Text

Sixth Partial Response to FOIA Request for Documents.Records in App J Encl & Will Be Available in Pdr.App K Records Withheld in Part (Ref FOIA Exemptions 7) & App L Records Completely Withheld (Ref FOIA Exemption 7)
	ML20207F620
Person / Time
Site:	Millstone, Vogtle, Oregon State University, 05000435
Issue date:	06/02/1999
From:	Racquel Powell; NRC OFFICE OF ADMINISTRATION (ADM)
To:	Ferraro D; AFFILIATION NOT ASSIGNED
Shared Package
ML20207F629	List: A08645, Responds to Requesting Info in Connection with Complaints Filed with DOL Wage & Hour Div by Three NNECO Employees; B13366, Informs That NNECO Has Provided Info to NRC Re Rosemount Transmitters at Plants; ML20207F620; ML20207F630;
References
	FOIA-98-32 NUDOCS 9906090156
	Download: ML20207F620 (11)
	v • d • e

_____-- _- - - . - -

NRC FORM 464 P4rt I ( U.S r UIAWA F.EsPoNsE NUMBER m

{

dt Q . NUCLEAR REGULATORY 98-032 COMMISSION p

p5# "'% 6

ESPONSE TO FREEDOM OF I

INFORMATION ACT(FOlA)/ PRIVACY RESPONSE - -

Er k*...*h ACT (PA) REQUEST TYPE -

REQUESTER DATE Donald P., Ferraro Jtill 021999 PART I -INFORMATION RELEASED

+

] No additional agency records subject to the request have been located.

j ] Requested records are available through another public distribution program. See Comments section.

APPENDICES Agency records subject to the request that are identified in the listed appendices are already available for

] public inspection and copying at the NRC Public Document Room r [APPENosCES~l Agency records subject to the request that are identified in the listed appendices are being made available for

[ J, K l public irispection and copying at the NRC Public Document Roum.

] Document Room. 2120 L Street, NW, Washington, DC. Enclosed is information on how you may obtain a

--- ry ' APPENDICES y 4g Agency records subject to the request are enclosed.

I l Records subject to the request that contain information originated by or of interest to another Federal agency have been

-i referred to that agency (see comments section) for a disclosure determination and direct ruponse to you.

Q We are continuing to process your request.

See Comments.

_I~ .,

~

PART l.A - FEES

- FAMouNT * ~ You wil: be billed by NRC for the amount listed. ~ None. Minimum fee threshold not met.

S You will receive a refund for the amount listed. Fees waived.

. ,,,n,,

PART l.B -INFORMATION NOT LOCATED OR WITHHELD FROM DISCLOSURE

] N' agency records subject to the request have boon located.

y~ the reasons stated in Part II.Certain information in the requested records is being withheld from disclosure pursuant to y This determination may be appealed within 30 days by writing to the FOIA/PA Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Clearly state on the envelope and in the letter that it is a "FOIA/PA Appeal."

PART l.C COMMENT 5 (Use attached Comments continuation page if required) 7 f ,/ ( j M ] [f BIGNATURE .IRELDOM OF INF ORMATIO9t 1 AND P Acy AL) OHIGLH l Russe o[e 9906090154 990602 PDR FOIA r

FERRAR099-32 PDR NRC FORM 464 Part 1 (61998) PRINTED oN RECYCLED PAPER This form was desgned using informs

i NRC FORM 464 Part 18

( U.S. NUCLEAR REGULATORY COMMISSION FOlA/PA DATE

{

f 7ESPONSE TO FREEDOM OF INFORMATION JilN 021999 )

ACT (FOIA) / PRIVACY ACT (PA) REQUEST PART ll.A - APPLICABLE EXEMPTIONS Records subject to the request that are described in the enclosed Appendices are being withheld in their entirety or in part under l,AND'CM_]K&L _

the Exempbon No.(s) of the PA and/or the FOIA as indicated below (5 U S.C. 552a and/or 5 U

] Exemption 1: The withheld information is property classifed pursuant to Executive Order 12958.

{

{ Exemption 2: The withheld information relates solely to the internal personnel rules and procedures of NRC.

R Exemption 3: The withheld information is specifically exempted from public disclosure by statute indicated. I m Sections 141 145 of the Atomic Energy Act, which prohibits the disclosure of Restricted Data or P.,rmerly Restricted Data (42 U.S C.

" 2161 2165).

{ Section 147 of the Atomic Energy Act, which prohibits the disclosure of Unclassified Safeg sards information (42 U.S.C. 2167).

~

41 U.S.C., Section 253(b), subsection (m)(1), prohibits the disclosure of contractor proposals in the possession and control of an L executive agency to any person under section 552 of Title 5, U.S C. (the FOIA), except when incorporated into the contract between the agency and the submitter of the proposal.

pJ Exemption 4: The withheld information is a trade secret or commercial or financial information that is being withheld for the reason (s) indicated.

] The information is considered to be confidential business (proprietary) information. l r] The information is considered to be proprietary because it concems a licensee's or applicant a physical protection or material control and u accounting program for special nuclear matenal pursuant to 10 CFR 2.790(d)(1).

((] The information was submitted by a foreign source and received in confidence pursuant to 10 CFR 2.790(d)(2)

Exemption 5: The withheld information consists of interagency or intraagency records that are not available through discovery dunng

.- litigation. Applicable privileges:

Deliberative process: Disclosure of predecisional information would tend to inhibit the open and frank exchange of ideas essential to the l deliberative process Where records are withheld in their entirety, the facts are inextncably intertwined with the predecisional information. TI.ere also are no reasonably sepyegable factual portions because the release of the facts would permit an endtrect inquiry into the predecisional process of the agency.

}] Attomey work-product pnvilege. (Documents prepared by an attomey in contemplation of litigation)

] Attomey client privilege. (Confidential communications between an attomey and his/ lier client)

- Exemption 6: The withheld information is exempted from public disclosure because its disclosure would result in a clearly unwarranted invasion of personal pnvecy.

y Exemption 7: The withheld information consists of records compiled for law enforcement purposes and is being withheld for the reason (s) endicated.

~ ~ (A) Disclosure could reasonably be expected to interfere with an enforcement proceeding (e g., it would reveal the scope, direction, and focus of enforcement efforts, and thus could possibly allow recipients to take action to shield potential wrongdoing or a violation of NRC requirements from investigators).

Q (C) Disclosure would constitute an unwarranted invasion of personal privacy.

(D) The information consists of names of individuals and other information the disclosure of which could reasonably be expected to reveal identities of confidential sources.

~~ (E) Disclosure would reveal techmques and procedures for law enforcement investigations or prosecutions, or guidelines that could reasonably be expected to nsk circumvention of the law.

~ (F) Disclosure could reasonably be expected to endanpa. the life or physical safety of an individual.

] OTHER (Specify)

PART ll.B - DENYING OFFICIALS Pursuant to 10 CFR 9.25(g),9.25(h), and/or 9.65(b) of the U.S. Nuclear Regulatory Commission regulations, it has been determined that the information withheld is exempt from production or disclosure, and that its production or disclosure is contrary _to the public interest. The person responsible for the denial are those officials identified below as denying officials and the FOlA/PA Officer for any denials that may be appealed to the Executive Director for Operations (EDO).

