ML20101E947
| ML20101E947 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 12/21/1984 |
| From: | Baxter T, Rosalyn Jones CAROLINA POWER & LIGHT CO., SHAW, PITTMAN, POTTS & TROWBRIDGE |
| To: | |
| Shared Package | |
| ML20101E945 | List: |
| References | |
| OL, NUDOCS 8412260458 | |
| Download: ML20101E947 (139) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
CAROLINA POWER & LIGHT COMPANY ) , . .
and NORTH CAROLINA EASTERN ) Docket No.-50..400 OL MUNICIPAL POWER AGENCY )
)
(Shearon Harris Nuclear Power )
Plant) )
APPLICANTS' PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW ON SAFETY MATTERS Thomas A. Baxter, P.C.
John H. O'Neill, Jr., P.C.
Pamela H. Anderson Michael A. Swiger SHAW, PITTMAN, POTTS & TROWBRIDGE Richard E. Jones Samantha Francis Flynn CAROLINA POWER & LIGHT COMPANY Edgar M. Roach, Jr.
HUNTON & WILLIAMS Counsel for Applicants December 21, 1984 8412260459 841221 FDRADOCK05000ggg
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
CAROLINA POWER & LIGHT COMPANY )
and NORTH CAROLINA EASTERN ) Docket No. 50-400 OL MUNICIPAL PCWER AGENCY )
)
(Shearon Harris Nuclear Power )
Plant) )
APPLICANTS' PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW ON SAFETY MATTERS Thomas A. Baxter, P.C.
John H. O'Neill, Jr., P.C.
Pamela H. Anderson Michael A. Swiger SHAW, PITTMAN, POTTS & TROWBRIDGE Richard E. Jones Samantha Francis Flynn CAROLINA POWER & LIGHT COMPANY Edgar M. Roach, Jr.
HUNTON & WILLIAMS Counsel for Applicants December 21, 1984
TABLE OF CONTENTS Page I. INTRODUCTION AND BACKGROUND............................ 1 II. FINDINGS OF FACT....................................... 5 A. Joint Contention I: Management Capability........................................ 5 B. Joint Contention IV: Thermoluminescent Dosimeters........................................ 5 C. Joint Contention VII: Steam Generators........................................ 20 D. Eddleman Contention 9: Environmental Qualification of Electrical Equipment............. 29
- 1. Introduction................................. 29
- 2. 9A: ITT-Barton Transmitters................ 32
- 3. 9B: Limitorque Valve Operators............. 38
- 4. 9C: Thermal Aging of RTDs................... 42
- 5. 9D: Instrument Cables....................... 51
- 6. 9E: Physical Orientation of Equipment............................... 56
- 7. 9F: Lubricants and Seals.................... 61
- 8. 9G: Type Test Reporting..................... 66 E. Eddleman Contention 11: Polyethylene Cable Insulation................................. 73 F. Eddleman Contention 41: Pipe Hanger Welding........................................... 73
( G. Eddleman Contention 65: Containment Concrete.......................................... 93 H. Eddleman Contention 116: Fire Protection....................................... 93 l
Page I. Eddleman Contention 132C(II): Control Room Design...................................... 113 III. CONCLUSIONS OF LAW.................................... 114 IV. ORDER................................................. 115 Appendix A: Written Testimony Received into Evidence Appendix B: Exhibits P
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"M December 21, 1984 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION :
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) 2
) :
CAROLINA POWER & LIGHT COMPANY ) .
and NORTH CAROLINA EASTERN ) Docket No. 50-400 OL -
MUNICIPAL POWER AGENCY ) .
)
(Shearon Harris Nuclear Power ) J m
Plant) )
i APPLICANTS' PROPOSED FINDINGS OF FACT -
AND CONCLUSIONS OF LAW ON SAFETY MATTERS '
I. INTRODUCTION AND BACKGROUND j
- 1. The Licencing Board will be issuing its second Par-tial Initial Decision in this contested proceeding on the ap- -
plication of Carolina Power & Light Company ("CP&L") and North j Carolina Eastern Municipal Power Agency (collectively "Appli- 2 cants") for a license to operate the Shearon Harris Nuclear Power Plant (" Harris Plant" or "SHNPP"). The general history of the case is summarized in the Board's Partial Initial Deci-sion on Environmental Matters, LBP-84 __, 20 N.R.C. __ _
(December __, 1984). -
- 2. This partial initial decision resolves all safety matters raised as contested issues by the parties, except for i emergency preparedness issues. The hearing on emergency -
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I planning is scheduled to begin on June 18, 1985, and will be the subject of the Board's third and final partial initial de-cision.
- 3. The parties agreed that the following contentions, admitted by the Board on September 22, 1982,1/ would be grouped as safety contentions (as distinct from the environmental and emergency preparedness issues):
Joint Contentions I, IV, V, VI, and VII CHANGE 44 Wilson III Eddleman 9, 11, 41, 45, 64(f), 65, 67, 116 and 132 i See Memorandum and Order (Reflecting Decisions Made Following Second Prehearing Conference) at 7 (March 10, 1983).2/ Subse-quently, the Board admitted Eddleman 132C(II) as a safety con-tention, and granted Dr. Wilson's motion to withdraw his Con-tention III. See Memorandum and Order (Ruling on Wells Eddleman's Proposed Contentions Concerning Detailed Control Room Design Review (DCRDR), Richard Wilson's Motion to Withdraw Contentions, and the Conservation Council of North Carolina's Motion to Withdraw Contentions) at 8 (Oct. 6, 1983); Tr. 777.
Joint Contentions V and VI were dismissed by default. Order 1/ Memorandum and Order (Reflecting Decisions Made Following Prehearing Conference), LBP-82-119A, 16 N.R.C. 2069 (1982).
