produced mechanical designs of reactor components and instrumentation.
From 1966 to 1969, I was employed by Combustion Engineering, Inc.
I was in charge of in-core instrumentation design and later was manager of Design Quality Assurance.
I developed CE's first Design QA Program which was approved by the NRC.
I received a patent for an "inegrated flow measuring device" while employed at the Bettis Atomic Power Laboratory.
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Guy S. Vissing -
Professional Qualifica'tions Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission My name is Guy S. Vissing.
I am currently employed by the U.S. Nuclear Regulatory Commission as a Project Manager, Operating Reactors Branch No. 4, Division of Licensing, Office of Nuclear Reactor Regulation. As such, my duties include managing the licensing activities related to assigned power reactors, managing the licensing functions related to assigned multi-plant issues (one of these include the Pressurized Thermal Shock Issue, MPA B-67, MPF-B-73) and serving on task forces, committees, inspection teams and other ad hoc: assignments.
I received a B.S. degree with a major in Civil Engineering from Michigan State College in 1948. I am a registered Professional Engineer in the State of Michigan.
I have had a broad range of progressive experience in Project Management, reliability engineering, civil engineering and management of the above.
From 1948 to 1962, I was with two consulting engineering firms in responsible charge of structural & civil engineering design & preparation of specification, estimates, and plans for a large steam-electric power plant, small dams, industrial buildings and sewage treatment plants.
I was individually responsible for surveys, and design and analyses on many phases of'large electric power plants, nuclear power plants, hydro-electric power plants, office buildings, industrial facilities, sewage treatment plants, earth dams, spillways, concrete dams, roads and railroads. I was a resident engineer on construction.
While with the Government (NASA /AEC) from 1962-1973, I have been a reliability specialist as applied to review of nuclear safety, a reliability
& system engineering specialist on management of reliability programs for nuclear rocket and isotopic power systems, a manager of technology utilizatidn reporting program, and a project manager for planning and design efforts for nuclear rocket test site facilities.
While with the U.S. Nuclear Regulatory Commission from 1973 to present, I have been a reliability specialist and for the past seven years, a project manager for operating reactors.
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I LAMBROS LOIS DIVISION OF SYSTEMS INTEGRATION OFFICE OF NUCLEAR REACTOR REGULATION PROFESSIONAL QUALIFICATIONS My name is Lambros Lois.
I am currently empicyed by the U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation, Division of Systems Integration, Core Performance Branch. As such, my duties include review of fast neutron flux and fast neutron fluence calculations to the pressure vessel and review of the pressure vessel neutron surveillance program as related to the l
pressurtzed thermal shock of pressurized water reactors.
I received a Doctor of Science degree in Nuclear Engineering from Columbia University, New York City in 1965.
l I have been associated with nuclear energy research, development and regu-lation for 22 of the past 25 years.
I was employed by Argonne National Laboratory for two years as a trainee at the International School of Nuclear Science and Engineering and as a nuclear engineer. My duties included criticality and control rod worth calculations.
I was employed by Bettis Atomic Power Laboratory as a Senior Scientist for about 5 years, where I did research and development work in radiation transport and radiation shielding.
I was employed by Stone and Webster Engineering Corp. for about one year and my duties were in the design of nuclear power plants and particularly in the evaluation of radiation sources and radiation shield design.
I was employed by the Environmental Protection Agency for about one year. I have been employed by the Nuclear Regulatory i
Commission for about nine years. My responsibilities included fast reactor safety analysis, reactor systems, light water reactor reload methodologies and reloads.
I have about 25 publications in the fields of radiation transport and reactor l
safety.
I also have co-authored a book in the series of " Progress in Nuclear Science and Engineering."
1 PROFESSIONAL QUALIFICATION OF DR.HUGHW.(ROY) WOODS I am currently the NRC Task Manager for the Pressurized Thermal Shock Unresolved Safety Issue.
In that position, which I have held since November 1981, I an responsible for coordinating and directing all NRC activities towards generic resolution of this issue.
I am, therefore, familiar with all of the various aspects of the problem and its proposed resolution in the many technical discipline involved, including reactor system considerations.
Since 1973, I have been employed by the Huclear Regulatory Commission or its predecessor, the Atomic Energy Commission, in various capacities as a Huclear Engineer, most recently (before my present assignment) as the Office of Inspection and Enforcement principal reactor systems specialis+
for Westinghouse supplied nuclear plants.
Prior to 1973, I was employed by the E. I. DuPont Company at the Savannah River Laboratory, where I was responsible for various safety studies for their nuclear materials production reactors.
I hold Ph.D., M.S., and B. S. degrees in Nuclear Engineering with minors in Mechanical Engineering, Materials Engineering, and Electrical Engineering.