~ ~

'DENEINEOFFICIA TITLE / OFFICE RECORDS DENIED ((OQCY

~~~~ ~ ~

J_ mes ticIAman ~ Director ' Office of Enforcement ~ Appendices E & L

- - . - . - . . . ~ . . _ - - - - - . ..

~ ~ ~ ~ ~ ~

5ipeafmust be made in writing witsn 30' days of ' receipt of this response. Appeals should' be maned to the FOIA/Prsacy Act Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, for action by the appropriate appellate official (s). You should clearly state on the envelope and letter that it is a "FOIA/PA Appeal."

PRINTED ON RECYCLLD PAPER TNs torrn was cestoned using inF orms NRc FORM 464 Part il (61998)

E J

i ,

l..

l Re: FOIA-98-246 APPENDIX J RECORDS BEING RELEASED IN THEIR ENTIRETY -

(if copyrighted identify with *) ,

1 lEL DATE DESCRIPTION #PAGE COUNT)

ENCLOSURES TO REPORT OF NU CONCERNING ALLEGATIONS OF PAUL M. BLANCH BEING RELEASED IN THEIR ENTIRETY:

1. 03/01/91 Letter to L. Martin from 1. Sikorsky. (1 page) RELEASE

2. 04/18/91 Letter to L. Martin from N. S. Reynolds. (2 pages) RELEASE

3. 05/10/89 Letter to C. M. Rice from E. Mroczka. (5 pages) RELEASE

4. 08/28/89 LRS Incorporated Report Visit # 1-89/2-89. (23 pages) l I

RELEASE

5. 02/07/89 Letter to NU Director of QA from S. Wanek. (4 pages)

RELEASE

6. 04/13/89 Letter to USNRC from E. J. Mroczka. (10 pages) RELEASE

7. 05/12/89 Memo to E. Mroczka from G. L. Johnson. (6 pages)

RELEASE

8. 06/28/89 Letter to E. Mroczka from E. Wenzinger transmitting NRC Inspection Report 50-423/89-04) (20 pages) RELEASE

9. 08/01/89 Letter to USNRC from E. Mroczka. (8 pages) RELEASE

10. 10/31/89 Letter to USNRC from E. Mroczka. (4 pages) RELEASE

11. 05/01/92 Letter to Chief, Rules and Directives, USNRC, from P.

Blanch. (8 pages) RELEASE

12. - 04/01/90 Northeast Utilities Organization Chart, NE&O. (5 pages)

. RELEASE

13. 08/31/88 Master Performance Rating of Paul M. Ble.nch (1988). (7 pages) RELEASE l

c

14. 08/31/86 Master Performance Rating of Paul M. Blanch (1986). (4 pages) RELEASE

15. 09/30/87 Master Performance Rating of Paul M. Blanch (1987). (7 pages) RELEASE l

l

16. 10/05/92 Memorandum to E. Mroczka from R. Wemer. (9 pages)

RELEASE

17. 11/30/88 Memorandum to A. R. Roby from Paul M. Blanch. (2 pages)

RELEASE

, 18. 11/30/88 Memorandum to Paul M. Blanch from A. R. Roby. (2 pages)

RELEASE

19. 03/09/89 Conflict of Interest Form - Paul Blanch. (1 page) RELEASE

20. .02/09/90 Conflict of Interest Form - Paul Blanch. (1 page) RELEASE

21. 12/06/92 Letter to J. Weiss from R. Lord. (25 pages) RELEASE

22. 01/19/89 Uiemorandum to R. C. Enoch from P. Blanch. (3 pages)

RELEASE

23. 02/24/89 Memorandum to M. F. Samek from P. Blanch. RELEASE l

24. 04/04/89 Memorandum to A. R. Roby from P. Blanch. (2 pages)

RELEASE

25. 05/11/89 Memorandum to M. F. Samek from Paul Blanch. (2 pages)

RELEASE

26. 05/12/89 Memorandum to R. M Kacich from Paul Blanch. (2 pages)

RELEASE'

27. 10/10/89' Letter to distribution regarding next scheduled meeting of the BWROG Committee on Rosemount Transmitters. (1 page) RELEASE l

i 28. 06/23/89 Memorandum to J. Becker from P. Blanch. (3 pages) !

l RELEASE

29. 12/05/89 Memorandum to R. C. Enoch from P. Blanch. (3 pages)

RELEASE f .

I i

30. 12/13/89 Memorandum to A. R. Roby from P. Blanch. (2 pages)

RELEASE

31. 03/28/90 Memorandum to P. M. Blanch from J. F. Opeka. (12 pages)

RELEASE -

32. 08/07/89 Memorandum to A. R. Roby from P. Blanch. (1 page)

RELEASE

33. 10/23/89 Memorandum to A. R. Roby from P. Blanch. (10 pages)

RELEASE

34. 11/01/89 Memorandum to E. J. Mroczka from P. Blanch. (2 pages)

RELEASE

35. Various News articles. (5 pages) RELEASE *

36. 08/05/92 Memorandum to E. Debarba from P. Blanch. (4 pages)

RELEASE ,

37. 09/03/92 Letter to L. Norton from P. Blanch (9 pages) RELEASE

38. 09/13/92 Newspaper article. (1 page) RELEASE *

39. 08/07/92 Complaint, Paul M. Blanch v. Rosemount, Inc., (Connecticut Supenor Coud. (10 pages) RELEASE

40. 11/05/92 Transcript Excerpt of NRC Public Meeting to Receive Comments on Northeast Utilities' Performance Enhancement Program. (9 pages) RELEASE

41. 11/05/92 Statement by Paul M. Blanch - NRC Public Meeting on 11/5/92, Niantic, CT. (7 pages) RELEASE ,

l

42. 04/07/89 Memorandum to R. P. Wemer from A. R. Roby. (7 pages)

RELEASE l l

43. 04/06/89 Memorandum to G. L. Johnson from Paul M. Blanch. (3 l l pages) RELEASE 1

44. 04/11/89 Notes to the File from A. R. Roby. (6 pages) RELEASE l

l 45. 10/20/89 Memorandum to Paul Blanch from A. Roby. (6 pages) l RELEASE l

p ,

l' l

l l 46. 04/06/89 Memorandum to the File from A. R. Roby. (12 pages)

RELEASE

47. 04/10/89 Memorandum to the File from G. L. Johnson. (1 page)

RELEASE

48. 05/03/89 Memorandum to R. P. Wemer from E. J. Mroczka. (2 pages)

RELEASE

49. 05/15/89 Memorandum to J. s. Keenan and C. Clement from A. Roby.

(1 page) RELEASE

50. 04/04/89 Draft Memorandum to Paul Blanch from A. Roby. (2 pages)

RELEASE

51. 04/06/89 Memorandum to G. L. Johnson from A. Roby. (2 pages)

RELEASE

52. 04/12/90 Memorandum to Paul Blanch from J. Opeka. (6 pages)

RELEASE

53. 10/03/89 Memorandum to J. Opeka from C. Frederick Sears. (2 pages) RELEASE

54. 06/12/92 Memorandum to the File from R. M Kacich. (2 pages)

RELEASE

55. 10/02/89 Memorandum to the File from B. M. fox. (3 pages)

RELEASE

56. 07/20/89 Letter to D. A. Rockwell from E. J. Mroczka. (1 page)

RELEASE

57. 04/17/90 Letter to Secretary of State (CT) from P. Blanch. (1 page)

RELEASE

58. 01/22/91 Letter to Allen Pollock from D. Guay. (2 pages) RELEASE

59. 11/14/89 IAD Report-Investigation of Alleged Time & Expense Abuse Generation Electrical Engineering. (6 pages) RELEASE

60. 10/19/89 Intemal Audit Report from M. Marinaccio to E. Mroczka regarding Blanch /EPRI Project Investigation. (3 pages)

RELEASE l .

. l

I i

61. 02/09/90 Letter to John Taylor from E. Mroczka. (1 page) RELEASE

62. 05/04/89 Memorandum to the File from G. L. Johnson. (4 pages)

RELEASE

63. 12/13/88 Memorandum to All NE&o Employees from E. J. Mroczka. (2 pages) RELEASE

64. 04/28/89 Memorandum to G. L. Johnson from A. R. Roby. (2 pages)

RELEASE

65. 10/11/89 Memorandum to R. P. Wemer from G. L. Johnson. (1 page)

RELEASE

66. 11/14/89 Memorandum to R. P. Wemer from E. J. Mroczka. (1 page)

RELEASE

67. 11/28/89 Handwritten notes of R. P. Wemer meeting with Nir Bhatt. (2 pages) RELEASE

68. 04/12/89 Memorandum to File from G. L. Johnson. (1 page)

RE!. EASE -

69. 10/16/89 Memorandum to A. R. Roby from E. J. Mroczka. (3 pages) i RELEASE

70. 10/16/89 Memorandum to G. L. Johnson from E. J. Mroczka. (3 pages) l RELEASE

71. 10/16/89 Memorandum to T. A. Shaffer from E. J. Mroczka. (3 pages) ;

RELEASE I l

72. 10/16/89 Memorandum to R. P. Wemer from E. J. Mroczka. (3 pages)

RELEASE ;

73. 10/04/89 Memorandum to B. M. Fox from Paul Blanch. (2 pages) i RELEASE

74. 10/12/89 Memorandum to Paul Blanch from B. M. Fox. (1 page)

. . RELEASE

75. 04/09/90 Letter to Thomas Martin from E. J. Mroczka with attachments. (45 pages) RELEASE

r .

]

l Re: FOIA-98-032 APPENDIX K RECORDS BEING WITHHELD IN PART N_4 DATE DESCRIPTION /(PAGE COUNTVEXEMPTIONS ENCLOSURES TO REPORT OF NU CONCERNING ALLEGATIONS OF PAUL M. BLANCH BEING RELEASED IN PART:

1. 11/24/92 Repen of Northeast Utilities Conceming Allegations of Paul l M. Blanch. (125 pages) RELEASE IN PART, EX. 7(C)

2. 02/20/90 Transcript of Conference Regarding issues of Concem to Paul Blanch. (128 pages) RELEASE IN PART - EX. 7(C)

{ NOTE: ONLY THE ODD PAGES OF THIS DOCUMENT !

l l COULD BE LOCATED IN NRC FILES)

3. 09/26/92 Trip Request Authorization (Paul M. Blanch) (7 pages)

RELEASE IN PART - EX. 7(C) ;

4. 05/22/89 Trip Request Authorization (Paul M. Blanch) (1 page) l RELEASE IN PART - EX. 7(C)

5. 10/26/89 Report of interview with Paul Martin Blanch. (13 pages) j RELEASE IN PART - EX. 7C l

6. 04/11/92 Resume of Amold R. Roby. (2 pages) RELEASE IN PART-I EX. 7(C)

7. 11/18/92 Resume of G. Leonard Johnson. (2 pages) RELEASE IN PART - EX. 7(C) l 8. Undated Resume of Thomas A. Shaffer. (2 pages) RELEASE IN PART - EX. 7C

9. Undated Resume of Edward J. Mroczka. (5 page). RELEASE IN PART - EX. 7(C)

10. May/1992 Resume of John F. Opeka. (1 page) RELEASE IN PART -

EX. 7(C)

11. Undated Resume of C. Frederick Sears. (4 pages) RELEASE IN PART - EX. 7(C) l 1

l

r .

1

\

12. July /1992 Resume of Bemard M. Fox. (2 pages) RELEASE IN PART EX. 7(C)

13. 09/22/89 Memorandum to the File from Allen Pollock. (4 pages)

RELEASE IN PART- EX.7(C) l 14. Undated Transcript of Interview with P. Blanch. (43 pages) l RELEASED IN PART - EX. 7(C) i

15. 02/28/85 Memorandum regarding sign out sheets. (1 page) l RELEASED IN PART - EX. 7(C) t l

i i

I 1

r ,

f Re: FOIA-98-032 APPENDIX L RECORDS BEING WITHHELD IN THElR ENTIRETY fgh p_ ATE DESCRIPTION /(PAGE COUNT 1/ EXEMPTIONS ENCLOSURES TO REPORT OF NU CONCERNING ALLEGATONS OF PAUL l M. BLANCH BEING WITHHELD IN THEIR ENTIRETY:

1. 09/18/89 Transcript of IAD Interview of Alleger. (38 pages)

WITHHELD IN ENTIRETY - EX. 7(D) .

2. 03/14/90 LAD Review of P. M. Blanch's Complaint Regarding Alleged improprieties. (7 pages) WITHHELD IN ENTIRETY- EX.

7(C)

3. 10/06/89 LAD Worksheet and attachments. (11 pages) WITHHELD IN l ENTIRETY - EX. 7(C) l

4. 5/23-6/13/89 LAD Worksheet. (3 pages) WITHHELD IN ENTIRETY - EX.

7(C)

5. 6/27-7/21/89 LAD Worksheet. (3 pages) WITHHELD IN ENTIRETY - EX.

7(C)

6. 7/28-8/29/89 IAD Worksheet. (2 pages) WITHHELD IN ENTIRETY - EX.

7(C)

7. 9/8-20/89 IAD Worksheet. (3 pages) WITHHELD IN ENTIRETY - EX.

i 7(C) l

8. 5/19 26/89 LAD worksheet. (3 pages) WITHHELD IN ENTIRETY - EX. j 7(C) l 9. 6/2-13/89 IAD Worksheet. (3 pages) WITHHELD IN ENTIRETY - EX.

7(C) i

10. 6/19-29/89 IAD Worksheet. '(3 pages) WITHHELD IN ENTIRETY - EX.

7(C)

11. 6/30-7/7/89 LAD Worksheet. (3 pages) WITHHELD IN ENTIRETY - EX.

7(C)

L

, 12. 7/18-8/4/89 IAD Worksheet. (3 pages) WITHHELD IN ENTIRETY - EX.

7(C)

13. 7/18-8/4/89 14D Worksheet. (3 pages) WITHHELD IN ENTIRETY - EX.

7(C)

14. 8/16-9/1/89 IAD Worksheet. (3 pages) WITHHELD IN ENTIRETY - EX.

7(C)

1 .

N EGAN & ASSOCIATES, P.C.

Counselors at law 2300 N Street, N.W Washington, D.C. 20037 l Telephone (202) 663-9200 January 20,1998 & (202) 663-9066 FOINPAREQUEST Natalie Brown TsoNo: 92- o 3 a.

FOIA/LPDR Branch .zlo Roc'& .__/.; M- 98 U.S. Nuclear Regulatory Commission 'cha Ofi: _ B n.oun<_.

l Mail Stop T6D8 Shd C*%' _ 71-gi2 Washington, DC 20555 0001 Re: Freedom ofInformation Act Reauest i

Dear Ms. Brown:

{

1 Pursuant to the Freedom ofInformation Act (5 U.S.C. Q 552) and the U.S. Nuclear Regulatory Commission ("NRC") regulations (10 C.F.R. Part 9, Subpart A), I request that you reopen FOIA-97-312, originally filed on August 15,1997. As you recall, FOIA 312 was revised several times, the most recently in October 1997. The latest version of that FOIA is as follows:

1. Any and all records (including without limitations, notes of I telephone conversations and teleconferences) that discuss, refer, or relate to the design, construction, licensing, operations, employee concerns, or maintenance of Millstone Units 1,2, or 3 for the period '

January 1,1992, through December 31,1996, that were produced or I

written by any person under the letterhead of the Washington, D.C.

law firm ofWinston & Strawn;

2. Any and all records (including without limitations, notes of telephone conversations and teleconferences) that discuss, refer, or relate to the design, construction, licensing, operations, employee concerns, or maintenance of Millstone Units 1,2, or 3 for the period January 1,1992, through December 31,1996, that were produced or written by any person under the letterhead of the Washington, D.C.

l law firm Bishop, Cook, Purcell & Reynold's; jfG U SO! M .3/P

3 f,.-

,4 1

Ms. Natalic Brown EGAN & ASSOCIATES, P.C.

January 20,1998 C "3*"

la' Page 2

3. Any and all records (including without limitations, notes of telephone convenations and teleconferences) that discuss, refer, or relate to the design, construction, licensing, operations, employee concerns, or maintenance of Millstone Units 1,2, or 3 for the period January 1,1992, through December 31,1996, that were produced or :

written by any pemon under the letterhead of the Washington, D.C.

law firm of Morgan, Lewis & Bockius (or its predecessors, Newman

& Holtzinger, and Newman, Bouknight & Edgar);

- 4. Any and all records (including without limitations, notes of telephone convenations and teleconferences) that discuss, refer, or relate to the design, construction, licensing, operations, employee l concerns, or maintenance of Millstone Units 1, 2, or 3 for the period January 1,1992, through December 31,1996, that were produced or written by any person under the letterhead of the Washington, D.C.

law firm ofNewman & Holtzinger;

5. Any and all records (including without limitations, notes of telephone conversations and teleconferences) that discuss, refer, or relate to the design, construction, licensing, operations, employee concerns, or maintenance of Millstone Units 1,2, or 3 for the period January 1,1992, through December 31,1996, that were produced or written by any person under the letterhead bf the Washington, D.C.

law firm of Newman, Bouknight & Edgar;

6. Any and all records (including without limitations, notes of telephone convenations and teleconferences) that discuss, refer, or relate to the design, construction, licensing, operations, employee concerns, or maintenance of M'illstone Units 1,2, or 3 for the period January 1,1992, through December 31,1996, that were produced or written by any pemon under the letterhead of the Washington, D.C.

law firm of Shaw, Pittman, Potts & Trowbridge; and

7. Any and all records (including without lintitations, notes of telephone comersations and :eleconferences) that were produced or written by any employee or agent of the NRC, which review,

l- .

Ms. Natalie Brown EGAN & ASSOCIATES, P.C.

. January 20,1998 couwlon or zu Page 3 ,

analyze, comment on, or refer to any of the documents requested in paragraphs 1 through 6, or of any communications between any

('

NRC employee and any of the law firms referred to in paragraphs 1 through 6 relating to Millstone Units 1,2, or 3.

'Ihe above-requested records should include those available at NRC Headquarters, Region I, and in the NRC offices at the Millstone site. To the degree that any or all of the above-requested records are already publicly .

available and readily retrievable in the Washington, D.C. area, please provide a list of such records with the appropriate references to facilitate acquisition.

As we discussed by telephone on January 14, your last statement of estimated fees for FOIA-97-312 was $14,141.40. Consequently, I have attached two checks payable to U.S. Nuclear Regulatory Commission, for $8,767.67 and $5,373.73. It is my understanding that if the actual costs are less than the estimate, the remaining funds will be refunded. It is also my understanding that if the actual costs exceed the estimate, you will bill me for that additional amount.

If you need any further information to process this request, please do not hesitate to call me at (202) 663-9200.

s Sincerely, Eka,~cQ Donald P. Ferraro

I*

l NORTHEAST UTILITIES R

a.n.,. omm . seen sue.i son C enu a 0- 1 - i e7 U$sNw.*

i lf P O c0X 270 k ' J

[*[,',*'[,',*,,**,",,*,",,a ,[,,, HAATFORD. CCNNECTicUT 06141-027;

,,o.. . % . .c,. co ,. 5203) 665-5000 l

April 13, 1989

]

Docket Nos. 50-213 50-245 50-336 50-423 B13178 i

U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

References:

(1) E. J. Mroczka (NU) letter to W. T. Russell (NRC), Report of Substantial Safety Hazard, B12863, March 25, 1988.

(2) Rosemount letter to Northeast Utilities, Rosemount Nuclear Qualified Transmitters, December 9,1988. l l

(3) Rosemount letter to Northeast Utilities, Notification !

Under 10CFR21, February 7, 1989. )

Gentlemen:

Haddam Neck Plant !

Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3 )

Rosemount Transmitters !

l This information letter is being submitted to provide a summary of our activ- ,

ities involving Rosemount transmitters. Since Rosemount transmitter failures !

were first identified at Millstone Unit No. 3 in 1987, Northeast Duclear Energy Company (NNECO) personnel have been actively involved in evaluating and addressing this concern. In addition, we have had a number of discussions l with the vendor, NRC inspectors, and various industry representatives, and j have met NRC reporting requirements on this issue. Since our knowledge has substantially increased on this issue since the docketing of Reference (1),

NNECO believes it is appropriate to ensure the NRC Staff is more fully j informed, by providing this information letter.