2/ The Joint Contentions are sponsored by intervenors Conser-vation Council of North Carolina (CCNC), Chapel Hill Anti-Nuclear Group Effort (CHANGE), Kudzu Alliance and Mr.
Wells Eddleman.
.(Ruling on Various Procedural Questions and Eddleman 15AA) at 6-7 (May 10, 1984).
- 4. The following safety contentions were decided by the Board on motions for summary disposition filed by Applicants pursuant to 10 C.F.R. 5 2.749, and are the subject of previous-ly written Board memoranda which record the grounds for those decisions:3/
Eddleman 64(f) (spent fuel cask safety valves):
Memorandum and Order (Ruling on Motions for Summary Disposition of Eddleman Contentions 29/30, 64(f), 75, 80 and 83/84) at 5-7 (Nov. 30, 1983); ,
CHANGE 44 and Eddleman 132 (reactor vessel level instrumentation): Memorandum and Order (Ruling on Motions for Summary Disposition) at 20-22 (April 13, 1984);
Eddleman 45 (water hammer) and 67 (low-level waste disposal): Memorandum and Order (Revision of Schedule for Filing Written Testimony on Eddleman Contention 9; Rulings on Eddleman Contentions 45 and 67) at 3-8 (July 24, 1984).
- 5. In addition, in a telephone conference on July 12, 1984, the Board announced its decisions to gr-nt motions for cummary disposition of Joint Contention VII (1, 2 and 3), and of Eddleman Contentions 11 and 132C(II). Tr. 2167. These sum-mary rulings were provided to meet the needs of the parties, in view of the hearing schedule, in planning the preparation of 3/ In many cases, while not cranting summary disposition of en entire contention, the Board did dispose of some issues.
These rulings are discussed in the Findings of Fact below in connection with specific contentions as relevant.
, their direct cases. Tr. 2166. It is anticipated that the Board will record its decisions on these motions in the Partial Initial Decision on Safety Matters.
- 6. The safety contentions which were the subject of the evidentiary hearing were Joint Contentions I, IV and VII(4),
and Eddleman Contentions 9, 41, 65 and 116. Hearings on man-agement capability (Joint I) were held in Raleigh, North Carolina, on September 5 through 7 and 10 through 14, 1984.
Hearings on the other contentions were conducted October 16-19, 23-26, 30 and 31, and November 1, and 13-15, 1984. An evening hearing session, on October 23, 1984, was devoted to limited appearance statements, under 10 C.F.R. 5 2.715(a).
- 7. The evidentiary record includes the written and oral testimony of 28 witnesses presented by Applicants, one witness presented by intervenor Eddleman, and 14 witnesses presented by the NRC Staff. In addition, in response to favorable rulings by the Board on intervenor requests to subpoena additional Applicant and Staff personnel to appear at the hearing to give testimony, Applicants voluntarily produced an additional nine witnesses, and the Staff produced one additional witness, to give oral testimony. Appendix A identifies, by witness, the location of the written testimony in the transcript.
- 8. The record also includes exhibits received into evi-dence. Appendix B lists the exhibits identified, and indicates the Board's ruling on any offer of the exhibit into evidence.
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II. FINDINGS OF FACT i
r A. Joint Contention I: Management Capability Y-
[ Filed under separate cover from CP&L in Raleigh.]
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B. Joint Contention IV: Thermoluminescent Dosimeters
- 9. The sole issue litigated in Joint Contention IV was: -
"Whether the TLDs [thermoluminescent dosimeters] and measuring equipment and processes to be used at the Harris facility can ._,
measure occupational doses with sufficient accuracy to comply hj with the NRC regulations." Tr. 2218 (Telephone Conference of August 10, 1984).
- 10. Thermoluminescent dosimeters are devices used for ._
measuring exposure to radiation. TLDs absorb and store energy 1; when they are irradiated by ionizing radiation. The dose re-ceived by the individual wearing the TLD can be calculated by -
measuring the stored energy that is released when the TLD is i
heated. TLDs will be used at the Harris Plant to perform rou- ^^
tine monitoring of personnel for official exposure record- [-
keeping purposes. They will be worn continuously by individ- .-
uals working in the radiologically controlled areas of the Harris Plant. Applicants' Testimony of Stephen A. Browne in -f Response To Joint Contention IV (Thermoluminescent Dosimators), if ff. Tr. 6407 (hereinafter "Browne"), at 3. A
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- 11. Joint Contention IV, concerning Applicants' use of TLDs, was admitted as a contention on September 22, 1982. The original contention consisted of four claims: 1) TLDs are in-accurate; 2) TLDs lack real-time monitoring capability; 3) TLDs are inadequate to assure worker health and safety; and 4) TLD readings should be corroborated by the use of portable pressur-imed ionization monitors. Memorandum and Order (Reflecting De-cisions Made Following Prehearing Conference), LBP-82-119A, 16 N.R.C. 2069, 2077 (1982). Applicants moved for summary dispo-sition of Joint Contention IV and the NRC Staff filed a re-sponse in support of this motion. Applicants' Motion for Sum-mary Disposition of Joint Intervenors' Contention IV (Thermoluminescent Dosimeters), January 9, 1984; NRC Staff Re-sponse in Support of Applicants' Motion for Summary Disposition of Joint Intervenors' Contention IV (Thermoluminescent Dosimeters), February 3, 1984.
- 12. On January 10, 1984, while Applicants' motion for summary disposition was pending, the Commission published a proposed rule which adopted the American National Standards In-stitute (" ANSI") standard on TLD accuracy (" ANSI Standard").
49 Fed. Reg. 1205-1211 (January 19, 1984). As was stated in Applicants' motion for summary disposition, CP&L uses the ANSI Standard as one part of its evaluation of its personnel dosimetry program.