These degrees were awarded respectively in 1969 and 1965 by the University of Florida and 1964 by North Carolina State University.
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PHYLLIS SOBEL, PH.D.
GEOSCIENCES BRANCH DIVISION OF ENGINEERING U. S. NUCLEAR REGULATORY COMilSSION
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.My name is Phyllis Sobel and I am employed as a Geophysicist.in.
the Geosciences Branch, Division of. Engineering, Office of Nuclear Reactor Regulation, Washington, D.C. 20555. -
PROFESSIONAL QUALIFICATIONS l
In 1969.I received a B.S. degree in Geological Sciences from the Pennsylvania State University. I also pursued graduate studies at Princeton University and the University of Minnesota.
In 1978 I received a Ph.D. degree in Geophysics from the University of t
t From 1970 to 1973 I was a teaching assistant and research assistant at the University of Minnesota. I taught undergraduate laboratories in physical geology, historical geology, and oceanography courses.
My activity as a.research assistant was in the deve opment and use of a program to simulate marine magnetic anomalies. My interests in graduate school included all areas of geophysics, structural geology, and marine geology. My dissertation was a study of seismic phases reflecting off structures below the Earth's crust under several geo-graphic regions.
From 1973 to.1977 I was employed by Teledyne Geotech in Alexandria, Virginia as a research geophysicist. At this corporation's research laboratory I worked on a vuriety of research problems in seismology related to the detection of nuclear explosions, including (1) the useoffilters.toextractsignalsfromseismograms,(2)thepropagstion of Rayleigh waves through heterogeneities, and (3) the characteristics
~ of.. earthquakes in areas of proposed underground nuclear tecting in Asia..
l I am a member of the American Geophysical Union and the Seismological Society of America.
I have authored or co-authored ten papers published j
as Teledyne Geotech reports or in the Bulletin of the: Seismological r
Society of America.
I have authored or co-authored.two papers presented l
at meetings of the Seismological Society of America.
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l From October 1977 to March 1978 I was employed as.a seismologist by the NRC Office of Standards Development.in the revision and development of new regulatory guides.and standards and the supervision of, technical-assistance contracts related to. generic problems.found in the licensing process.. Since March 1978 I have been employed by the~Geotciences Branch in the evaluation of the seismological and geophysical data submitted to the NRC in support of a proposed seismic design basis.for nuclear facilities.
I have participated in the licensing activity for approx'imately fifteen sites.
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Professional Qualifications Division of Engineering Office of Nuclear Reactor Regulation
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Nuclear Regulatory Commission My name is Paul C. S. Wu.
I am currently employed by the U.S. Nuclear Regulatory Commission as a Chemical Engineer, Chemical Engineering
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Branch, Division of Engineering, Office of Nuclear Reactor Regulation.
As such, my duties include review of primary and. secondary water chemistry control and the corrosion of materials used in the con-struction and operation of nuclear power plants.
I received a Ph.D. degree in Materials Science from the Iowa State University in 1972.
i I have been associated with nuclear enery research, development, and construction as a research scientist, principal-engineer, and super-visor since 1967.
From 1967 to 1972, I was employed by Ames Laboratory, USAEC as a research scientist, responsible for nuclear materials research in coolant technolog and fuel reprocessing.
From 1972 to 1980, I was; employed by Westinghouse Electric Corporation on a variety of management and engineering positions, responsible for all materials 4
and corrosion programs related to nuclear fuel reprocessing and waste-management.
I wasN 1so in charge of the Mechanical Properties Laboratory 4
responsible for research and developmental prcgrams concerning fatigue, creep, fatigue-creep interaction, and stress rupture evaluation of nickel structural alloys.
I was also responsible for advanced nuclear control materials development.
I have more than 30 publications in the field of mechanical, physical, and chemical properties of nuclear materials.
I have also given 6 invited presentations in various national laboratories and research institutions.
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Louis Frank Professional Qualifications Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
-My name is Louis Frank.
I am a Senior Materials Engineer in the Inservice Inspection Section, Materials Engineering Branch, Division of Engineering, Office of Nuclear Reactor Regulation, of the. United States Nuclear Regulatory Commission.
In my present position, I am responsible for performing technical reviews and evaluations of PWR steam generator tube surveillance and repair programs for NTOL and operating plants.
I hold a Bachelor of Science Degree in Metallurgical Engineering and a Master of Science Degree in Metallurgy from the University of Kentucky and New York University, respectively.
I am also a Registered Professional Engineer in the State of Maryland.
I have a total of thirty-one years of professional experience of which thirty years has been in the nuclear field.
I was employed as a materials research engineer at General Telephone & Telegraph's Atomic Energy Labs in Bayside, N.Y. starting in 1952.