Backaround During the first cycle of Millstone Unit No. 3 operation, five Rosemount capacitive-type differential pressure transmitters failed in the reactor coolant system. Twelve such transmitters are used to monitor reactor coolant i flow in the primary loops and provide a reactor trip signal to the reactor protection system. There are three transmitters in each loop. If any two of the three transmitters in a loop sense low flow, a trip signal is initiated.

The failures occurred individually over a period between March and November 1987, such that two transmitters never failed imultaneously in any one loop.

mi..

,b wn

4 9

0 U.S. Nuclear Regulatory Commission B13178/Page 2 April 13, 1989 4

Each failed transmitter was taken out of service and the affected channel was placed in the trip condition. The transmitters that failed were all Rosemount mooel 1153 HD5PCs.

Attempts to calibrate the failed transmitters were unsuccessful and the transmitters were replaced. The five failed transmitters were returned to Rosemount and destructive testing determined a loss of oil to be the cause of the failed condition.

The individual failures of these transmitters were evaluated for reportability in accordance with station procedures. The plant remained in compliance with technical specifications and no reportable condition existed under 10CFR50.72 and 50.73. However, NNECO concluded that the number of failures was of concern. Accordingly, an evaluation was initiated in late November 1987 to determine if a Substantial Safety Hazard existed. This included technical reviews performed by engineering disciplines and operations personnel, meetings with the vendor, deliberations by the Millstone Unit No. 3 Nuclear Review Board, and a final determination by a senior corporate officer. A conclusion was reached that the five failures represented a Substantial Safety Hazard, and the NRC Staff was accordingly notified under 10CFR21 on March 25, 1988 (Reference 1). That notification also informed the NRC Staff that the Rosemount problem existed at Millstone Unit No. 3 and not at NU's other plants.

The Haddam Neck Plant does not have Rosemount transmitters. Millstone Unit No. I has ten safety-related Rosemount transmitters, model 1152, that are used only for ATWS mitigation. Millstone Unit No. 2 has Rosemount transmitters, :

three of which are used in safety-related applications, and two of these are ;

similar in model number (ll53HD5PA) to the Millstone Unit No. 3 failures. ;

Reference 1 also reported that Millstone Unit Nos. I and 2 have not exper-ienced this problem. I In addition to having replaced all failed transmitters by the end of the first cycle, NNECO's corrective actions included the priparation of an in-service test procedure, and a monthly test of all twelve (12) Rosemount transmitters i in the reactor coolant system throughout Cycle 2 operation. Millstone Unit l No. 3 is scheduled to shut down for the second refueling on May 20, 1989.

Since the first cycle, no additional failures of Rosemount transmitters have been observed. .

Continuino Review Recogkizing .that the monthly test may not in itself provide total operability information, in addition to the monthly surveillance being performed at Millstone Unit No. 3 over the past year, NNECO has continued to investigate the Rosemount transmitter issue. During a meeting on February 6,1989, NNECO concluded that all new information should be evaluated to ensure operability and reportability reauirements were being fully met. Accordingly, new eval-uations were initiated on February 8,1989 for Millstone Unit Nos.1, 2, and

U.S. Nuclear Regulatory Commission B13178/Page 3 April 13, 1989

3. The evaluations focused on new information provided by Rosemount and on information that was learned as a result of increased analysis of data from Millstone Unit No. 3.

In Reference (2), Rosemount confirmed to NNECO that there was a problem affecting Rosemount transmitter models 1153 and 1154. The letter stated that a small number of these transmitters may respond sluggishly to input changes or may drift outside normal specifications. The source of the problem was identified as a loss of oil within the sensor cell. Several oil loss paths ,

were possible: the glass-to-metal interface, fill tube and damaged isolator diaphragm. The letter stated that the failure was a random and low probabil-ity event. Finally, Rosemount identified that increased acceptance criteria were added to the cell manufacturing and testing process, to assure a reliable product.

In Reference (3), Rosemount provided a 10CFR21 Notification to NNECO concern-ing the problem affecting Rosemount transmitter models 1153 and 1154. In this letter, Rosemount stated that the loss of oil from the sensing cell may cause a reduction in transmitter performance (such as drift, lack of response and an increase in response time). The letter also stated that the problem may be unidirectional. In addition, no firm limits could be placed upon the perfor- l mance reduction of a failed transmitter.

The scope of the problem as indicated in Reference (3) is believed to be limited. Information used in Rosemount's assessment suggests that trans- l mitters in service longer than 36 months may not exhibit the loss of oil '

failure, i.e., failures exhibit an infant mortality nature. This is because the failure is related to a specific manufacturing process and is not directly service related. No known loss of oil failure has occurred after 30 months of service. The Millstone Unit No. 3 transmitters now have approximately 36 months of service.

The information in Reference (3) did not rule out generic applicability. The l letter stated that ... prior to detectable fail're, u the transmitter may ;

continue to provide a signal but not respond over its full range and/or time i response may be significantly degraded. This may be a safety concern at your l plant." Rosemount also stated that action has been taken to correct the i source of the problem by " improving the manufacturing process and intensifying test criteria." Therefore, "

,.. potential for failures of this nature in transmitters currently being produced has essentially been eliminated."

Reference (3) also identified that all failures reported to Rosemount have ;

occurred in certain groups of transmitters. Transmitters from these groups that were shipped to the Millstone site were identified in an attachment to the Rosemount letter. A total of 16 transmitters were listed as having come from batches with confirmed failures. Our review of this list confirmed that 13 of these were used in Millstone Unit No. 3 and three of these were used in Millstone Unit No. 2. We have identified the locations of these transmitters

i 1

{

l.* U.S. Nuclear Regulatory Commission B13178/Page 4 April 13, 1989 i

l i

in our plan's, t and determined the effect that reduced transmitter performance may have in each application as described below.

Millstone Unit No. 3 13 Transmitters from Susoect Batches The identification, function, failure effects and calibration results for each of the 13 transmitters from manufacturing batches with confirmed failures is described below:

Additional actions that provide increased operability confidence are described later in this letter.

Model No. Serial No. Plant Tao No. Function 1153DB5 408073 3CHS*LT102 Boric Acid Tank Lvl. TK5A 11530B5 408074 3CHS*LT104 Boric Acid Tank Lvl. TK5A 1153DB5 408076 3CHS*LT106 Boric Acid Tank Lv1. TK5B L

These transmitters are used to measure boric acid tank level. Two of the transmitters are on Tank 5A and one is on Tank 58. These transmitters are listed in the Unit 3 Technical Specification Section 3.3.3.5 as being part of the required Remote Shutdown Instrumentation. There are two sensors on each tank and the minimum number of sensors required is one. Therefore a failure of any one of these sensors would not jeopardize remote shutdown capability.

These transmitters were calibration checked in February 1989 and found to be in normal working order. 1 l

Model No. Serial No. Plant Tao No. Function 1153085 408078 35WP-FT59A CTMT Recire Cooler A !

outlet Flow l 1153085 408079 3WSP-FT59B CTMT Recirc Cooler B ;

. outlet Flow These transmitters are used to measure service water outlet flow through the containment recirculation coolers. They provide only a monitoring function.

Failure of the transmitters would not prevent any of the recirculation coolers from performing their safety function. These transmitters were calibration checked in February 1989 and found to be in normal working order.

Model No. Serial No. Plant Tao No. Function Il53HD5 408188 3RCS*FT424 RCS Flow Loop 2 1153HD5 408190 3RCS*FT426 RCS Flow Loop 2 1153HD5 408193 3RCS*FT436 RCS Flow Loop 3 These transmitters are used to measure reactor coolant loop flow. There are a total of three transmitters installed in each loop. Two out of three logic is used to provide a. low flow reactor trip at full power. Two of the above l

l.

l L

1 U.S. Nuclear Regulatory Commission B13178/Page 5 April 13, 1989 transmitters are installed in loop 2 and one is installed in loop 3. If the two transmitters installed in loop 2 fail simultaneously then the plant would not be able to provice a loop 2 low flow reactor trip. In February 1989, all RCS flow transmitters were monitored for performance using high speed data acouisition equipment and/or using the Offsite f acility Information System (0FIS). All transmitters were found to be in normal working order.

Model No. Serial No. Plant Tao No. Function 1153HD5 408198 3RCS*LT461 Pressurizer Level This transmitter is used to measure pressurizer level. There are three transmitters measuring pressurizer level and providing a 2 out of 3 high-lesel reactor trip. Should the transmitter fail there would still be two other transmitters that could provide the reactor trip function. This transmitter was calibration checked in February 1989 and found to be in normal working order.

Model No. Serial No. Plant Tao No. Function 1153GD8 411114 3RCS*PT403A RCS Pressure Wide Range This transmitter is used to measure wide range RCS pressure. It is used to provide indication, alarm, input RHR valve interlocks, and input to the inadequate core cooling monitor. A redundant channel utilizing Foxboro transmitters is provided. The Rosemount transmitter channel was checked against the Foxboro transmitter channel using 0FIS data and no anomalies were observed. This transmitter was calibration checked in February 1989 and found

, to be in normal working order.

Model No. Serial No. Plant Tao No. Function 1153HD5 408197 Not Installed None 1153005 410157 Not Installed None 11540P4 414993 Not Installed None The above transmitters are not installed and therefore pose no safety con-cerns.

Millstone Unit No. 2 - Three Transmitters from Susoect Batches Reference (3) identified three transmitters from suspect batches which were used at Millstone Unit No. 2. The identification, function and failure effects of each of these are described below:

1 Model No. Serial No. Plant Tao No. Function ll53HD5PA 411943 RCS-LT-110Y Pressurizer Level l

e '

, i* U.S. Nuclear Regulatory Commission B13178/Page 6 April 13, 1989 This transmitter is one of two transmitters used to provide pressurizer level indication signals. The level signals are used for control and indication.

and they provide no reactor protection or engineered safeguards actuation functions. The selection of either of these two transmitters is made man-ually. Technical Specification sections 3.3.3.5 (Remote Shutdown Instrumen-tation) and 3.3.3.8 (Accident Monitoring) both require a minimum of one channel to be operable in modes 1, 2, and 3. Loss of both transmitters would require plant shutdown if not restored within seven days. The pressurizer i level signals are Type A variables per Regulatory Guide 1.97 and they are also l referenced in the plant Emergency Operating Procedures as one means of eval-uating RCS inventory conditions during POST-LOCA operation. This transmitter i was calibrated and response checked in February 1989 and found to be in normal working order.

Model No. Serial No. Plant Tao No. Function l

l 1153H05PA 411942 Not installed None 1153HD5PA 411944 Not installed None i l )

The above spare transmitters are not installed and therefore pose no safety j l concerns. !

i Millstone Unit No. 1 - Transmitters from Susoect Batches !

Millstone Unit No. I has not received transmitters from the batches in j i question.

Haddam Neck Plant - Transmitters from Susoect Batches !

l The Haddam Neck Plant has not received any transmitters ~ from batches in l question, and Rosemount transmitters are not used in any applications at the l plant.

~

, Summary of Transmitters from Susoect Batches Rosemount informed NNECO under 10CFR Part 21 that a potential failure could occur in Rosemount model 1153 and 1154 transmitters. The loss of sensor fluid failure mode may cause the transmitters to exhibit reduced performance prior to a detectable failure. Rosemount indicated that the reported failures all occurred in certain manufactured batches. Millstone Unit No. 3 received 13 transmitters from suspect batches and Millstone Unit No. 2 received 3 trans-i mitters. The functions and failure effects of these transmitters have been identified. In addition, all of the installed transmitters from these suspect batches were performance checked in February 1989 and found to be in normal working order.

n

.{ U.S. Nuclear Regulatory Commission B13178/Page?7

April 13, 1999 insoection of Rosemount'. Inc.

NNECO has responsibility under 10CFR50 Appendix B to assure that its suppliers of 1 safety-related components are complying with applicable requirements.

Because of our concerns on the Rosemount transmitters, we sent an experienced mechanical . engineer to the Rosemount facilities in February 1989 to review

, their transmitter sensor. manufacturing and inspection process, and to verify z,

improvements that have been made.

The inspection focused on the following specific concerns:

1) Rosemount's determination of the root cause of the failures.

2) The methodology used by Rosemount to restrict the problem to a subset of transmitters.

Rosemount has concluded, and we concur, that the root cause of failure is the lack of bond between the glass and metal cup within the transducer cell. The critical manufacturing operation is the furnace glass seal and its preparatory steps. In August of 1986, Rosemount undertook a process optimization program which, when completed in March of 1987, cut their reject rate from approx-imately 20 percent to 1-to-2 percent. This, in conjunction with increased testing and inspection criteria, is the basis for Rosemount's determination that' the problem will not exist with units manufactured after the first quarter of 1987.

The failures to date have been numerically small, the first documented failure occurring with a 1979 shipping date. The latest shipping date for a failed

' transmitter was in the first quarter of.1987. There were approximately 300 lots totaling just over 14,000 units produced between those dates. As of February 1989, Rosemount has had 84 reported failed units classified as " low oil" failures. Upon testing, 6 of those were found to be due to other causes.

Each of the 78 low oil. failures was traced back to its manufacturing lot and 20 lots' were classed as suspect. There are 1004 units in those 20 lots of which 16 units were supplied to NNEC0. Factors mitigating the conservative nature of this data are:

1. The failure is of the infant mortality type; the metal-to-glass bond delaminates upon'cooldown from the furnace temperature to ambient.

2. The overall failure rate is low.

l

3. The failures are due to small process variations that are lot dependent.

'Therefore, while ' both Rosemount and we believe that additional lots may be ,

added' to the suspect: list due to additional failure information, the total i number.of failures will remain low (in the range of one-percent or less of all j units produced). In addition, units in service are less likely to be found !'

failed due to . low oil as service life increases. Of the known loss of oil

U.S. Nuclear Regulatory Commission B13178/Page 8 April 13, 1989 failures, the shortest length of service time was a few days and the longest 30 months. The most frequent lengths of service before failure were 12,18 and 27 months. This data is derived from calibration or refuel outage dates and is not truly actual service time. The Millstone Unit No. 3 transmitters have now operated for approximately 36 months.

Additional Actions Taken to Provide Ooerability Confidence l