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- 13. On April 13, 1984, summary disposition was granted on -j$
a three of the four issues originally contained in Joint Conten- ;
tion IV. The Board found that Applicants' use of TLDs was ade-
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quate to protect worker health and safety, that TLDs were not 5! "
required to have real-time monitoring capability, and that _ -!
m Applicants were not required to use portable pressurized ion- rr
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ization chambers to corroborate TLD readings. The remaining ==
7 issue in Joint Contention IV was articulated as follows: "Does -;
compliance with the 1983 ANSI standard insure compliance with the NRr regulations?" Memorandum and Order (Ruling on Motions -
as fc Summary Disposition), at 20 (April 13, 1984).
- 14. Applicents subsequently moved for reconsideration or clarification of the ruling on Joint Contention IV. Appli- )
cants' Motion for Reconsideration or Clarification of Board ))
a Memorandum and Order on Joint Contention IV, July 18, 1984. On iff ME August 10, 1984, Applicants' motion for reconsideration wasj 7 granted in part and denied in part. Tr. 2217. The Board ruled
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that, since the ANSI Standard was the subject of a proposed 2; "d Commission rule, it would be inappropriate to consider, in an ;mg individual licensing proceeding, the desirability of that stan-dard. See Potomac Electric Power Company (Douglas Point Nucle- ___ ar Generating Station, Units 1 and 2), ALAB-218, 8 A.E.C. 79, 88 (1974). Therefore, that portion of the one remaining issue 55 in Joint Contention IV was excluded from consideration. Tr. -i! 2217-18. As stated above, the final, unresolved issue of Joint b A s L
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Contention IV was clarified for Applicants and the Staff as follows: "Whether the TLDs and measuring equipment and pro-t cesses to be used at the Harris facility can measure occupa-tional doses with sufficient accuracy to comply with the NRC regulations." Tr. 2218.
- 15. Mr. Stephen A. Browne testified on behalf of Appli-cants at the hearing on the remaining issue of Joint Contention IV. Mr. Browne currently is employed by CP&L an a Project Specialist-Health Physics. During the last eight years Mr.
Browne has bee. involved with the supervision and direction of donimetry programs using TLD systems manufactured by Harshaw, Teledyne and Panasonic. He han been employed by CP&L since 1979. In his present position Mr. Browne in responsible for the technical direction of personnel donimetry programs at all CP&L nu *iar plants. Mr. Browne also is a consultant to the National Bureau of Standardo ("NBS") and has been involved with annessing and evaluating personnel radiation dosimetry procos-nors for the National Voluntary Laboratory Accreditation Pro-gram ("NVLAP"), sponsored by NBS, in which CP&L in a partici-pant. Browne at 1.
- 16. Mr. John P. Cusimano, Mr. Seymour Block, and Mr. Ross Albright testified on behalf of the NRC Staff. Mr. Cunimano in employed by the United Staten Department of Energy, Ra-diological and Environmental Sciences Laboratory, an a Senior Physicint in the Donimetry branch. He han been involved
directly in the evaluation, development and implementation of dosimetry systems over the past twenty-three years. Mr. Cusimano also has served as a technical expert to the NVLAP program. NRC Staff Testimony of John P. Cusimano and Seymour Block Concerning Joint Contention IV, ff. Tr. 6560 (hereinafter "Cusimano and Block"), at 1-2. Mr. Block is employed as a Se-nior Health Physicist in the Division of Systems Integration, Office of Nuclear Reactor Regulation at the NRC. 14. at 2. Mr. Block's duties include evaluating the design and operation of operating nuclear power plants and reviewing the Safety Analysis Reports of applicants for licenses. This responsibil-ity includes reviewing health physics radiation protection pro-grams. Id. Mr. Block has published numerous articles in the dosimetry area. Id. at Attachment 2. Mr. Albright is a Radia-tion Specialist with the NRC. His duties include inspecting radiation protection programs at operating nuclear power plants, including CP&L's Brunswick and H. B. Robinson plants. NRC Staff Testimony of Ross H. Albright Concerning Joint Con-tention IV, ff. Tr. 6567 (horoinafter "Albright"), at 1-2.
- 17. At the request of the Board, Dr. Phillip A. Plato testified as part of the NRC Staff panel. Dr. Plato is a Pro-fossor of Radiological Health at the University of Michigan.
Personal Resume of Phillip A. Plato, ff. Tr. 6561. Dr. Plato has directed numerous funded programs on dosimetry and has pub-lished twenty articles in the field. Id. The Board requested
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that Dr. Plato appear because he served as a member of the Health Physics Society working group which developed the ANSI Standard that is the subject of the proposed Commission rule. Tr. 6562. Dr. Plato directed the performance testing program that the NRC commissioned to perform a pilot test of the ANSI Standard as it was being drafted. Id. He is a co-author of
" Performance Testing of Personnel Dosimetry Services,"
NUREG/CR-2892 (February 1983) and " Performance Testing of Per-sonnel Dosimetry Services," NUREG/CR-2891 (February 1983) (hereinafter "NUREG/CR-2891"), which were prepared to document the results of the program of testing dosimeter processors under the ANSI Standard. Personal Resume of Phillip A. Plato at 12, ff. Tr. 6561.
- 18. CP&L uses TLDs as one aspect of a well-established, state-of-the-art dosimetry program which has operated success-fully for over ton years. The dosimetry equipment, which in-cludes TLDs and TLD readers, is operated and maintained by
( trained and qualified personnel who provide dedicated support to the dosimetry program. The program is operated under de-tailed quality assurance procedures which incorporato elaborato quality control checks on all aspects of dosimetry processing and recordkooping, including the uno of TLDs. Browne at 25-26.