From 1955 thru 1963 I was a supervisory engineer in nuclear materials research and development at the Martin Co's. nuclear divsion.
From 1963 thru 1973 I was with two consulting firks.. engaged in nuclear safety studies.
Since joining the NRC in June 1973 I have been involved in corrosion and steam generator issues.
In the Office of Standards I prepared regulatory guides on steam generator inspection and plugging.
In the Office of Research I managed programs involving eddy-current inspection, particularly developing advanced techniques for conducting eddy-current inspections.
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David E. Smith Professional Qualifications Division of Engineering Office of Nuclear Reactor Regulation March 80 Materials Engineer to Materials Engineering Branch Date Division of Engineering Knowledgeable and experienced in welding, fabrication and inspec-tion of materials and other related engineering aspects of nuclear reactors.
Serves as a qualified materials engineer in the Materials Engineering Branch, Division of Engineering.
Responsible for reviews, analyses, and evaluation of safety issues related to structural and mechanical components of reactor facilities licensed for power opera-tion.
Participates as a technical reviewer in evaluating applications for construction permits and operating licenses for power and non-power reactors _and operational and design modifications of DOE and D0D-owned operating facilities exempt from the licensing process.
Specific assignments include review of operating license applica-tions for compliance with Standard Review Plans for which the Materials Application Section is responsib1'e.
EDUCATION:
Bachelor of Metallurgical Engineering, Rensselaer Polytechnic Institute, 1959 EXPERIENCE:
(Prior to joining NRC) l May 1967 to Materials Engineer Naval Sea Systems Command, Code 05E2, i
March 80 Washington, D.C.
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Responsible for materials specifications, Hull material development l
programs, consultant on welding, fabrication and inspection of metal structures, material selection, corrosion, machinery materials problems.
The hull materials development programs involved basic alloy research, the making and processing of all structural metal forms (castings, forgings, plate, extrusions, weld wire, rolled product), their fabri-cation (welding, cutting, machining, forming, painting), structural l
tolerances, and evaluation of structural performance, strength, toughness, corrosion, fatigue, compatibility with other materials, etc.
I would interface with material manufacturers, suppliers, ship-yards, and designers, and the type desks responsible for providing ships to the fleet.
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2-April 66 to Student.
Acquired commercial and instrument ratings for May 67 single engine land airplanes.
Dec 64 to Manufacturing Engineer for Ling Temco Vought, Centerline, April 66 MI.
Developed welding procedures for the LANCE missile tankage assembly.
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AFFIRMATION OF PREPARATION I, Lambros Lois, being duly sworn, state that I was reponsible for preparing the foregoing response to Interrogatories 1-9, 14-19, 21-24 Conten-tion 3.
Those responses are true and correct to the best of my knowledge.
L-w Lh Lambros Lois Subscribed And sworn to before me this day of July, 1983 n$ N) ary Publi
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AFFIRMATION PREPARATION I, Guy Vissing, being duly sworn, state that I was oc:ponsible for preparing the foregoing response to Interrogatories 10-12, Contention 3.
Those responses are true and correct to the best of my knowledge.
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My Comission expires: 7//)flo Subscribed and sworn to before me this lo* day of July,1983
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.. o AFFIRMATION OF PREPARATION I, Colode Requa, being duly sworn, state that I was reponsible for preparing the foregoing response to Interrogatories 25-29, Contention 3 and Interrogatories 8-10, Contention 8.
Thosa responses are true and correct to the best of my knowledge.
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Subscribed 3nd sworn to before me this (7 day of July,1983 Ykcuht./1#cm+
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AFFIRMATION OF PREPARATION I, Louis Frank, being duly sworn, state that I was reponsible for preparing the foregoing response to Interrogatories 30-41, 64-67, 78 and 85-87, Contention 3.
Those responses are true and correct to the best of my knowledge.
Ak Louis Frank Subscribed nd sworn to before me this /o day of July,1983 OCr7+L>
tary Publy
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My Comission expires: 7k 4 ), ) N[o V
t 3-AFFIRMATION OF PREPARATION I,BernardMann,beingdulysworn,statethatIwasrgp_qnsiblefor preparing the foregoing response to Interrogatories 45, 48 a~dd 54, Con-tention 3.
Those responses are true and correct to the best of my knowledge.
Ib R Bernard Mann Subscribed nd sworn to before me this (o day of July, 1983 b)1Y1 OCXJh
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My Commission expires: 7[/ %
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AFFIRMATION OF PREPARATION I, Paul Wu, being duly sworn, state that I was reponsible for preparing the foregoing response to Interrogatories 58 and 60-62, Contention 3.
Those responses are true and correct to the best of my knowledge.