In addition to those transmitters specifically listed by serial number and identified by Rosemount as being of concern, we have reviewed all other Rosemount transmitters in safety-related applications. All safety-related Rosemount transmitters were evaluated by specific function to assess what, if any, further testing or analysis was appropriate. These fell into two primary categories, transmitters that for various reasons did not require further testing and/or analysis and transmitters that did require further testing and/or analysis. Further testing and analysis consisted of calibration checks, review of 0FIS data and/or review of transient analysis data as appropriate to detect any performance anomalies. To date, none of these other transmitters have exhibited symptoms associated with the loss of sensor fluid problem. Rosemount has stated in References 2 and 3 that industry data has shown that the transmitter failures are random with the highest probability of occurrence during the beginning of operation.

i NNECO has not discovered any additional transmitters that have experienced l this failure mode. We believe this supports the infant mortality conclusions I that were independently reached as a result of our inspection of Rosemount's )

facilities.

Future monitoring of our Rosemount transmitters includes a program at Millstone Unit No. 3 'which will (1) verify transmitter performance . utilizing 0FIS data and (2) instruct instrument technicians of symptoms to be alerted to, which would be indicative of degraded transmitters. ,

Conclusions on Failure Symotoms Our evaluations and analyses of Rosemount transmitters described in this letter have lead us to conclusions that any of the following symptoms may be an indicatinn of transmitter failure.

1. Slow drift in either direction on the order of 1/4% per month.

2. "One sided" noise from flow signals.

3. Slow response to a transient or inability to follow a transient.

4. Decrease in the RMS noise level. -

5. Deviation from redundant channels.

l

i '

U.S. Nuclear Regulatory Commission i B13178/Page 9 April 13, 1989 Transmitter calibration when exercised over the calibrated range is considered the best indicator of a loss of oil at the present time. If any of the following symptoms are observed during calibration, further evaluation is recommended.

1. Inability to respond over the entire range.

2. Slow response to either an increasing or decreasing hydraulic test pressure. (e.g. Response at either the high or low end of the calibrated range may be on the order of 2 minutes to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .)

3. Any drift of greater than 17. from original calibration.

We have shared the above conclusions with the industry via Nuclear Network notification on February 14, 1989 and we have factored these conclusions into our monitoring program.

Conclusion We have had a high level of involvement in the Rosemount transmitter issue.

Information which we have helped to identify and share with the vendor, the NRC Staff, and various industry representatives, is contributing to the under-standing and resolution of this issue. We have continued to ensure that operability and reporting requirements have also been met.

Our confidence in the continued operability of the Rosemount transmitters is

~

based on a number of factors including:

o Specific evidence to believe Rosemount has identified and also !

corrected the deficiencies in the manufacturing process.

o Specific evidence to believe Rosemount has identified and improved their inspection process to preclude defective units being shipped, o Infant mortality evidence supported by data collected by Rosemount and by NNECO. The mortality is such that the transmitters that are subject to failure can be expected to have already failed.

o Identification and review of all safety-related Rosemount trans-mitters used in our plants, principally to determine their function and potential failure effects.

o Specific operability verifications provided by transmitter response testing, calibration, review of recorded transmitter perforrance, use of high-speed data acquisition monitoring, and use of Offsite Facility Information System monitoring.

o A monitoring program that will selectively verify transmitter performance.

U.S. Nuclear Regulatory Commission 3

B13178/Page 10 April 13, 1989 This summary of our activities on the Rosemount transmitter issue is intended to further contribute to the understanding and resolution of this concern.

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY NORTHEAST NUCLEAR ENERGY COMPANY d 2d/

E.J.froczka //

Senior Vice President cc: W. T. Russell, Region I Administrator A. B. Wang, NRC Project Manager, Haddam Neck Plant J. T. Shedlosky, Senior Resident Inspector, Haddam Neck Plant M. L. Boyle, NRC Project Manager, Millstone Unit No. 1 G. S. Vissing, NRC Project Manager, Millstone Unit No. 2 D. H. Jaffe, NRC Project Manager, Millstone Unit No. 3 W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 i

e 4

F~

l t k

. i

.....~..

. *I:

% vNITED STATES NUCLEAR REGULATORY CCMMISSION 3

5 ,f MEGION I

'g, f 47s ALLENDALE ROAD

- - - - KING OF PMUS$1A. PENNSYLVANIA 194o6 JUN 2 81989 Docket / License: 50-423/NPF-49 Nortneast Nuclear Energy Company ATTN: Mr. Edwarc J. Mrotzka 3dCEIVCU Senior Vice President - Nuclear JUL 5 *tS89 Engineering and Operations Group P.O. Box 270 8tartford, Connecticut ec 06101-0270

[.J.tnN.ViCf

. Jg.pcf.ra,10:Pa3r3 W iiO M Gentlemen:

Subject:

Millstone 3 Routine Inspection 50-423/89-04 (4/5/89 - 5/15/89)

The enclosed report refers to the routine resident safety inspection conducted on April 5 througn May 15, 1989 at the Millstone Nuclear Power Station, Unit 3.

The results of the inspection are described in the NRC Region I Inspection Re-port enclosed with this letter and were discussed with Mr. C. H. Clement of s your staff at the conclusion of the inspection.

Report Section 8.0 describes further NRC review of Rosemount transmitters.

Your initiatives and leadership in evaluating transmitter failures and notify-ing the NRC of this safety issue are commendable. We remain very interested in .

your evaluation of the continued aperability of installed transmitters, and in '

your conclusions regarding failure probability and the adequacy of the present instrument calibration interval. Please address these matters in a written response to this inspection wi. thin 30 days of receiving this letter,, ' ;

i Y: r ::: : ration with us is appreciated.

I S rely, e

Edward C. Wenzinger, Acting hi f Projects Branch No. 1

, Division of Reactor Projects

Enclosure:

NRC Region ! Inspection Report 50-423/89-04 l

l 1

D ,

f Attachment to 000153

( E 5 &liO &&i5' Ef */} }

e

.' . Northeast Nuclear Energy Comoany 2 JUN 2 8 l989 I

cc w/ enc 1:

l W._ D. Romoerg, Vice President, Nuclear Operations R. M. Kacich, Manager, Generation Facilities Licensing l

D. O. Norcoutst, Director of Quality Services S. E. Scace, Station Superintencent C. H. Clement, MP3 Superintencent D. B. Miller, Station Superintancent, Haddam Neck Public Document Room (PDR)

Local Public Document Room (LPDR)

Nuclear Safety Information Center (NSIC)

NRC Senior Resident Inspector State of Connecticut 1

i e

e 9

e !

e

~'

k..

1 1

U.S. NUCLEAR REGULATORY COMMISSION REGION I Report No. 50-423/89-04 Docket No. 50-423 License No. Np F-49 Licensee: Northeast Nuclear Eneroy Company P.O. Box 270 Hartfore. CT 06101-0270 Facility Name: Millstone Nuclear power Station. Unit 3 Inspection At: Waterford. Connecticut Inspection Conducted: April 5 - May 15, 1989 Reporting Inspector: W. J. Raymond, Senior Resident Inspector Inspectors: W. J. Raymond, Senior Resident Inspector G. S. Barber, Resident Inspector Approved by: bbb <

E. C. McCane, Chief, Reactor Projects Section IB bhM Date Inspection Summary: Inspection on 4/5/89 - 5/15/89 Areas Inspected: Routine onsite inspection (112 hours0.0013 days 0.0311 hours 1.851852e-4 weeks 4.2616e-5 months ) of plant operations, previous inspection findings, Plant Incident Reports, reactor scrams on May 6 and May ll, a shutdown to repair high unidentified leakage on April 11, physi-cal security, repairs to service water piping, allegations and the MAP 4.16 l Allegations Resolution Program, Rosemount transmitters, maintenance, and sur-veillance.

Results: No violations.were identified. Reviews of plant operational status during operations and during the refueling shutdown identified no unsafe con-

' ditions. Further NRC review is warranted on whether additional licensee docu-me'ntation of their program for temporary repairs to code class piping is re-quired to NRR (Detail 7.0). Rosemount transmitter failures due to oil loss were found to be adequately addressed; additional information will be needed to follow the licensee's actions on this matter.

e

+#

w * * *

[

TABLE OF CONTENTS PAGE 1.0 Persons Contacted.................................................... 1 2.0 Summary of Facility Activities....................................... 1 3.0 Status of Previous Inspection Findings............................... 2 !

1 3.1 (Closed) IFI 85-09-01, Inspector Findings Regardino the Modified / Amended Security Plan................ ............... 2 3.2 (Closed) UNR 87-24-01, Issue Amendment to Allow Testing of Containment Overcurrent Devices to NEMA Criteria. . . . . . . . . . . . . . 2 3.3 (Closed) IFI 85 . Review Solid Radwaste System and FSAR Amendment Reflecting As-Built 0esign.......................... 2 3.4 (Closed) IFI 85-56-03, Review Dewatering Procedure and Fill-Head Venting Modifications......................................... 3 3.5 (Closed) IFI 85-56-04 Review Radwaste Procedures for Reference to New Dewatering and Solidi fication Processes. . . . . . . . . . . . . . . . 3 4.0 Plant Operational Status Reviews..................................... 3 5.0 Excessive RCS Unidentified Leakage................................... 4 6.0 Followup of Plant Trips.............................................. 5 6.1 Reactor / Turbine Trip due to Low Condenser Vacuum................ 5 6.2 Reactor / Turbine Trip due to Control Rod Testing................. 5 i.o non-Coce Re pai rs to Cods Cl a s s Sy stems . . . . . . . . . . . . . . . . . . . . . . . . . .'. . . . . 7 1

8.0 Followup on Rosemount 1153 and 1154 Transmitters..................... 9 90 Followup on Allegation RI-89-A-38.................................... 14 10.0 Maintenance.................., ...................................... 15 l 11.0 Surve111ance......................................................... 16 12.0 Management Meetings.................................................. 16 em os a,o e o e manone een e o

DETAILS 1.0 Persens Centacted Inspection findings were discussed periodically with the supervisory and management personnel identified below:

S. Scace, Station Superintendent C. Clement, Unit Superintendent, Unit 3 M. Gentry, Operations Supervisor R. Rothgeb, Maintenance Supervisor K. Burton, Staff Assistant to Unit Superintendent J. Harris, Engineering Supervisor D. McDaniel, Reactor Engineer R. Satchatello, Health Physics Supervisor M. Pearson, Operations Assistant 2.0 Summary of Facility Activities The plant began the inspection' period at full power and operated until 7:20 a.m., April 8 when a power reduction to 90% was necessary to perform condenser thermal backwashes. Power was returned to 100% by 6:55 p.m.

that day.

A power reduction was commenced at 11:45 p.m. April 11 when unidentified leakage (See Detail 5.0) increased from 0.5 gpm to 2.5 gps. The leak was confirmed by increases in containment radiation levels and chemistry samples. The leak was from a cracked weld at a letdown line valve. A plant shutdown was completed at 10:00 a.m., April 11 and cooldown was com-plate at 10:48 p.m., April 12. A new valve was welded in place and estab-lishment of containment vacuum began at 3:05 p.m., April 14 Heatup began with Mode 3 being reached at 10:48 p.m. April 14. Reactor startup began with criticality occurring at 2:11 p.m., April 15. Full power was achieved at 8:38 p.m., April 17.

The plant continued to operate at full power until 8:10 a.m., May 6 when a

manual reactor /tur61ne trip (Detail 6.1) was initiated due to the loss of two circulating water pumps due to seaweed blockage on the intake screens.

The storm subsided and the reactor was started up and made critical at 10:39 p.m., May 6. 'The plant returned to full power operations.

The reactor automatically tripped on high negative flux rate from full t I

power at 3:14 p.m. on May 11 (Detail 6.2). The trip occurred as I&C per- !

sonnel were installing test equipment in preparation for control rod scram i time testing.. The plant was kept shutdown to begin the refueling outage. j l

l i

E

2 Refueling Outage #2 is seneculed to last 52 days and will incluce the foi-lowing major activities: installation of the ATVS mitigation system; ser-vice water system insoection anc repairs; ISI weld insoections; refueling; installation of reactor vessel level monitoring system for mid-loop opera-tions; and, a containment integrated leak rate test.

3.0 Status of previous Inseection Findinos (92701) 3.1 (Closec) IFI ES-09-01. Insoector Findinos Recardino the Mocifiec/Amencea Security Plan The inspector reviewed the results of inspection 50-423/85-64 and noted that the 11spection reviewed whether the Millstone 3 Physical l Prote'etion Program, including personnel, equipment, systems and facilities, was being effectively integrated into the proposed com-bined security program for the Millstone site. The review also in-cluded a special evaluation of the security force training program to determine the ability of security force personnel to carry out their duties and responsibilities by observing licensee acministered ex-aminations of a statistically selected sample of security force per-sonnel, in the tasks in which they were cualified, to obtain results at a confidence level of 955. Additionally, the review included in-

' tarviews of key members of the security organization and project engi-neering staff responsible for the installation and testing of secur-ity systems and equipment. The review indicated that the Unit 3

, security program conformed to NRC requirements in the areas examined and was being effectively integrated into the Millstone site program.

In addition, recent reviews and inspection have concluded that the Millstone 3 Modified Amended security plan was adequately integrated into site security. Any violations and/or deviations have been ad-dressed and effective corrective action was noted. Based on the lic-ensee perfomance to date, this item is closed.

3.2 (Closed) UNR 87-24-01. Issue Amendment to Allow Testino of Containment Overcurrent Devices to NEMA Criter13 Amendment 13, dated January 20, 1988, was issued to revise the in-

, stantaneous overcurrent trip setting from plus or minus 205 to plus 405 minus 255 per the NEMP A8-2 criteria. Past testing conformed to the espanded tolerance. This item is closed.

3.3 (Closed) IFI 85-56-02. Review Solid Radwaste System and FSAR Amenament Reflectino As-Built Desion FSAR Section 11.4.1 was revised in 1987 to reflect changes in the i waste solidification design. This revision replaced the in-site system with mobile solidification equipment. Existing system con-nections are compatible with the mobile solidification equipment.

This item is closed.

1

3 3.4 (Closed) IFI B5-56-03. Review Dewaterine Procecure and Fill-Head Venting Mecifications OM-43 replaces the OM-38 regarding operation of the NuPac Resin Dry-ing/ Dewatering System. Procedure OM-43 contains the necessary in-structions and emergency actions relative to system alarms. In ad-dition, the procedure requires an operational check of the system interlocks prior to initiating a resin transfer. Also, the exhaust from the resin drying equipment is dt. charged through a HEPA venti-lation unit. Therefore, the plant's vent system would not be re-quired to handle any contaminated exhaust ventilation from the equip-ment. These procedure changes were responsive to the inspector's concerns. This item is closed.

3.5 (Closed) IFI 85-56-04, Review Radwaste Peocedures for Reference to New Dewaterino anc Solicification Processes OP 3338A, Radioact1.ve Solid Waste and OM-43, NuPac Dewatering System describe the proper operation of their respective systems. In the event that waste solidification is necessary, a vendor's Process Con-trol Program will be reviewed and 50RC approved in order to accomp-lish the solidification. The solidification will be done with a mobile solidification system provided by a vendor. This activity is routinely inspected by the core inspection program. Future deft- !

ciencies noted will be addressed by the licensee during these inspec- l tions. This item is closed.