- 19. CP&L unos the Panasonic Model UD-802AQ TLD donimetry system to perform its routino personnel monitoring. Browne at
- 3. In the opinion of the exports in the field, the Panasonic
equipment is state-of-the-art. Tr. 6631 (Plato); Tr. 6568 (Cusimano). The degree of accuracy is established in technical specifications stated in the product literature and introduced as Applicants' Ex. 25. The system has been tested in indepen-dent studies and found to be accurate and reliable. Tr. 6568 (Cusimano).
- 20. As one aspect of its TLD processing program, CP&L has participated in the series of tests directed by Dr. Plato at the University of Michigan. CP&L's performance in the tests conducted in 1980 and 1982 is documented in NUREG/CR-2891, which lists CP&L as processor #187. Browne at 8. CP&L passed each category of the 1982 tests by wide margins. Id. at 9; Cusimano and Block at 8.4/
- 21. CP&L also participates in the NVLAP accreditation program for dosimeter processors.5/ Browne at 8. Testing for accreditation was conducted in 1984 at the University of Michigan. Although those results have not been published, CP&L has provided the results of its favorable performance to the Board. Id. at Attachment C. The NVLAP program also includes 4/ CP&L used a Teledyne TLD system at the time of the 1980 tests. Tr. 6437 (Browne). As discussed above, CP&L now uses a Panasonic system. Therefore, although CP&L panned each catego-ry of the 1980 tests, the results of those tests are not di-rectly relevant to this contention.
5/ NBS has selected the University of Michigan as its offi-cial testing laboratory for the NVLAP program. Browne at 8.
on on-site inspection by a NVLAP assessor. CP&L's on-site in-spection was conducted in August 1984, and in October 1984 CP&L was informed that it had been awarded accreditation under the , NVLAP program. Browne at 25; Tr. 6409 (Browne). CP&L is one of only sixteen processors to have achieved such accreditation at this time. Tr. 6409 (Browne).
- 22. Because the current NVLAP program relies on the ANSI Standard to assess TLD accuracy and that standard is the sub-ject of a proposed Commission rule, CP&L uses the ANSI Standard as one assessment of the success of its effort to ensure the continued superiority of its dosimetry program. In formulating the ANSI Standard, the working group of the Health Physics So-ciety considered standards proposed by prestigious organiza-tions such as the International Commission on Radiation Protec-tion, the International Commission on Radiation Units and Measurements, and the National Council on Radiation Protection and Measurements. Tr. 6601 (Plato). Those organizations were in general agreement that for doses around the annual regula-tory limit of five rem, accuracy should be in the range of 50%.
For lower dosos, such an one rem or less, accuracy to a factor of two or three was deemed acceptable. Id.; Browne at 10. The working group accepted these parameters as a starting point for developing a performance standard. Because the standards used .. by those organizations lacked a clear definition of accuracy, however, the working group was required to search the availab5e
literature to find a suitable definition. Tr. 6601 (Plato). The group adopted a standard that combined bias (the average deviation of the measured dose from the true value) and stan-dard deviation (the variation or spread about the average mea-sured dose), in a single formula. That formula was mathemati-cally expressed as P + 2S 4 L, where P is bias, S is standard deviation and L is accuracy. Id. at 6602. A flaring limit for L was set from 0.5 at the five rem level to about a factor of two at lower levels, in accord with the recommendations of the various committees. Tr. 6602 (Plato).
- 23. Two sets of dosimeter processing tests were conducted using this standard. Each processor entered fifteen dosimeters into each category of dose type and range, with the result that over twenty thousand individual dosimeters were processed dur-ing the course of the tests. Id. Each dosimeter was irradi-ated at the testing laboratory with a known dose and returned to the processor for reading. The doses read and reported by the processors were compared to the true dose value to evaluate the processor's performance. The results of the first two tests revealed a serious problem with the flaring limit on ac-curacy. The test procedure specified dose ranges for the low dose level categories in order to provide a flaring tolerance limit. However, the done range in a particular category some-times was so small that the processor could guess at the cor-rect number within a factor of two, as required by the
I standard. Tr. 6603 (Plato). In response to this problem the I standard was changed to require 50% accuracy at all ranges, ex-cept at accident levels where an accuracy of 30% was required. Tr. 6604 (Plato).
- 24. The standard also was changed to provide that P + S 4 L, rather than P + 2S 4 L. Tr. 6607 (Plato). As adopted, the ANSI Standard assures that virtually every dosimeter tested by a particular processor must be accurate to 50% (or to 30% in the case of accident level categories). Id. This standard is appropriate for three reasons. It is in general agreement with the less specific standards suggested by other nationally rec-ognized health physics organizations. Tr. 6599 (Plato); Browne at 12. It is attainable with available equipment, assuming that human errors such as poor calibration factors and clerical errors are minimized. Tr. 6609, 6614 (Plato). It does not compromise protection of worker health and safety. Tr. 6617 (Board). Moreover, the ANSI Standard is the most appropriate performance standard for evaluating processors proposed by any institution because it is the only standard which provides suf-ficient details about radiation sources and procedures to be used for pass-fail testing purposes. Tr. 6492 (Browne).
- 25. CP&L's success in the NVLAP accreditation program and University of Michigan tests indicates that the TLDs to be used at the Harris Plant will be sufficiently accurate to ensure that doses to workers do not approach the regulatory limits.
1 i As noted above, using the Panasonic system, CP&L passed the University of Michigan performance testing by a wide margin in all categories in 1982 and 1984. Browne at 9; Cusimano and Block at 8.
- 26. In addition to participating in the bi-annual NVLAP accreditation program, CP&L has contracted with the University of Michigan for quarterly intercomparison checks of its TLD processing system under the ANSI Standard. Browne at 12.
Although the NVLAP accreditation program does not require such a procedure, CP&L has added new test categories for low energy beta and mixtures of low energy beta with high energy photons.6/ Id. at 13. These categories were added to test CP&L's TLD system capability more fully under the exposure con-ditions similar to actual working conditions. CP&L's results on those two categories were well within the ANSI Standard. Id.