Paul Wu Subscribed and sworn to before me this day of July, 1983 Notary Public My Commission expires:
AFFIRMATION OF PREPARATION I, David Smith, being duly sworn, state that I was reponsible for preparing the foregoing response to Interrogatories 74-77, Contention 3.
Those responses are true and correct to the of rny knowledge.
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David Smith.
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AFFIRMATION OF PREPARATION I, John Crooks, being duly sworn, state that I was reponsible for preparing the foregoing response to Interrogatory 19(b) and 79, Conten-tion 3 and Interrogatory 58, Contention 8.
Those responses are true and correct to the best of my knowledge.
John Crooks Subscribed and sworn to before me this day of July, 1983 Notary Public My Consnission expires:
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AFFIRMATION OF PREPARATION I, Paul E. Norian, being duly sworn, state that I was reponsible for preparing the foregoing response to Interrogatories 80-84, Contention 8.
Those responses are true and correct to the best of my knowledge.
Paul E. Norian Subscribed and sworn to before me this day of July, 1983 Notary Public My Commission expires:
AFFIRMATION OF PREPARATION I, Hugh W. Woods, being duly sworn, state that I was reponsible for preparing the foregoing response to Interrogatory 20, Contention 3.
That response is true and correct to the best of my knowledge.
D b enrtQ Hugh W(/ Woods Subscribed nd sworn to before me this 7 day of July, 1983 i
a otary Publ y My Comission expires: 7!l L i
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. 0 AFFIRMATION OF PREPARATION I, Phyllis Sobel, being duly sworn, state that I was reponsible for preparing the foregoing response to Interrogatories 12-14, Contention 8.
Those responses are true and correct to the best of my knowledge.
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Subscri nd sworn to befora is day of July,1983
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AFFIRMATION OF PREPARATION I, Richard Serbu, being duly sworn, state that I was reponsible for preparing the foregoing response to Interrogatories 52-57, Contention 8.
Those responses are true and correct to the best of my knowledge.
Richard Serbu Subscribed and sworn to before me this loc day of July, 1983
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7,!/)dlo My Comission expires:
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAF_ETY AND LICENSING BOARD In the Matter of:
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Docket No. 50-261 QLA CAROLINA POWER & LIGHT COMPANY
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(H.B. Robinson Steam Electric
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Plant, Unit 2)
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AFFIDAVIT OF PAUL BEMIS I, Paul Bemhs, being duly sworn, depose and say, 1.
I am a Section Chief with the U.S.N.R.C. assigned to Region II, Atlanta, Georgia.,
Theanswerstoknterrogatories21thru55onContentionslaandIb 2.
contained in the " Staff's Response to Hartv111e Groups Interrogatories" are true and correct to the best of my knowledge and belief.
The sources of information on which I base this statement are personnel of the U.S.N.R.C. assigned to Region II, Atlanta.
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f UNITED STATES OF ;-: ERICA NUCLEAR REGULATORY C0:4:4]SS10N BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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CAROLINA POWER AND LIGHT COMPANY
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Docket No. 50-261
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(H.B. Robinson Steam Electric
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(Steam Generator Repair)
Plant, Unit 2)
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CERTIFICATE OF SERVICE I hereby certify that copies.of "NRC STAFF RESPONSE T0. INTERROGATORIES OF THE HARTSVILLE GROUP" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk, through deposit in the Nuclear Regulatory Commission's internal mail system, this 7th day of July, 1983:
Morton B. Margulies, Chairman Administrative Judge Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, DC 20555*
Atomic Safety and Licensing Board Dr. Jerry R. Kline U.S. Nuclear Regulatory Commission Administrative Judge Washington, DC 20555*
Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Appeal Washington, DC 20555*
Board U.S. Nuclear Regulatory Commission Dr. David L. Hetrick Washington, DC 20555*
Administrative ~ Judge Professor of Nuclear Engineering Docketing and Service Section University of Arizona Office of the Secretary Tuscon, AZ 85721 U.S. Nuclear Regulatory Commission Washington, DC 20555*
George F. Trowbridge, Esq.
Shaw, Pittman, Potts & Trowbridge Bradley W. Jones, Esq.
1800 M Street, N.W.
Regional Counsel Washington, DC 20036 USNRC, Region II 101 Marietta St., NW B. A. Matthews Suite 2900 Hartsville Group Atlanta, GA 30303 P.O. Box 1089 Hartsville, SC 29550 Samantha Francis Flynn, Esq.
Carolina Power & Light Company P.O. Box 1551 Raleigh, NC 27602
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i0r.: John Ruoff'
~P.O. Box'96 Jenkinsville,:SC 29065 M
- Myron prman Deputy Assistant Chief Hearing. Counsel
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