4.0 plant Operational Status Reviews (71707)

The inspector reviewed plant operations from the control room and reviewed the eperational status of plant safety systems. Actions taken to' meet technical specification requirements when equipment was inoperable were reviewed to verify the limiting conditions for operations were met.

Plant logs and control room indicators were reviewed to identify changes in plant operational status since the last review and to verify that changes in the status of plant eq11pment was properly communicated in the logs and records, dentrol room instruments were observed for correlation !

between channels, proper functioning and conformance with technical spect- l fications. Alarm conditions in effect were reviewed with control room operators to verify proper response to off-normal conditions and to verify operators were knowledgeable of plant status. Operators were found to be cognizant of control room indications and plant status. Control room man- '

- ning and shift staffing were reviewed and compared to technical specif t-

. cation requirements. No inadequacies were identified.

j 1

1 l

L l

5.0 Excessive Unidentified RCS Leakace (937021 j l

At 11:45 p.m., April 11, with the plant at full power, a shutdown com- l mentec when RCS unidentifiec leak rate reached 1.5 gpm (Technical Speci- 4 fication (TS) limit is 1.0 gpm). The leakage was caused during Engineerea l Safety Features (ESF) slave relay surveillance testing earlier in the day (4:00 p.m. to 6:00 p.m.). The RWST suction valves to the charging pumps were being openec for the test when operators closed down on the charging )

flow control valve to limit the amount of 2300 ppa boron that was injected I to the RCS. The charging flow reduction was necessary to limit the post surveillance dilution required to return boron concentration to its and of cycle concentration. Charging flow was reduced rapidly, causing a de-crease in heat transfer across the regenerative heat exchanger. This ac-tion caused flashing to occur downstream of the letdown orifices due to higher inlet temperatures. The subsequent manual increase in charging flow collapsed this bubble causing a water hammer to lift and reseat the letdown relief valve. An unidentified leak began at this point.

The increase in leak rate manifested itself to the operator between 8:00 p.m. and 10:00 p.m. as increases were noted in containment radiation levels and sump pump rates. The shift supervisor ordered the sump sampled l for baron and activity. The sample boron concentration and activity was reported back at 11:30 p.m. representative of RCS liquid, confirming the leak indications. An Unusual Event was declared at 11:48 p.m. April 11 and the NRC was notified via the ENS at 12:07 a.m., April 12. A power decrease was conducted to reduce inside containment radiation levels in preparation for containment entry. The entry team received their medical exam and was briefed on potential leak locations. The containment entry team located the leak on a cracked weld on a leakage monitoring valve (3 CMS *V995) letdown line at 3:20 a.m., April 12. The weld connected the valve body to a 3/4 inch vent line. Spray was observed in a 270 degree are coming from the weld. The leak is isolable by the inboard containment isolation valve and the orifice isolation valves. Leak repair required cold shutdown. Reactor shutdown was completed at 9:11 a.m., April 12 and the licensee continued r;1 ant cooldown to effect repairs. The leak was isolated during the cooldown, after. Residual Heat Removal (RHR) cooling was initiated. The Unusual Event was terminated at 8:05 p.m., April 12.

The inspector reviewed licensee actions during this event. Licensee ac-tions were generally effective and showed due regard for safety. Tech-nical Specification time limits were met. The plant was cooled down with-out incident. '

The inspector questioned the need for the sump activity and boron sample

. to confirm the leak. The licensee had two supporting indications (con-tainment radiation levels and sump pump rates) that confirmed the initial leak rate. Sampling the sump resulted in an unnecessary delay in plant .

shutdown. Licensee use of real time supporting indications for offneraal conditions will be reviewed in future inspections.

l I

I

~

Licensee use of the RCS leakage abnormal operating procedure (AOP) was lacking. Neither the swing snift or midnight shift shift supervisor (55) used the AOP. They felt it was unnecessary since they knew wnat they I wanted to do. The inspector emonasized the need to follow procedures and emonasized that operators receive the wisdom of all the memoers of Plant Operations Review Committee when they use and follow proceduras. Prece-dural acherence curing off normal conditions will be reviewed during future inspections. This item will remain unresolved pending further NRC review of licensee procedural use and adherence (UNR 89-04-01).

6.0 Followuo of Plant Trios (93702) 6.1 Reactor / Turbine Trio due to Low Condenser Vacuum J l

On May 6, at 8:10 a.m., while at 905 power, a manual reactor / turbine l trip was initiated due to lowering main condenser vacuum. Power had previously been lowered to complete a condenser backflush. The "A" and "B" main circQ1ating water (CV) pumps had been automatically tripped at 8:08 and 8:09 a.m., respectively due to high suction dif-farential pressure (DP). The high OP condition was the result of an early morning storm. Excessive blockage on the intake screens re-sulted in the inability of the screenwash system to continue to clean the screens. The loss of both CW pumps on high OP resulted in a low vacuum in the "C" condenser shell which forced the manual reactor trip. All safety systems responded appropriately to the trip. Opera-tors stabilized the plant in hot shutdown. The trip was reported in accordance with 10 CFR 50.72(b)(2)i1 at 8:3G a.m., May 6.

The intake screens were cleaned subsequent to the trip and the plant was restarted after the stormy conditions subsided. The approach to criticality was initially begun at 1:00 p.m., but was aborted when the 1/M plot showed that criticality was predicted beyond the cap-abilities of control bank D. The licensee had anticipated that rod worth could overcome the effects of Xenon buildup after the trip.

The ECP was recalculated for later that evening with the approach to criticality commencing at 9:58 p.m., May 6. The reactor was made critical and Mode 1 was entered at 12:38 a.m., May 7. 1 The inspector reviewed the sequence-of-events printout and the opera-ter questionna4res. Items of minor significance were questioned and satisfactorily addressed by the licensee. No inadequacies were

. noted.

6.2 Reactor / Turbine Trio due to Control Rod Testino The reactor scrammed automatically on negative flux rate trip from 100% full power at 3:14 p.m. on 5/11. The seras occurred when !&C personnel turned off a " rod drop monitor computer" which caused

F I

6 l'

i control recs to insert. The rod test equipment was set up in pre-paration for scram time testing scheduled during the plant snutcown l for the refueling outage on May 12.

Operators stabilized the plant in het shutdown at 560F and 2250 psig.

Plant response to the transient was normal except for minor proelems with a sticky pressurizer spray valve; problems re-opening the main j

feedwater control valve to the A steam generator in the post-trip recovery phase; and problems with the auxiliary feedwater flow con-l trol valve to the D steam generator.

The steam driven auxiliary feedwater pump was out of service for l

' overspeed testing at the time of the trip; a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action statement per TS 3.7.1.2 was in effect. Steam generator level centrol was I maintained with main feedwater system and the two motor driven aux 111ary feedwater pumps as required. No ESF systems were required to operate. The resident inspector responded to the control room and verified stable plant conditions. Inspector review of main control l

board indications verified the plant responded as expected (except as noted above) for a reactor / turbine trip, and that reactor operator responses were proper. No inadequacies were noted.

The licensee reported the scram to the NRC Duty Officer as. required by 10 CFR 50.72(b)(2)(11) at 3:39 p.m. May 11. The licensee kept the plant shutdown to begin the refueling outage since fuel exposure was within the required burnup window.

The negative rate trip occurred when two or more control rods in-sorted while !&C technicians were installing the control rod timing computer. I&C personnel had entered the containment to install the i equipment and test the communication links between the remote rod l position signal units and the computer. The communication link tested I satisfactory per procedure SP 3451N21. The technician also connected the computer' to the control rod logic cabinets (3RDS-RHK1HC)~ per the same procedure. The trip occurred as the technician powered down the computer. This action caused a spurious signal in the logic cabinet that resulted in rod insertion. This same test set up had apparently been performed in the past without incident. The exact mechanism on how the text equipment caused the rod insertion was still under lic-ensee investigation at the end of the inspection period.

1.icensee review of equipment performance following the trip noted the following:

! (1). A misaligned limit switch on main feedwater isolation valve FWS*CTV41A for the ' A' steam generator prevented re-opening the ,

valve on demand by the reactor operator following the scram. I The misaligned NAMC0 switch failed to indicate the valve was

" closed" and prevented satisfying the reset permissive in the

~

valve isolation circuit (reference S&W Drawing ESK-7JN). The L_.

r 7

valve was opened by the operators with the assistance of main-tenance personnel at 4:52 p.m. Licensee corrective actions in-

! cluded plans to review switch mounting on other valves and to improve the switch mounting bolt arrangement.

(ii) Following the scram, the auxiliary feedwater flow control valve (FWA*FCV3101) to the 80' steam generator closed in response to l

demand by the reactor operator to control level and prevent ex-cessive cooldown rates. The valve failed to reopen and genera-tor level was controlled using main feedwater. Licensee review determined that the valve controller failed on the main control board when a drive cord broke between the thumbwheel and the potentiometer that provides the position demand signal. Cor-rective actions included review of actions necessary to peri-odically inspect the cords on other controllers.

(iii) Following the scram, the reactor operator noted reactor coolant system pressure decreased to about 2000 psig due to inadvertent operation of pressurizer spray valve RCS PK 4558. Operator ac-tion to place the controller in manual to cycle the valve, and operation of the pressurizar heaters limited the RCS pressure decrease. The valve controller on the main board was trouble reported on 10/31/88 due to suspected problems with the con-troller " sticking" sometimes. The loop 2 cold leg pressurizer spray valve, RCS PK 455C, had been out of service since 2/25/89 due to a controller problem. Licensee corrective actions in-cluded replacement of both controllers during the refueling out-age.

Inspection of the event included interviews witti operators and man-agement personnel, review of control room indications, and a review of the seguence-of-events printout, post trip data available from the plant comouter, and the licensee's post-trip documentation (PIR 66-89 EPIP 4112-3 and OPS Fors 3263). No anomalies were noted in plant system performance. Licensee review and evaluation of the event was proper. No inadequacies were identified.

The resident inspector will follow the licensee root crise evalu-ations, equipment repairs, LER summary and corrective actions on a subsequent routine inspection. No inadequacies were identified.

7.0 Non-Code Repairs to Code Class Systems (71707)

NRC Inspection Report 88-24 describes previous NRC review of licensee ac-tions to identify and correct leaks in ASME Code Class 3 piping in the service water system. Previous inspections found licensee evaluations of known leaks were technically acceptable to assure piping system integrity during interim periods of operation. The licensee used temporary repairs to limit the leakage from the system untti the plant shuts down for an i

outage.

l

7 .-

i - 3 I

Service Water leaks I

One proelem reviewed previously and aedresseo again during this inspection period, concerned a leak in the service water outlet from the "A" reactor

! plant component cooling water (RPCCW) heat exchanger, just down stream of l valve 35WP-V35. A through wall defect in the copper-nickle clad, carbon steel pipe was founc to be highly localized anc had dimensions that were less than the critical size that would jeopardize structural integrity of the 18 inch diameter pipe. The licensee installed a leak limiting device (a " soft-patch" - wood plug held in place by metal and or weobed bancs) to

! reduce water spillage to the floor area. The licensee removed the "A" l RPCCW heat exenange from service on April 18 to further evaluate the de -

! feet after noting leakage had increased to about 5 to 10 gpm on April 17.

Further licensee investigation found that although the size of the defect had increased slightly, the conclusions from the prior technical indica-tion remained valid. Actions were taken to reduce the letkage by im-proving the patch and,then returning the system to service. The inspector reviewed the technical evaluation with engineering personnel and identi-fied no inadequacies.

License Recuirements l

Further inspector review on April 18-19 of the regulatory requirements for j construction, inspection and repair of safety class piping raised a ques-tion whether the plant was operating within the licensing basis for the interim period of operation under a " temporary repair" (soft patch). The ASME code of reference for MP3 specified in 10 CFR 50.55a(g)(3)(ii) is ASME III and ASME Section XI. Final Safety Analysis Report Table 3.2-1 i establishes the correlation between ASME code classification and safety l classification for piping systems, and classifies the RBCCW and . Service l Water Systems as Safety Class 3. Technical Specification 3.7.3 provides

! the operability requirements for the RBC CW system. Technical Specifica-tions 4.0.3 and 4.0.5 requires that safety class systems be tested per ASME Section XI and further states that failure to meet the TS surveil-lance reoutrements constitutes a failure to meet the limiting condition for operation.

l l Section IWD 2600 and Table IWD 2500-1 of ASME Section VI applies to safety l class 3 piping systems. The code requires that safety class 3 piping be l subjected to periodic visual examination with an acceptance standard that no leakage is permitted in the pressure retaining boundary. The ASME Sec-tion XI code recognizes approved repair methods for code class piping, but does not recognize temporary repairs or " soft patches." The use of repair methods on code class piping that is not recognized by the ASME Code con-4 stitutes a condition contrary to the requirements of 10 CFR 50.55(a)(g) that is outside.the NRR approved licensing basis for the plant.

en

_1

F 9

Safety Stonificance This matter was discussed with licensee technical staff and management on April 19 and during a conference call between licensee and NRC Region and NRR staff on April 20, 1989. The licensee further described his program to monitor and address degradation in both small- and large-core piping in the service water system in a letter dated April 28, 1989. The licensee stated that the soft patches are not ASME code repairs, but are interim non weld repairs intended to be used until an outage of sufficient dura-tion occurs to' allow permanent code repairs. The licensee's engineering evaluations assured structural integrity criteria of the ASME code was not met even with the defects, and that operability of equipment in the area of the leak would not be compromised by flooding, assuming no credit is taken for the soft patch. The interim repair is considered a maintenance activity that limits the amount of water leakage from the piping. NRC l

staff review of the licensee's position identified no inadequacies with i the technical evaluations or the program to address service water system leaks.