- 27. The continued efficiency of CP&L's dosimetry program also is ensured by its use of extensive quality control mea-sures. The measures taken by CP&L minimize all of the most common causes of inaccurate measurement observed during the University of Michigan tests. For instance, incorrect calibration factors are minimized by using NBS traceable
@/ CP&L has dropped the accident categories, which differ only in dose levels, for the purposes of the quarterly inter-comparison. Browne at 12-13.
radiation standards to determine calibration factors. Browne at 21. The factors were verified during the accreditation tests at the University of Michigan, and are reconfirmed during the quarterly intercomparison and through a monthly cross-check program performed at CP&L. Browne at 21. Each TLD reader is calibrated semi-annually and after preventive maintenance. During the semi-annual tests, neither bias nor standard devia-tion can exceed 10%. Finally, a daily reader calibration check is performed and requires that each TLD be read within 15% of actual dose. Id. at 22.
- 28. Dosimeter variability is minimized by carrying out an initial acceptance test of TLDs received from the manufacturer.
Each TLD in a batch of five hundred must be accurate to within
+15%. The same test procedure is performed semi-annually to determine whether any TLDs should be removed from service. Id.
at 23.
- 29. Clerical errors are minimized through the use of an automated system for reading TLDs. The TLD readers are interfaced to a computer system and most readings are elec-tronically transferred directly to the individual's dose histo-ry record. All manual data entries are subject to a detailed verification procedures. Id.
- 30. CP&L has ensured that its TLDs will respond properly at accident done levels by performing in-house tests to estab-lish TLD response up to doses of one hundred rom. CP&L's
performance in the ANSI Standard tests demonstrates that poor calibration at accident doses is not a problem at CP&L. Id. at 24.
- 31. CP&L also has instituted detailed written procedures for performance of all routine dosimetry operations, formal training and qualification of all operating personnel, and for-mal supervisory review of quality control records at CP&L.
These measures ensure that TLDs will be used and processed cor-rectly at the Harris Plant. Id.
- 32. Although, in some respects, the ANSI Standard may be less restrictive than the checks imposed by CP&L in its own program,2/ CP&L has no intention of relaxing its own in-house program standards if the ANSI Standard is adopted as a final rule. Tr. 6536 (Browne). CP&L is committed to maintaining an excellent donimetry program through the use of state-of-the-art equipment and techniques, regardless of what standard ulti-mately is enacted. Id.; Browne at 26,
- 33. There has been considerable discussion and some dif-forence of opinion about the extent to which current NRC rog-ulations provide a specific standard for TLD accuracy.g/ It in 2/ I.e., the accuracy required in the initial acceptance tent (115%); the accuracy required in the nomi-annual calibration of TLD roadorn (110%); the accuracy required in the daily reador calibration tootn (+15%).
9/ The Board han suggested that an implicit standard for ac-curacy could be derived from the current regulations. Tr. 2218 (Continued Noxt Page)
clear, however, that any final rule adopted by the NRC will be binding on all present and future licensees, including Appli-cants. In the meantime, TLD accuracy can be enforced through oxisting regulations which require radiation protection proce-dures to be cet forth in the technical specifications. Tr. 6631 (Albright). A licensee sets forth its own acceptance criteria, quality control programs and other requirements as radiation protection procedures. Id.9/ If an NRC inspector finds that any part of the dosimetry program, including the use (Continued) (Telephone Conference of August 10, 1984). The Staff dis-agrees. Albright at 4. The Commission has recognized the fact that current regulations "do not address the competency of TLD processors" and has chosen rulemaking as the "only acceptable alternativo proceduro for solving the problem." 49 Fed. Reg., supra, at 1205, 1211.
" Duplication with the pending rulemaking proceeding in to be avoided." See Wisconsin Electric Power Company, et al.
(Point Beach Nuclear Plant, Unit 2), ALAB-78, 5 N.R.C. 319, 326 (1972). A generic datormination of tne issue of TLD accuracy through rulemaking clearly is appropriate because the problem ( is not plant specific and the rulemaking procedure will promote administrative officiency and consistency of docinion. See, n.q., Baltimore Gas and Electric Co. v. Natural Resources Defense Council, Inc., 103 S.Ct. 2254 (1983). While the Board may believe that a specific accuracy standard in donirable as an aid to enforcement, the Board should not impone a standard on those Applicants that in different from that required of other liconnoon. This is particularly true in light of Appli-cants' demonstrated performance in achieving TLD accuracy be-yond that required by the proposed rule and their commitment to maintain the quality of their program in any event. 9/ Applicants' articulation and implementation of those pro-grama han not boon contanted in this proccoding. l I
and processing of TLDs, is insufficient, a citation could be issued against the licensee for violating its own procedures. Tr. 6623 (Albright). This enforcement mechanism provides as-surance that licensees will maintain accurate TLD dosimetry programa during the pendency of the rulemaking proceeding.
- 34. Notwithstanding any possible ambiguity in the en-forcement standards, Applicants have demonstrated that the hardware to be used for TLDs at the Harris Plant is state-of-the-art and that they have taken appropriate measures to mini-mize sources of human error. Applicants' results on a variety of tests indicate that their TLD processing system is accurate enough to meet any reasonable standard that is imposed by the NRC. See Memorandum and Order (Ruling On Motions for Summary Disposition) at 17-19 (April 13, 1984). Applicants have ex-pressed their firm commitment to maintain the current quality I
of their program. Thus it is clear that Applicants have demon-strated both the present capability of their TLDs, measuring equipment and processes, and their commitment to maintaining a superior dosimetry program throughout the lifetime of the Harris Plant. This ensures that occupational doses will be measured at the Harris Plant with sufficient accuracy to comply with NRC regulations.