! Based on the above, the inspector identified no safety concerns 3 systetts. continued plant operation with temporary repairs on safety class regarding This matter requires further review by NRC management to determine what further actions, if any, are required by the licensee. An NRC staff post-tion to address this issue generically for the industry is pending. This item is unresolved pending completion of the review by the NRC (UNR 89-04-02).

8.0 Followup on Rosemount 1153 and 1154 Transmitters (92700)

Peeblem Summary 1

The inspector reviewed ongoing licensee actions to. address suspect Rose-mount transmitters. The inspection review incluc3d a meeting with cor-porate engineering personnel on March 30 and a meeting between NRC tech-nical staff in headquarters with industry groups on April 13, 1989. The licensee updated his initial 10 CFR Part 21 report by providing supple-l mental information in a letter dated April 13, 1989.-

The inspection reviews, along with input on April 3 from a licensee employee with safety'cc.::rrns about the issue, identified new information .

l to the NRC about the number of failures, the failure mode, the relatively high failure probability, and the significance of having a failed instru-ment that was not detectable. The employee's safety concerns involved addressal of the issue at other nuclear plants since actions to address the issue at Millstone station were either complete or in progress. Based on the additional information, the NRC staff took action to address the issue to plant operators (see below). NRC Inspection Reports 50-423/88-05 and 49-02 describe NRC review of actions at Millstone.

l

.. 10

)

Rosemount 1153 and 1154 pressure sensing units are used extensively in the safety related, environmentally qualified applications in the nuclear in-dustry. Of 106 units in use at Millstone 3 (MP3), five failed in service in 1987. Subsequent reviews by'the licensee concluded a potentially sig-nificant safety hazard (SSH) existed due to the failure moce and the wide-spread use of the instruments.

The results of the SSH were reported to the NRC in a March 1988 10 CFR 21 report. Subsequent review by the vendor, Rosemount, attributed the cause of the defects to the manufacturing process. Rosemount provided update information to users in a February 1989 letter that identified suspect units on a site specific basis, and recommended that utilities review the units for safety impact at their sites.

The problem was further studied by the Electric Power Research Institute using Millstone 3 data supplied from the Off Site Information System (OFIS). The failure mode was characterized as follows:

The units fail from a loss of oil in tho sensing chamber. That re-suits in an inability to respond over the ful1~ span, an inability to respond in the increasing pressure direction, and a loss of dynamic response capability.

The deterioration is gradual but has an " infant mortality" aspect. ,

Failue can occur within 30 to 36 months after being placed into ser- I vice in a high pressure application (greater than 1000 psi). ]

The sensors can be significantly degraded in place (experience a loss of safety function) prior to the onset of detectable (observed) fail- i ure. A channel that looks operable to an operator during a panel !

check can be " failed-as-is" and incapable of responding in the up-scale direction. ;

The detection of degraded units and identifying failures attributable to loss of oil is difficult. Identification relies upon accurate root cause l determinations of failures at plant sites. It is probable that, as lic-

. ensees review the failure data base for their plants, oil loss failures i may exist but may not be identifiable. The ability to detect the failures is only as good as the level of detail with which the failure was de-scribed when entered into the data base (usually as part of the mainten-ance work ort'er process), and only as good as the root cause analysis done.

Furthsr. revu m of the calibration record for the plant may give a false security that tne failure is not present, unless the review is done based on thorough guidance. Since the degradation is a slow process, a unit undergoing calibration during the early stages of failure may show only a slight amount of drift and this drift could be calibrated out during a test, or successively over several tests.

11 The satisfactory completion of response time testing has not been adeguate l

to assure installed units are acceptable. Out of the 78 identified fail-ures, only 1 or 2 were identified during response time testing. A problem in meeting the response time acceptance criteria will show up only when the unit is on the verge of obvious failure or already significantly de-graded. The problem can be seen in its incipient stages curing tests when units are subjected to their full span of pressures. This includes cali-brations and response time tests.

Licensee engineering reviews with.the vendor determined that the first documented 1153/1154 failure (due to loss of oil) occurred in a unit with a 1979 shipping date, and the latest shipping date for a failed unit was in 1987. Just over 14,000 units were produced in that time in 300 lots.

Of about 85 failures suspected to be from loss of oil, Rosemount found that 78 were due to low oil. Each of these failures were traced to its manufactured lot and 20 lots were identified as suspect. There were 1004 units in these 20 lots, and 16 of these units were supplied to Millstone.

The list of suspect batches for units identified in the February 89 Part 21 letter was developed from transmitters examined by Rosemount and known to have failed from loss of oil. Since the vendor does not have the cap-ability to handle contaminated units, all failures have not been examined.

It is estimated that the total number of failures from loss of oil may involve hundreds of units industry-wide.

In correspondence with the licensee dated December 8, Rosemount stated there have been no reported failures in sensors built in the last 3 years (1986 - 1988). The vendor reportedly made improvements in the process in that period to address'the loss of oil problem. The vendor reportedly has more recently, through a combination of improvements in the manufacturing i process and in-process testing criteria, improved the manufacturing reject rate. Although the manufacturing process for 1151 & 1152s transmitters is the same as that for 1153/1154s, the former units are used mostly in con-trols applications and non-EEQ safety appiteations, and reportedly have not experienced the failure mode.

Estimated Failure probabilities The licensee estimated the failure rate by relating the 78 loss of oil failures to the full manufacturing base produced from 1979 to 1987. The 78 failures represented 0.05655 of that manufacturing base. Using a round-up value of 1% defects in the manufacturing base, the failure rate for 1 year of service was then (0.01/8760 =) 1.1 x 10-6/ hour. This fail-ure rate compares favorably (1/30th) to the rate assumed for random fail-ures in probabilistic safety analyses of 3.85 x 10-5/hr according to the licensee.

Inspector review concluded that the above predicts 0.014 failures in an 18 month operating cycle, instead of the 5 failures that actually occurred at Millstone 3. Also, of the 106 units installed at Millstone 3, 16 were 1

1

12 from the suspect lots. Five of these failed in service over the period from March - Novemeer 1987. The inspector therefore concluded that the licensee's estimate (1.1 x 10 '/ hour) was overly optimistic.

The inspector reviewed the methodology employed by Westinghouse in WCAP-10271-P-A, Evaluation of Surveillance Frequencies and Out Of Service Times For Reactor Protection Instrumentation System. In that methodology, the detectability of failures is an important factor that is accounted for explicitly in the calculations. Undstectable failures are defined by IEEE 379 as failures that cannot be detected by periodic testing or cannot be detected by alarm or anomalous indicatio.1s.

The Millstone 3 studies show that the Rosemount 1153 transmitters can be significantly degraded in place for months (incapable of providing a trip function in the upscale direction), with this condition not detectable by any alarm or anomalous , indication. The condition might be detected by periodic test, but the calibration surveillance interval is 18 months (13,140 hr) and the mean time to detect failures is 13,140/2 = 6570 hours0.076 days 1.825 hours 0.0109 weeks 0.0025 months . -

On the other hand, the WCAP methodology assumes a failed instrument will be detected within two operating shifts (i.e.,16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months ), a significant difference.

While the WCAP failure rate is 2.8 x 10 8/hr and the licensee estimated a 1% defect (1.1 x 10 */hr), actual experience at Millstone 3 appears to be much worse. The WCAP assumed detection interval was 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , at Millstone 3 it is 6570.

The WCAP concluded that the failure probability (P) using 1=2.8 x 10 8/hr and T=16 hour detection interval was acceptable, where:

.P = AT/2 = 2.2 x 10 '

Using T=13,140 hours0.00162 days 0.0389 hours 2.314815e-4 weeks 5.327e-5 months and the licensee's assumed failure rate (1.1 x 10 '/hr), the failure probability results would be approximately 7 x 10 3 Considering the 5 failures experienced at Millstone 3 in 18 months, the failure rate would be 4 x 10 */hr. This is a much worse failure rate than

. 1.1 x 10 '/hr. .

This matter will be reviewed further with the licensee to determine whether a more definitive failure probability can be obtained.

Further NRC Action /Followuo The NRC technical staff issued an Information Notice (IN 89-42) on April 21, 1989 advising the industry of the information available to the staff and requesting other utilities to review the matter for applicability for their plants. The present action plan for Millstone is to continue review of the 1153s and 1154s in safety-related applications for acceptable per-formance. The review includes testing as necessary to see if there is l

i

, 13 !

l evidence of degraced performance of the type cescribed in the INPO Sig-nificant Event Notice (SEN) 57; 1.e., sluggish response, slow crift of 1/4 percent or more, recucec noise in signal or change in normal system signal fluctuations, inability to responc over the entire operating range, etc.

Suspect units would be further tested to confirm operacility.

Wholesale changeout of susoect units that have yet to be proven inocerable was considerec by the licensee out is not considerec, by the licensee, to ;

be the best course of action at this time. Until the full scope of the suspect lots is cetermined there is a chance that new replacement units will also be susceptible to loss of oil failure. Further, unless addi- ;

tional data proves the " infant mortality" period invalid, a good confir- I mation that a given unit is not susceptible to the failure comes from the ,

service time in high pressure applications without degradation. The best i course of action appears to be to test installed units to detect signs of degracation. Test criteria from Rosemount are pending to assist this in- l' dustry effort. ,

The licensee noted that preliminary Rosemount findings have suggested that the cause for tne defects was related to a design change replacing an elastomer 0-ring with a metal one to qualify the transmitter-for harsh accident environments. The metal 0-ring resulted in increased stress being applied to the glass sensing chameer. That caused cracking of the silica and eventual leakage. The 0-ring was installed during final as-sembly after intermediate pressure tests of the sensing chamber were com- !

plated; thus, the assembleo unit could be shipped with the incipient fail- !

ure mode undetected. Rosemount review is in progress to recoassend testing !

for the industry to better cetect degraded units.

The inspector reviewed the bases for the licensee's conclusions regarding j Rosemount transmitters and identified no inadequacies. Licensee'initi- ;

atives to further review this issue included: testing on May 10 using the j Great Hack Road training facility labs to record instrument pressure drop -

versus time to measure the rath of oil loss versus operating pressure; l and use of the mockup flow loop to characterize the dynamic response characteristics of a degraded unit. Licensee actions during the refueling

~

outage to further. evaluate installed Rosemount transmitters will be re-viewed during subsequent routine inspections. ,

Previous NRC resident inspection (IR 89-02) reviewed licensea actions to assure operability of Rosemount transmitters in safety-relat A applica-tions at Millstone. These actions included verification of operability through response testing, calibrations, review of 0FIS data, and review of channel perforwance during transients. Licensee actions were assessed as

. thorough in addressing the issue at Millstone 3. .

v ~.2

.s

4

. 14 9.0 Fo11ewuo on A11ecations (40500)

Followuo of a Soecific Safety Issue (RI-89-A-38)

A licensee employee contacted the inspector on April 5 to allege harrass-ment by a supervisor for bringing technical concerns to the NRC (reference Section 8.0 above concerning Rosemount transmitters).

The employee stated the harrassment occurred after presenting differing views to the NRC inspector during an inspection meeting at corporate engi-neering on March 30. The employee's supervisor was not in attendance dur-ing the meeting. Upon his return, the supervisor allegedly criticized the employee's performance at the meeting and restricted his further involve-ment with the problem. The employee stated it was inappropriate for him to be removed from the project due to his particular expertise in the issue.

The March 30 meeting was held at the inspector' request as part of the NRC inspection of the licensee's technical resolution of the Rosemount trans-mitter issue. During the meeting, the employee expressed views on the issue that differed from the general engineering consensus regarding the significance of the Rosemount problem, and in particular, the significance of the failure rate. Review of the matter by the NRC then concluded that the failure rate was not acceptable (see Detail 8.0, preceding) and the employee's safety concerns were substantiated.

Inspector observations during the March 30 meeting were that the es-playee's actions and statements were appropriate and beneficial to achiev-ing thorough examination of all facets of the technical issue. The in-scoctor noted further that the employee was especially capable in dealing with this issue due to his extensive involvement and study of the issue on an industry wide basis.

The inspector informed the employee of the need to work with his chain of command and his rights to pursue the discrimination concerns with the Oe-partment of Labor s The employee pursued the issue through tb NU Alle-gation Resolution Program. The employee contacted the Allegations Program coordinator and was referred to an outside consultant (LR$ Associates) for followup of the technical issue. The administrative concerns were re-ferred to licensee management.

The inspector reviewed licensee actions to resolve this issue. The tech-nical and administrative issues were resolved to the employee's satisfac-tion. The licensee concluded that the employee's concerns were valid and that his continued participation in this matter was appropriate. Based on further NRC discussion with the employee, the licensee's allegations program was effective in resolving this instance of an employee-identified safety concern.

1 15 Review of A11ecatiens Precram imolementation in General The inspector met with the NU Allegations Program Coordinator to review !

the history of issues resolved by the program in its first year since im-plementation in June 1988. The licensee stated that, while several in-dividuals had initiated contact to resolve concerns, the only case in-tended to be addressed by MAP 4.16 (Millstone Employee Allegation Resolu-tion Program) was the one summarized above. The inspector reviewed with the licensee the specifics of the worker concerns in two other instances.

The concerns appeared to have been properly resolved. -

The licensee stated that a new corporate procedure, NEO 2.15, was issued to incorporate the map 4.16 guidance on handliag resolution of employee-safety concerns. The licensee plans to eventually have NEO 2.15 supersede MAP 4.16. The licensee s assessment was that, while use of the program 8

(in terms of the number of cases at Millstone and Connecticut Yankee) was low, the program was effective in resolving legitimate safety concerns.

The licensee intends to continue with the present program.