C. Joint Contention VII: Steam Generators
- 35. The sole issue litigated in Joint Contention VII was whether Applicants are required to consider multiple tube rup-tures in their steam generator tube rupture analysis.
- 36. As originally admitted on September 22, 1982, Joint Contention VII stated:
Applicants have failed to demonstrate that the steam generators to be used in the Harris Plant are adequately designed and can be operated in a manner consistent with the public health and safety and ALARA ex-posure to maintenance personnel in light of (1) vibration problems which have developed in Westinghouse Model D-4 steam generators; (2) tube corrosion and cracking in other Westinghouse steam generators with Inconel-600 tubes and/or carbon steel sup-port plates and AVT water chemistry; (3) present detection capability for loose metal or other foreign objects and; (4) ex-isting tube failure analyses. Memorandum and Order (Reflecting Decisions Made Following Prehearing Conference), LBP-82-119A, 16 N.R.C. 2069, 2077 (1982). On July 12, 1984, the Board granted Applicants' motion , i for summary disposition of parts (1), (2) and (3), of Joint Contention VII, leaving only part (4) for hearing. Tr. 2167. [ Board to insert decision granting summary disposition. See Tr. 2167; Applicants' Motion for Partial Summary Disposi-tion of Joint Contention VII (Steam Generators), May 16, 1984; NRC Staff Response to Applicants' Motion for Partial Summary Disposition of Joint Contention VII, June 5, 1984.]
- 37. In January 1984, Mr. Wells Eddleman proposed Conten-tion 180 which related to an open item in the SHNPP Safety Evaluation Report relating to operator response time for steam generator tube failure analyses. The Board ruled, on March 8, 1984, that Eddleman Contention 180 was included within the scope of part four of Joint Contention VII. Tr. 771. Subse-quently the Board approved an Agreement of Settlement entered into by Applicants and Joint Intervenors with respect to the operator response time issue raised in Eddleman Contention 180, and the issue was dismissed. Memorandum and Order (Ruling on Various Safety and Procedural Questions) at 1-2 (July 27, 1984). Thus at the time of the hearing the only issue re-maining for consideration was whether Applicants are required to consider multiple tube ruptures in their steam generator tube rupture analysis.
- 38. Mr. Michael J. Hitchler testified on behalf of Appli-cants. Mr. Hitchler in Manager of Plant Risk Analysis with the Nuclear Safety Department of Westinghouse Electric Corporation, which designed and supplied the steam generators that will be used at the Harris Plant. Applicants' Testimony of Michael J.
Hitchler In Response to Joint Intervenor Contention VII(4) (Steam Generator Tube Rupture Analysis), ff. Tr. 4012 (herein-after "Hitchler"), at 2-3.
- 39. Messrs. Ledyard B. Marsh and Herbert F. Conrad testified on behalf of the NRC staff. Mr. Marsh is employed as
+ g a Section Leader in the Reactor Systems Branch. His responsi-bilities include supervision of safety reviews of the reactor coolant, emergency core cooling, accident and transient analy-ses submitted for review by applicants for operating licenses. NRC Staff Testimony of Ledyard B. Marsh and Herbert F. Conrad Regarding Joint Contention VII Part (4), ff. Tr. 4176 (herein-after " Marsh and Conrad"), at Attachment 1. Mr. Conrad cur-rently is employed as a Senior Materials Engineer in the Mate-rials Engineering Branch. His responsibilities include evaluating materials application, heat treatment, fabrication, inspection and corrosion control. Id. at Attachment 2.
- 40. The Harris Final Safety Analysis Report ("FSAR")
contains an analysis of a single double-ended rupture of a steam generator tube, consistent with Section 15.6.3 of the NRC
" Standard Review Plan," NUREG-0800, Revision 3. Hitchler at 4.
This analysis demonstrates that release of radioactive materi-als during such an event would not exceed the limits of 10 C.F.R. Part 100. Marsh and Conrad at 2. Joint Contention VII(4) does not challenge this analysis, but questions whether an additional analysis is required for the multiple tube rup-ture scenario.
- 41. A data base of more than four million tube years of experience is available for Westinghouse steam generators using Inconel steam generator tubes similar to the tubes in steam generators installed at the Harris Plant. Hitchler at 5. Over
' Pe. s, a
4 atha' course df,the extensive period of time represented by that data base, only five tube rupture events have occurred in Westinghouse steam generators. All five of these events had ll flow ratesplarge enoughtto cause a plant trip and to initiate
~y i .4 cafety injection, but c'nly one of these'1 events had a flow rate even approximating the full double-en ed tube rupture event an-clyzed in the Applicants' FSAR. Id.
- 42. Based solely'on these historical data for i
Westinghouse steam generators it would be predicted that a tube rupture event could' occur at the Harris Plant with a frequency
, s of1.6xIO-6/ tube-year'; or once in ev,ery 45 years. Id. at
- 6. This prediction was based on Mr. Hitchler's conservative assumption of 3.6 million tube years of experience. The data i
base of 3.6 million tube years used in'Mr. Hitchler's analysis was derived by discounting the available data for four million tube years by ten percent to allow for < tubes removed from ser-vice clue to p, lugging. Tr. 4060 (Hitchler). This is a conser-vative assumption because, in fact, the historical record of steam generator tObe plugging is significantly less than 2.5 percent of tube years. Tr. 4061 (Hitchler).
- 43. One of the five tube rupture events at a nuclear plant of Westinghouse design was caused by phosphate wastage
, 3 , and stress' corrosion cracking. A second tube rupture event was ! caused by denting. Two other tube rupture events resulted from the presence of a foreign object in the steam generator. The H \.
)"
+
~ . _. . _ .
1 l i l
. risk of each of these types of' events has b'een minimized for
~
1 the: Harris Plant, because of design improvements in the Harris
- cteam.generato'rs and secondary system. Therefore the frequency l-E of tube rupture events at the Harris Plant is predicted to be significantly less than the historical data alone would sug-gast. Hitchler at 7-9.
- 44. Tube failure _due to phosphate wastage will be elimi-
- nated at the Harris Plant because AVT water chemistry will be used throughout the life of-the plant. Id. at 8. Thus the predicted frequency of such events at the Harris Plant in a data base of 3.6 million tube years will be zero, rather than one, as in the historical record.
- 45. Denting will develop more slowly than the historical everage, if at all, at the Harris Plant because of the use of
, AVT water chemistry, the reduction of copper in the secondary
- l i
side system, and the une of fresh water condenser cooling rath-or than sea or brackish water. Id. Stress corrosion cracking t
- is unlikely at the Harris Plant because of the limitation on
, the use of copper, and design advances which minimize crevices bntween the tube and tubesheet and reduce the accumulation of overlying sludge. Id. at 8-9. Moreover, any tube degradation at the Harris Plant is likely to be identified before rupture ' could. occur due to extensive in-service inspection, including i i full inspection of all tubes prior to operation, eddy current l
.--n- -- , , , , , , . -, - .-- -,, ,w, -,----.----.-,n
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testing,10/' ultrasonic inspection techniques, profilometry probes, and continuous monitoring of water quality, secondary-side radioactivity levels and leakage rates. Marsh and Conrad at 4; Hitchler at 9. Tubes with tube wall degradation of forty l percent or more will be plugged or sleeved pursuant to Harris Plant technical specifications. Marsh and Conrad at 4; Tr. 4179-81 (Conrad). Due to the reduced likelihood of denting and stress corrosion cracking, and early identification of any-tube degradation, the likelihood of tube rupture due to denting and stress corrosion cracking will be reduced by a factor of five from the historical frequency. Since the historical frequency of rupture events from these causes-is two (out of five events from all causes), it follows that the frequency of such events j at the Harris Plant in the data base of 3.6 million tube years can be anticipated to be 0.4. Hitchler at 9.
- 46. The Harris Plant will implement a rigorous quality assurance procedure to reduce the likelihood that loose parts will be left in the steam generators. Due to the loose parts 10/ This method of testing creates a magnetic field inside the cteam generator tube in order to produce eddy currents in the tube walls. Any discontinuity or flaw can be observed as a disturbance in the eddy currents. Eddy current testing is an important innovation in providing a reliable method of testing tube integrity. See Applicants' Motion for Partial Summary Disposition of Joint Contention VII (Steam Generators), May 16, 1984, at 12 n.6, citing " Safety Evaluation Report related to the D4/D5/E Steam Generator Design Modification," NUREG-1014 (October 1983) at 3-10.
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i l \ monitoring program and other quality assurance procedures, the frequency of tube rupture events caused by loose parts is ex-pected to be reduced by a factor of two from the historical frequency of two. Id.
- 47. Based on the improvements to be implemented at the Harris Plant, the number of historical tube rupture events
- which are applicable to the Harris Plant would be 1.5 (virtu-ally none due to. phosphate wastage, 0.4 due to denting and stress corrosion cracking, one due to loose parts), rather than five (one due to phosphate wastage, two due to denting or ovality and stress corrosion cracking, two due to loose parts) as in the historical data. Id. at 7, 10. On this basis, sin-gle tube rupture events at the Harris Plant would be predicted to occur at a frequency of 0.6 x 10 -6 / tube-year, or only i
once in 120 years. Id. at 10.
- 48. Since a multiple tube rupture event never has oc-curred, Westinghouse developed an alternative to the analysis of the historical data to evaluate the frequency of multiple
, tube ruptures. .The " pressure pulse" model developed for this purpose relates the pressure differential across steam genera-tor tubes to tube failure probability. Id. at Attachment A.
4 Based on this pressure pulse model, a single tube rupture event is' calculated to occur at a frequency of 7.5 x 10 -3 / year. This is consistent with the 8.2 x 107 3/ year figure derived from tube experience data. Id. at 11. Using the pressure 1
pulse model, multiple tube ruptures can be predicted to occur
~
at the Harris Plant with a frequency of 7 x 10 / year, or I once in 14,000 plant years. Id. at 12.
- 49. The NRC Staff does not require that multiple, double-snded ruptures be postulated within the design basis. Marsh nnd Conrad at 3. The NRC Staff position on this issue is rea-sonable because: 1) as demonstrated in Mr. Hitchler's testi-
. mony, the likelihood of multiple double-ended tube ruptures is extremely low; 2) the assumption of a single double-ended tube break covers a spectrum of smaller leaks, including leakage from a few tubes; and 3) the design basis accident analyses are not intended to predict the consequences of an expected event, but are stylized scenarios which bound the consequences of sim-ilar, more probable and less serious events. Id. Probabilistic Risk Assessment analyses indicate that public risk from multiple tube rupture events will not contribute sig-nificantly to overall public risk at the Harris Plant. Hitchler at 12. Moreover, the Harris Plant has an inherently large margin of. safety to accommodate the possibility of multi-pie tube ruptures. Marsh and Conrad at 6. Thus in the unlike-ly event that a multiple tube rupture accident did occur at the Harris Plant, the existing safety systems are sufficient to allow the operator sufficient time to assess the accident and take appropriate actions. Operators will be specifically trained to recognize and manage such events. The safety
l l l cystems and operator actions would ensure that the overall con-scquences were acceptable. Id. at 6. l
- 50. In summary, it is clear from the historical data that single steam generator tube rupture events are rare. These i
svents are predicted to occur even less frequently at the l Harris Plant, where design and procedure modifications have buen implemented to minimize the probability of such occur-( rcnces. Multiple tube ruptures have never occurred, and are postulated to occur with a frequency on the order of once in 14,000 reactor years -- far less than that predicted for single tube ruptures. In any event, however, Applicants' present analysis for a single double-ended break encompasses the sig- ! nificant consequences of a multiple tube rupture event. Nei-ther the Commission's regulations nor the NRC Staff's guidance for design basis analyses require an analysis of multiple tube ruptures.11/ During cross-examination of Applicants' and the Staff's witnesses, Mr. Eddleman developed no thesis which sug-gested such an analysis should be undertaken. It is concluded that, contrary to Joint Contention VII(4), Applicants need not perform an analysis of multiple tube rupture events. 11/ Thus this contention is similar to that aspect of Eddleman Contention 116 that sought to have Applicants demonstrate the capability to fight two simultaneous fires of independent ; cources. That issue was rejected outright by this Board in a ' sua sponte ruling. See Tr. 4352, 4831-32 (Board). l
4 D. Eddleman Contention 9: Environmental Qualification l of Electrical Equipment
- 1. ' Introduction
- 51. Eddleman Contention 9 as litigated in this proceeding I
consists of a " preamble" and seven specific subcontentions, A through G. The preamble states as follows: The program for environmental qualifi-cation of electrical equipment at Shearon i Harris is inadequate for the following rea-sons: Memorandum and Order (Revision of and Schedule for Filing Writ-ten Testimony on Eddleman Contention 9; Rulings on Eddleman
- Contentions 45 and 67) (July 24, 1984) (" July 24, 1984 Memoran-dum and Order"), at 1-2.12/ The statements of Eddleman Conten-tions 9A through 9G are provided in the discussions of individual subcontentions, infra.