Results of a recent licensee survey of employee views were being tallied by the licensee. The licensee stated the preliminary conclusions con-firmed previous perceptions: workers have a high degree of personal regard )

for quality and safety in the performance of their jobs; and most workers do not feel the need to use the MAP 4.16 program since concerns can be worked out at the level of the immediate supervisor. The inspector noted the licensee's comments.

Routine NRC review of the licensee's programs for resolving employee safety concerns will continue on subsequent routine inspections.

10.0 Maintenance (627031 The inspector observed and reviewed selected portions of preventive and corrective maintenance to verify compliance with regulations, use of ad-ministrative and maintenance procedures, compliance with codes and stand-ards, proper QA/QC involvement, use of bypass jumpers and safety tags, personnel protection, and equipment alignment and retest. The following activities were included:

-- Weld Repair to Leaky Letdown Valve (3CHS*V995)

- Aus111ary Feedwater Valve FWA*FCV3101 Controller

- Main Feedwater Valve FW5'CTV41A Limit Switch No inadequacies were identified.

l 4

I

\c

, , 16 11.0 Surveillance (61726)

The inspector observed portions of surveillance tests to assess perform-ance in accordance with approved procedures and Limiting Conditions of Operation, removal and restoration of equipment, and deficiency review and resolution. The following tests were reviewed:

-- Diesel Fuel Oil Transfer Pump Readiness Test, dated 5/1/89 "A" Motor Driven AFW Pump Readiness Test, dated 5/1/89

-- LPSI Injection Valve Stroke Timing, dated 5/8/89 No inadequacies were noted.

l 12.0 Management Meetinos (30703)

Periodic meetings were held with station management to discuss inspection .

findings during the inspection period. A summary of findings was also {

discussed at the conc 1'usion of the ainspection. No proprietary information I was covered within the scope of the inspection. No written material was i given to the licensee during the inspection period.

l 1

e e

9 S

t (

. W}

l. JoAnsor. /d053

? . .

nonnemasy_vpupss _

c o sorrro t

l eT; T;;;_= ~ ~

= . g .-- - ggcowwacTicur osm. caro l

August 1, 1989 Decket No. 50 423 A08112 Re: Inspection 50-423/89-04 U.S. Nuclear Regulatory Connission Attention: Document Control Desk Washington, DC 20555 E. C. Wenzinger letter to E. J. Mroczka, ' Millstone 3

References:

(1) Routine Inspection 50-423/89-04 (4/5/895/15/89)," dated June 28, 1989.

(2) E. J. Mroczka letter to U.S. Nuclear Regulatory Coeuris-sion, 'Rosemount Transmitters," dated April 13, 1989.

Gentlemen:

Millstone Nuclear Power Station. Unit No. 3 Resoonse to Insonetion 50-423/89-04 ,

Reference (1) requested additional information regarding use of Rosemount transmitters at Millstone Unit No. 3. Specifically, our evaluation of contin-ved operability of installed transmitters, our conclusions regarding failure pr:bability, and the adequacy of the present instrument calibration interval were requested. This letter responds to that request.

Since the concern (slow oil leakage) relating to these transmitters was

. originally identified, Northeast Nuclear Energy Company (NNECO) has (and will continue to) aggressively pursue resolution of this issue with Rosemount and industry groups. Although there is still more to be learned, NNECO concludes that sufficient information is available to support a detersination that all Rosemount transmitters installed at Millstone Unit No. 3 are operable (Refer-

~

ence[2]).

Reference (2) provided NNECO's last update on this issue and our conclusions

.on operability.. Information obtained since then has cor.tinued to support the conclusion that these transmitters are operable. In particular, the data.

continues to indicate that this failure mode is strongly dependent on time in

service. Thct is, if a transmitter is failing, it will exhibit symptoms in a l relatively short tieel very likely within the 3 years of service which the If the oil leak were of such Millstone Unit No. 3 transmitters have accrued.

a low rate that no detectable degradatten has been observed thus far, then that rate is low enough to ensure the acceptability of our present calibration intervals. This has been established by an analysis of the failure mechanisa and its cause and is supported by the fact that although transmitter failures f*

i ttachment tt 000153

?OWG N EW

1 4 .

4 U.S. Nuclear Regulatory Connission A08132/Page 2 August 1, 1989 occurred during the first operating cycle, none have occurred since. It should be noted that our prior correspondence, including the latest (Reference (2]) letter, discussed five failures during the first cycle in the reactor protection system. Our continuing analysis of all of our Rosemount transmit-ters and their performance history leads ur to now believe that a sixth trans-mitter may have also failed due to oil loss during the first operating cycle.

The additional transmitter measures charging system flow rate. This transmit-ter is only suseected to have failed due to the loss of oil mechanism, although this has not been proven. The suspected transmitter was replaced during the first refueling outage.

Comprehensive testing performed during the second refueling outage has also supported our conclusions that all of the installed Rosemount transmitters are:

not exhibiting any oil loss symptoms at this time. At this point in transmit-ter life, failure probabilities due to this mechanism are no greater than randos failure probabilities. Protection system design can accommodate these!

low probability transmitter failures without loss of system function.

Duetothenatureofthisfailureanditspotentialeffectonsystemfunction,l NNECO will continue to investigate this issue. Transmitters which can bo:

monitored for early signs of a failure will be so monitored. This monitoring program is essentially a computer aided channel check which can detect smalB amounts of drift between redundant channels and changes in signal distributies of certain parameters. The program will provide early indication of potential l problems leading to increased scrutiny of the affected channels. Trending thoi data over several weeks and evaluating the trends will enable a determinatios of incipient failures to be made before tranruitter function is adverselpl affected. This monitoring program will be implemented by September 15, 1989.i

' Transmitters which cannot be included in the monitoring program will bo:

calibrated each cold shutdown. Attachment 1 provides, in addition to time is service considerations, a sussary of our additional assurances of continues operability. Attachment 2 provides a correction to the Reference (1) discuss sion on failure probability.

Ifthereareanyhostions,donothesitatetocontactus.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY

/

E.'Uf M' roc 1ka f Senior Vice President cc: W. T. Russell, Region I Administrator D. H. Jaffe, NRC Project Manager, Millstone Unit No. 3 W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and :

I 1

1 ' Docket No. 50-423 b9.E.L12 Attachment No. 1 l Millstone Nuclear Power Station, Unit No. 3 1

i Respor.se to NRC Inspection 50-423/89-04 Summary of Additional Assurances of Operability 1

August 1989

4 U.S. Nuclear Regulatory Comission A08132/ Attachment 1/Page 1 August 1, 1989

- Millstone Nuclear Power Station, Unit No. 3 Response to NRC Inspection 50-423/89-04 Summary of Additional Assurances of Goerability The entire group of Rosemount 1153 and 1154 transmitters was evaluated for operability. All transmitters were grouped into one of the foll,0 wing subsets:

A. Transmitters subject to low static pressure ($ 100 psia) are not. subject to this failure mechanism. Thus, no additional monitoring is necessary.

Redundant' transmitters in low range codes (Codes 3, 4, and 5 1.e.,

B.

i 100 psid) will be monitored by comparison to each other to detect oil loss before transmitter operability is affected.

C. Nonredundant transmitters which monitor a noisy process will be moni-l tored. By monitoring signal noise, failures can be detected before the '

l transmitter response time has significantly degraded (up to one minute).

This includes differential pressure transmitters operating at midrange lj and pressure transmitters operating near the upper end of the range.

These transmitters either perform no automatic function or perform automatic functions for decreasing pressure. ,

D. Certaio transmitters monitor parameters which have redundant or diverse instruments. For these transmitters, additional assurances beyond time in service are not considered necessary.

1 I

i b

c.

a.

Docket No. 50-423

'o Aall2Z i

l Attachment No. 2 i

N111 stone Nuclear Power Station, Unit No. 3 Response to NRC Inspection 50-423/89-04 Failure Probabilities l

l l

1 l

August 1989

U.S. Nuclear Regulatory Commissicn

' A08132/ Attachment 2/Page 1 August 1, 1989 Millstone Nuclear Power Station, Unit No. 2 Response to NRC Inspection 50-423/89-04 Failure Probabilities MRLCDMMENT:

Estimated Failure Probabilities The licensee estimated the failure rate by relating 78 loss of-oil failures:to the full manufacturing base produced from 1979 to 1987. The 78 failures represented 0.0565 percent of that manufacturing base. Using a round up value base, the failure rate for 1 year of of service was then (0.01/8760 =) 1.1 x 10 s/ hour. This failure rate compares 1 percent defects in the manufacturing favorably (1/30th) to the rate assumed for random failures in probabilistic safety analyses of 3.85 x 10 8/ hour according to the licensee.

Inspector review concluded that the above predicts 0.014 failures in an 18-month operating cycle, instead of the 5 failures that actually occurred at Millstone Unit No. 3. Also, of the 106 units installed at Millstone 3,16 were from the suspect lots. Five of these failed in service over the period from March to November 1987. The inspector therefore concluded that the licensee's estimate (1.1 x 10 8/ hour) was overly optimistic.

NNECO RESPONSE:

In the above coment, the NRC calculated that 0.014 failures would be pre- 1 dicted in an 18-month operating cycle. This is incorrect. The 0.014 number is correct for only one transmitter ever an 18-month period. With !

107 1153/1154 Rosemount transmitters installed at Millstone Unit No. 3, the correct number of predicted failures for an 18 month period would be:

(1.1 x 10 s/ hour) x (13,140 hours0.00162 days 0.0389 hours 2.314815e-4 weeks 5.327e-5 months ) x (107 transmitters) = 1.5 transmitters The failure rate was

established from a conservative estimate of known fail-ures, fatture times, and the total manufacturing base. Its applicability to Millstone Unit No. 3 is appropriate as only 15 percent of the installed transmitters were frosi suspect lots. Evaluated over a more applicable time period (April 1986 to April 1989), the failure rate would have predicted 3 transmitter failures. . This is a more statistically appropriate period upon which failure rate assumptions may be evaluated.

Whil'a the actual number of failures at Millstone Unit No. 3 is slightly higher, the predicted failures are close to the actual number; therefore, we conclude our estimate was not everly optimistic.

Additionally from the information available to industry, Rosemount data, and our own evaluation of the failure mode, we can be confiderit that the failure probability decreases significantly with time in service. Beyond the

U.S. Nuclear Regulatory Coanission A08132/ Attachment 2/Page 2 August 1, 1989 ,

36 months identified, the assumed failure rate should be significantly lower if used to support predictive failure probability calculations.

IGtC COMMENT: .

The inspectnr reviewed the methodology employed by Westinghouse in l WCAP-10271-P-A, Evaluation of surveillance Frequencies and Out of Service '

Times for Reactor Protection Instrumentation System. In that methodology the i detectability of failures is an important factor that is accounted for explic- ;

itly in the calcolations. Undetectable failures are defined by 1EEE379 as i failures that cannot be detected by periodic testing or cannot be detected by '

alarm or anomalous indications.

The Millstone 3 studies show that the Rosemount 1153 transmitters can be significantly degraded in place for months (incapable of providini a trip ;

function in the upscale direction), with this condition not detectable by any alarm or anomalous indication. The condition might be detected by periodic test, but the calibration surveillance interval is 18 months (13,140 hour0.00162 days 0.0389 hours 2.314815e-4 weeks 5.327e-5 months ) and l the mean time to detect failures is 13,140/2 = 6570 hours0.076 days 1.825 hours 0.0109 weeks 0.0025 months . On the other hand, l the WCAP methodology assumes a failed instrument will be detected within two operating shifts (i.e., 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months ), a significant difference.

While the WCAP failure rate is 2.8 x 10 s/ hour and the licensee estinated a 1 percent defect (1.1 x 10 s/ hours), actual experience at Millstone 3 appears to be much worse. The WCAP assumed detection interval was 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , at Millstone 3 it is 6570.

The WCAP concluded that the failure probability (P) using 1 = 2.8 x 10 s/ hour and T = 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months detection interval was acceptable, where:

P = AT/2' 2.2 x 10 s )

. Using T = 13,140 hours0.00162 days 0.0389 hours 2.314815e-4 weeks 5.327e-5 months , and the licensee's assumed failure rate 4 (1.1 x 7 x 10,10**/ hour), the failure probability .results would be approximately a.

Considering' the 5 failures experisinced at Millstone Unit No. 3 in 18 months, the failure rate would be 4 x 10 */ hour. This is a much worse failure rate than 1.1 x 10 s/ hour.

. This matter will be reviewed further with the licensee to determine whether a more definitive failure probability can be obtained.

19tECO RESPONSE:

The use of WCAP 10271-P-A is inappropriate to calculate acceptable or unac-captable failure detection times based upon independent transmitter failure probabilities. NNEC0 has contacted Westinghouse for their interpretation of the NRC's usage of this WCAP. Westinghousa stated that the purpose of the

g ..

. d

U.S. Nuclear Regulatory Commission A08132/ Attachment 2/Page 3 August 1, 1989 WCAP was to determine the increase in overall plant risk resulting from a Westinghouse relaxation of surveillance frequency from 1 month to 3 months.

also stated that although the philosophy was appr transmitters with 18-month surveillance intervals was inappropriate.

The WCAP is a risk-based study concerned only with increases in ris recognizes that there can be a wide spectrue of failures wh different detection periods.

failures which could be detected by mornal operating shift checks. (16 Westinghouse stated that in the case of common probability mode(during of failure failures thewhic be detected by daily channel checks, the '

increase surveillance period) quantified in terms ofto overall plant risk isAltho the appropriate evaluation. evaluation and analysis would be required transmitters plant risk resulting from applications which use Rosemount subject to loss-of-oil considerations, with 18-month surveillance inter transmitters are now As we have previously stated, the Millstone Unit Mc. 3 Using the substantially beyond their 36-month infant mortality period.

assumed failure rate (1.1 x 10 s/ hour) in failure probability calculations will not provide appropriate input to such a calculation.

i l

.. i e

9 D

l (- I

{ May 1.' : 9c2 e- \

1 Chief, Rules and Directives Review Branch U.S. Nucicar Regulatory Commission Washington, DC 720555

Subject:

Com ents on Proposed Supplement to Rosemount Bulletin I am an employee of Northeast Utilities (NU) in Berlin, CT., however the comments which follow are my personal opinions and those of other knowledg9able industry individuals, including some of the NUMARC Ad Hoc Advisory Committee (AHAC) members. These comments do not ' necessarily reflect the positions of NU. I have provided comments directly to NUMARC as a representative of the AHAC chanered problems. ' with the resolution of the Rosemount loss of fill oil f

From my personal knowledge, the loss of fill oili problems were first identified to the VRC Staff in March 1988 by NU via 10 CFR Part 21 notification and sossibly before this by other utilities. In April 1989, I myt. with the NRC Staff and made a presentation discussing the technical aspects of the failure mechanism and the possible safety implications. Fellowing this meeting the NRC issued an Information Notice l which discussed the loss of oil symptoms and the fact that these failures moy not be detectable during operation. This Notice required no actio n by utilities even though it was understood by NRC Staff and pther technical experts, that this failure mechanism not only preseAted d significant risk, but was clearly in violation of existing NRC Reg 21stions .

During Novemb'e r 1989, I again met with the NRC Staff to discuss the potential risk of these failures. In March 1990, the NRC finally issued Bulletin 90-01 on the Loss of Fill Oil in Rosemount Transmitters. This Bulletin required some action by utilities,

'I NRC Information Ifotice 89-42 " Failure of Rosemount Models !!S3 and itS4 Transmitters" dated April 21,1989 2

10 CFR 50.55 a (h) requires compliance with IEEE 279 and IEEE 379 which state that non detectable failures must be censidered as failed along with a postulated single failure. 10 CFR 50 Apx. A GDC 21 requires "..a capability to l est et,annels iodependently" during power operation "to determine failqres asi losses of redundancy that may have occurred."

f t !!

w i

a,--

O 9 May 1.1992 e

l l

however it did not require compliance with existing regulations. The Bulletin failed to discuss or attempt to address the risk associated with the failures. As a result of the weak wording of the Bulletin.

many utilities took little or no action and did :not address the fact that " undetectable failures" are nrohibited by NRC reculations. The NRC, Rosemount EPRI, NUMARC, General Electric, and most utilities have all acknowledged, or otherwise implied that this failure mechanism regulations. is undetectable and therefore in violation of existing l

During the Fall of 1990, I informed NRC Region 1 personnel of my concern that many utilities are not l only operating outside regulatory requirements, but that this undetectable failure mav also pose a sienificant risk to the safe operation of some nuclear' plants. I still have not had any response from the NRC Staff on this particular issue.

The NRC's proposed NRRtoStaff revision met with me in November 1991 to discuss a the Bulletin. At that time II informed them of both my technical and regulatory concerns associated with the initial Bulletin. i I was also informed. that a proposed supplement to the Bulletin would be issued in December, or at the latest in January 1992.

' On April 23, 1992, 1 received a copy of the proposed supplement to the Bulletin along with the Brookhaven Techfical Report 3 from l

NUMARC. My review of this report indicates that it is well documented and founded upon a strong technicall basis. The report concludes that the failure rates of Rosemount transmitters are significantly 4

greater than originally calculated by the NUMARC report .

Comparing the failure rates from the NUMARC report to the rates predicted by the Brookhaven' report indicates that the latter is 20 times greater than the failure rate previously predicted using the l

3 "Evalustion of Surveillance and Technical Issues Regarding Rosemount Pressure Transmitter Loss of Fill-oil Failures" dated December 21, 1991 4

Summary Report of NUMARC Activities to Address Oil Loss in Rosemount Transmitters dated May 1991 2

i May 1,1992 same data base. , This does not include the very large number of failures due to hther mechanisms such as " particle contamination" and normal electronic failures.

Using the calculatLon given on Page 17 of the Brookhaven report,and assuming the mean failure rate of Ix10-5, the unavailability of the RPS is predicted to be 0.1% for a 2 out of 4 logic and 1.3% for a 2 out of 3 Idgic. This relates to 1.3 failures per 100 demands and is not assuming the worst case failure rate for Rosemount transmitters.

Unless je can be demonstrated that the 95 percentile failure rate is less than 5x10-7 failures per hour for a 2 out of 3 logic, there is no justification for c mtinuing with 18 month calibrations. Even with this, the risk of fai ure of the RPS is increased by 100%.

To relate this ris) to other industries, the historical failure rate of commercial airlinen landing gear is about I per 10 million landings.

The automobile in .ustry would issue an immediate recall with this 4 projected failure rai e if the' breaks on a car failed I out of 100 or 1000 applications.

risk. I believe the Process industries would not tolerate this level of nuclear industry must take immediate action to mitigate this situation. With more than 100 operating reactors in the US, I am personally very uncomfortable with the use of unmonitored and undetected failures of Rosemount transmitters in safety systems.

I am enclosing two graphs which depict the relationship between mean time to detect the Rosemount failure and RPS unavailability.

Data is plotted for four values of failure rates which covers the expected range for undetected failures of R,osemount transmitters used in both 2 out of 3 and 2 out of 4 logic configurations. These graphs are for illustrative purposes only and have not been independently verified.

This discussion relates to only the Reactor Protection System reliability and does not consider the fact that many of these sensors, also, provide inptlt to ATWS ESF and ECCS systems. This common mode failure ' mechanism chullenges our

" defense in depth" philosophy as one may postulate a common mode failure of RPS, ESF, ECCS and the ATWS systems.

! From my review of the Brookhaven report and the proposed supplement to the Bulletin, the mHsing piece of information is that in order tf . determine a surveillance interval, one must first determine i 3

L

[ 1 l I 1

May 1.1992 an availabklity goal. This goal is n 3t provided in either the report or l the proposed supplement to the Bulletin. My discussions with i

knowledgeable industry personnel j rom Brookhaven, MIT, NU and EPRI indicate that an unavailability for the RPS should be less than 5 x 10-5, or 1 failure per 20,000 de mands.

This does not take into accoumt the fact that some of th ese Rosemount sensors may be i l

common inputs to RPS, ESF, ECCS an:1 ATWS systems. The NRC or the industry must establish an accepaable number for these failure ;

probabilities. Utilities should then t interval - based upon this number.

e able to justify their monitoring Using' the failure probabilities given i l

in the Brookhaven report along with the equation on page 17, an appropriate 'mean time to detect "

or sutveillance interval can be dete rmined. This is similar to the approach taken by NU and transmitled to NUMARC via letter dated May !!S,1990. Graphs similar to Il i lose enclosed may also be useful in determining the appropriate detection intervals.

The FRC has excluded Rosemouns transmitters manufactured after July 1)S9 from the requirements of khe proposed Bulletin. The NRC has also excluded Rosemount 1151 and 1152 models from 1 consideration, even after Rosemount has' issued a recall of some 1151 models.due to loss of oil. I do not .believe that there is any fitm basis l for thistransmitters.

these exclusion until a valid statistical data base is established for The only / difference between these new transmitters and those manufactured prior to July 1989, is that their failure probability may be lower initidlly, but in my opinion, will likely increase with time in service.

If this does occur, then it may be necessary to issue another supplement to the Bulletin, which will be of further embarrassment to the nuclear industry.

. I t

e 1

4

a i

May 1.1992 3

When we combir e this failure mechanism with other known, high probability, common mode failures such as the Potter Brumfield MDR relays 2 , ITE J10 relays' and the failure of pressurizer level condensate pots, and one may question the overall regulatory philosophy of the Nuclear Regulatory Commission. I have personally witnessed the expenditure of billions of dollars on issues that have less safety significance than the failures of Rosemount transmitters.

In other situations I have seen large expenditures, which in my opinion, have resulted in a decrease in reactor safety such as the recent NRC requirement to install an Emergency Response Data System GRDS).

My personal concem is not NL"s nuclear plants, as very responsible actiorf has been taken to resolve these issues due to NU's total comrr ilment to true nuclear safety gand the dedication of the NU opera ing and engineering staffs.

other, plants may be My overriding concern is that operating with a risk which is totally unacc eptable due to these unaddres~ sed issues The industry conce rns are and the general public need to know that ::sfety are id being fully addressed and prioritized whenevet they We need to provide the assurance that the safety of the p,c,entified.

opie living near the plants, those operating the plants, the protection of the environment, the economic viability, and the integt: ty of the industry are the primary concerns.

L t

5 NRC Information Notice 92 04

Potter & Bnimfield Model MDR Rotary Relay Failures" dated January 6.1992 5 NRC Information Notice 92 27 " Thermally Induced Accelerated Aging and Failure f ITE/G01:1.D A.C. Relays used in Safety Related Applications" dated Apnl3. 1992 I

i 5

~

May 1.1992 a-We all recognize tnat the reculations discussed herein are unrealistic and that there is a combination of failure probability and detection interval which provides an acceptable risk. I believe that the proposed supplement to the Rosemount Bulletin. incorporating all the ;

above comments, provides a reasonable compromise between an i

unrealistic regulation and acceptable nsk.

I am not requesting that the NRC develop any new regulations. I am not even asking for total compliance with existing unrealistic, regulations. I am only requesting that the NRC take prudent and timely action to assure the continued safety of the operating nuclear power plants.

Very truly yours, j 9/. /$h Paul M. Blanch 135 Hyde R d.

West Hartford, CT. .

06117 xc: Mr. Ernest Hadley, Esq.

l 6

o

0

.. . c :- va.

.. '. ' n. . o O A.

6 C *u I r '." \i / b +-

4+

sa

\ 11 a )

s.a s.* e N x x .,a x x x

.=

o c- c, o cw o

' w. a e#

.. .s

, ,. C. : - ' ' tes.

- 3. : > ;..~. ,, , ,,

rw W $ ..J . .

. _.,o

.., w

.......&.,,,.,}

".. : :? '~~; ~:::;.;=.. =.r=...:..::,.:: ; k.~ ..; .. ::. :::T*.- .. ....

n :trt p::-

?::::i=*~gi:

.. .::.r..,,,,~ . :1b=,.:. 1: ::

. x 10, , , . .

1.,: A y:,- ,

.J.: ,:, ,,,: .

, ,. _.- o . . i. . ,

y.. i.,.i.

h . ~ . . - - - .

.~ ..} _ _ _

. .7 .

. .u, m <*,

c.

E

. m,8 i . ..=...?.:..:.......%.... .. ._. , =, *, n ... .s=.n ............t__..&.~._.&.,.:v.

-~. g.:~.~+=r t..-.- . = ?.:..::.r,.* .. , . . ;r, w:: ::. . . . ,:.3.1

...._.._t. O

~. .

V e N

4

, ...:2. t: ,.2...,._ . , ..ry ,_,,

4 I l i 3 l ? i' ! l $ . Ii'

'1

.- :+ lj f l8 A .. ; 4 1

.aa..A.-.-.-..

.f 4

gf g - . f.h. :. ,

p.m.s

_ .. ..a y p,)

. . . . . . . .. . . . . . - :::. .?,? . . . . 9,..

e:.~::.n.?. .:.,-n.... -~ .

. x10-7

. .. ....es.... . . . -

i. ,:?,:: v;:- . . .~-~.. ~t...

. t t:;. ~m', . :... ,

?,.. . ...r: . . 94 r;...;; V C*

.~

g i , . . :,

i*gq .

..N , ; ; ;i." i i ii;m 'O ***

4.,, w .- . ..

c.- . .

4..N- .i : .**. .. .

.<+---*++t..t .. , Q *'".','

y.............._ p

(,Q . . . . . N. . . . - . . . _ .

e. - ..A"r :. .:;:. . .{.:=....=. . . . . n.~, . ,.:. r.._ . ........ .... .-... . , A. c.. ~*~;-*:::r.q=-.:y- p...y3: :-

y""1 t-*

..?=..

c .3.. . .:p:. .q. p...

1x10~ ; , ,;;..

C

....1.

i

_..-.... . i g ; e 3

i N. .

.. -! i . ; .. n. ,i O u7 ..........x. ---.....w.~.

. W -

. y . .. g . r .

.......A.......x _ . .

3

-, p .x,0-5

- . p .. .v2 x=:...; re;;;p . v.: p.. : .r J

. , l :, .. I x , . , ,:1 N;, .

. g ! l : .1, i g ; , :i;i r-t

,3,,,

, r-d

).

L..

3 e

g yU' 2'

. . . . N.... .:.,.. ...

g; . ... .4.

>#.u... . ] .e a ;x10~4 8- . .-d . , ,,.x . .,

7s 4 ... . f.y CA

.,n N i : :::in

[Q P i

o O' -

. i g

. i .

Ni: # it, N i ; i a,i e , . .

v

- ) s. u. .; ._

a. x<. .

_. u.

. . ._ q. ..

o

.m . _ . o.,...a gn m, . . . . . -. .,

.l',s b"d n n.r .. : C W Fd. 1x10~g 1- .: um- i . r*

N l T- -

' t~h t e 8"9 E ,;

if rT 4 ,

j i ; i h;I . ; \: i m : 1N;y l \! i i{

"$ M

, , ;i '

. \.: . .

is.a Q N

.a,. . -

1 ,

s ,.0 w .u _..

. . - z s.

.. . r-t

. , , I" . . . . .

lx10 2 , , , , y, g

. m ..

s .r ,,3A , . . . . .'

f1, c ("t

>-3

~

t 1

,,,1 , , , ,;ig, \ , ; ,,fg, 3 , , ,it , , ,_

O. . . . x

,._..s.. . y. C, *1

.a . . .6 i ,

,, a

-j

, _ . .. - _. ..... . .. . X :. : .. .

2

_x.

.. . . .. es

. . . . . . - . . , . . . . K j.. ..

r - \ e.- b 1 x10~1

.~ . i i ! ; i.,;; h.i :

! . . ;..- i n. . .

.h p3+. ;a: 1 i Nin e,,)

y _, r-t

. ---: .- . . . .. . . .. :- e . . . -

. . r.--*::

.2

h. .

-t --. -~_-M .Y F

31 O

1x10 0

g

i ;;;m

, i -~

i

ni mp i i iion i i ic--t- nt. % y! i i pMt

( ,. ta t G g

.. _= . w.. . i. 1 0

, . . . tre,r

.x

. .-; x , e:r..:3-....t=....t.in - -. . : .. .. -.s . .. c.4 rt 1

.m .t. . . ..:.. : = - - _. . .v. e.z

.... .. . .t. .s. .tCo g 1x10 i i ip. l ,it i ; ;;nn i !

6 g i Nj_g. n 1 -

I. . . .

3( I . i

=*"'*"*~ *:.} *:1.1.i[*[,':.*.*.**.;1',.*::! "[:*l:tf _"*"'** .g,

....g

. . . q.. *. *y;

.. : .+ '- ,

.; . ~

4 -. ... 4 . f**i g** '[Y, ='

Y'{**m ~:. g .. g ,";** *.1, I 1x10 2 '

o o n m ,,m

.- J u

tt o 1 J u 11 M M-O N a ia m.*l f.0

.i.

I 8

N"" lW""*

--i "t:m

- I I

i I

l 2

0-

.?,qaa. n '.*' _* .~..e .. o Dse t e c r=.

e

( '.T' ) ( h .w= .eo. )

x -

x x x x

- - x O C O a O.

r.o s,.s O. o O

. X .t a 40 s vs _

w

$I l ' #

,~

.% . .' k

- *I;,

4 u=4

. . . .- . _ . _ . _ . _ . ..a ... :_ a ._...t

.I. v

. x10 5 EMai-iiiihaA; emf =- WRiW.I *m=a=mip *Wimii g

. .i. -

. m_ x l - -- i i . i.

u. u. , . . . .

u- :;

. - : N : r r -- ~ r- _4 ..

a v .-

M .: ;wn..

s-7 c.

. . . 1:: . ' x .; .

=: =~~ ~ ' c- :. =r::";:. **J $

1x10,,g

i p 4 --

Ariu o

, : g -,

4 4 g, .

. . m, . . e p;.

g O

q P

r;m2w_ g:.:::,=,...::..:.:.:q.=..- up,=m .p,.._s ,, =.,=. ,:=,

. . . ,=m. :.=, ..,=. .: g ~

. .. x..,,.-.

ato , i% . , me i .% i ,om p

i ,i..

-.a.k. O ~. O'

% j

% - .. .:. w

=.s .%.;.y . h 64.. p p :: .ghs .m...: p

.. 3%. .;.wg ;;.

_ - l.14:s p

= = ~.s'" m W

T 1x10~ i ,N $ i. ! ,

i.: .

1 q.. ggpi g

(,a u> , >

t_ N..A.1

.. 5.

_.a.. ...

. . ..t. 4 , 4J .;1 l "**"".!?..T.JJ.".J{t.'

.::** '*." :: . ..,* :.".:. :U*:

',,7;;;,, ,"?.?.a'

..). .:*.2:C*i = . .* .

i% .

1. m 1

Z.*.?? % .*-*::--::::

' ". .~.. .- .5 .

e *** ...*} &w".*.1.. ."././.t./4.. . . /* .gt..T. !*. t .. g g._. . . '.., .pg .. .. .. . . . ~. m".

.. j 1 x10' ..

r e .i

,,, i g i K.x 1

u 3.~_ ,,.;_2_g

. 4 ,,,4.... . w j

a . m a s3 e Y ,C E <

" m .,x10q a e i. n i

, 2.: 3...

i s r n:

.i. '^v - ri_i O M

- 1 i

k H*

-F, .N-y r,.,e.. .

% yte .m.-- i ,

. . i ,y..t : i 4 Q !

k .

W p,,a J

.........' w .. , ~--'2 ~r*11p

~L. ' . . . . . . ,

~----*.

._s.- *sg !

I e O

. 1:-

'~ '"'l 1x10*? 4: $ T'"Ik, I

5 D' ic

. ! i. 12.

. , i. .,.

T N: t. ?)h .I N !

7 j .:-- :3n:

-- g

$ g p:wmg :a.; me: y r.;

.4 Y. '4....-

p. -

. 4. OgN M

- - ::. :p= m ..

>.a .

1x10 i g pa. , , > -

.:m ..

n i en n;f N . -.

-._.g.

tr

. . i,on , x

.o i. .

ix++n pg.. g p$

,,,!!'J

. 4. .

y j ::::: teras-i 4g. : r. 9:An .ffid.;[? e*. T5'JK!)?i**Tik, 9 3.*/ -_-l"*** ,,, n ..g . . 1 '

, , , ,

M

>=8 1x10*1 t

: mt , _4j E i.y? _ M"N.}-. h**2 fI M

4 ;. *"<u .iq

+.1 i , inn 6 A i,i:; m m ' '

44 .e.4. !

i.. - - .

3

..!.4 ,

O

.- & 4 O

2 :\ p puse

-9:!:T:.1rt

'att '

i "- -~ ~;;.t;L:;ege;

...pm-:msn 2n" 9- :4nt--ff

. . r&< :2 .r.a:r:3*.:&

w:rmasterm ja s,m -- -a:tm.. * . . . . . , . n3p ,

r.?tf "

N k- ne .t; 1.r.t;&v.t?i 11 8"") pe

.x 10') .

v is .iti i: _ . t t e, .

o i .....

ws.m m

. - i 3 0,

e rii

- - -. .-gi1 2q TQ i

. . i

~.-- r. :.v .?:.w.tzur

..~ . .1 . u: . . . . _. .

1x10 1 dM...?:r:.tte'i 6,.oi -=.==. .. r:-?.:- . -

t,.

i E MI

.tn i eiin ,o.+ M. . i l.x.nt':.

E:. > M, o I >=**

1 no i t ie -O

'--..--*4 p. 4 = '

.. ; . . . .. : .:-"* C -- 4 4 \ i L.

i m. :mt.rctr :,.4.r.pi a=m q .~. . n . ysq . ..,,.w pt.m e r ,

..t-rT !*Ito M 1x10 2 i

i; t, a ni . i-

>m ; Ani q v ti ..no O

. *~ i i i i *

.t.m !. ! .t--- i .S O

. t -

.s,.- ___ ,w

.t,.

R 3,, j ' "**'.!: 1, l ,t* sh..} . u J" .$g '*{ .4. .. ) *@.J. l .:{ "1. l l[ ,' **f*;U.4 ***%4

,.i M

) M 1x10 ' j "?] .3 '

l$l

. i r i , .ii ,

1U M 71'UJ .{l"i Q

i. . 3 i .j i i e i;ini e i isp J lit i 1, O

..4: . ;

.li i- te M s

lx10 4 ) : = ar-I~ W 5 ==.4.d:fG Q EEFV&un.kNTG=':ad'5WYY k&TII5N **

yc vi u v

-3 M A m y

C N 1 Go

=-r* hae a0*

Nf I'

"=4 I

~, G a l

(

ML20207F620

Contents

Text

Dear Ms. Brown:

References:

Subject:

Enclosure:

References:

Subject:

Navigation menu

ML20207F620

Text

Dear Ms. Brown:

References:

Subject:

Enclosure:

References:

Subject:

Navigation menu

Search