12/ Eddleman Contention 9 originally was admitted by the Li-censing Board in its Memorandum and Order (Reflecting Decisions Made Following Prehearing Conference), LBP-82-119A, 16 N.R.C. 2069, 2091 (1982). The contention subsequently was reworded. See Memorandum and Order (Addressing Applicants' Motion for Codification) (January 17, 1983), at 3; Applicants' totion for Codification of Admitted Contentions (December 17, 1582), at 6, A-14. Following discovery and based on negotiations among , Applicants, Mr. Eddleman and the Staff, Eddleman 9 was revised to consist of the seven specific allegations mentioned above. See Applicants' Motion for Substitution of Contention and for I Revision of Schedule to File Direct Written Testimony on Eddleman Contention 9 (July 12, 1984). Mr. Eddleman proposed en open-ended preamble for the rovised contention which would have called into question the entire program for environmental i qualification at the SHNPP. See id. at 4. That proposed pre-table was rejected and Applicants' proposed preamble was adopt-Ed for the rsasons stated in the July 24, 1984 Memorandum and Order, at 1-2.
- 52. Addressing Eddleman Contentions 9A through 9G on be-half of the Staff was Mr. Armando S. Masciantonio. As Equip-ment Qualification Engineer in the Equipment Qualification Branch, Division of Engineering, Office of Nuclear Reactor Reg-ulation of the NRC, Mr. Masciantonio is responsible for techni-cal reviews, analyses and evaluations of the adequacy of envi-ronmental qualification of electrical equipment important to safety in nuclear power plants, including the SHNPP. NRC Staff Testimony of Armando Masciantonio on Eddleman Contention 9, ff.
Tr. 5567 (hereinafter "Masciantonio"), at 1.
- 53. ' Addressing Eddleman Contentions 9A through 9G for Applicants was a series of expert panels. Those panels are identified infra in the discussions of specific subcontentions. I In addition, Applicants' panel consisting of Mr. Robert W. Prunty and Peter M. Yandow provided, for informational pur-poses, an introductory piece of testimony which described briefly Applicants' program for environmental qualification of olectrical equipment ("EQ Program"). (Mr. Masciantonio's testimony also included general discussion of Applicants' EQ Program.) Mr. Prunty is employed by CP&L as a Principal Engi-neer in the Electrical Group and Instrumentation and Control Group at the SHNPP. Applicants' Testimony of Robert W. Prunty and Peter M. Yandow in Response to Eddleman Contention 9 (Envi-ronmental Qualification of Electrical Equipment), ff. Tr. 4971 (hereinafter " Applicants' Introductory Testimony"), at 2. As
i lead electrical engineer of the newly formed Harris Plant Engi-l l neering Section in December 1979, Mr. Prunty was responsible for establishing the EQ Program. Id. at 4. He continues to be l responsible for the EQ Program in a supervisory capacity. Id. at 3-4, Tr. 4989 (Prunty). Mr. Yandow is employed by CP&L as a Senior Engineer in the Instrumentation and Control Group, i Applicants' Introductory Testimony at 6. Mr. Yandow is the lead engineer currently responsible for the detailed aspects of the EQ Program at the SHNPP. Id. at 6-7; Tr. 4989 (Prunty).
- 54. Equipment that is relied on to perform a necessary safety function must be demonstrated to be capable of main-teining functional operability under all service conditions postulated to occur during its installed life for the time it in required to operate. The purpose of.the EQ Program for j electrical equipment at the SHNPP is to ensure all safety-
) related electrical equipment and other electrical equipment im-portant to safety is qualified to be capable of performing its cafety functions in the environment postulated for design basis cvents. Environmental conditions include temperature, pres-cure, humidity, radiation, chemical spray and submergence. i Applicants' Introductory Testimony at 9; Masciantonio at 3-5.
- 55. The Commission's regulations at 10 C.F.R. 5 50.49 es-tablish requirements for environmental qualification of elec-trical equipment important to safety. Equipment "important to safety" includes safety-related electrical equipment,
n:nsafety-related electrical equipment whose failure under pos-tulated environmental conditions could prevent satisfactory ac-ccmplishment of safety functions by safety-related equipment, cnd certain post-accident monitoring equipment. In general, l l cnvironmental qualification is required to meet General Design Criteria 1, 2, 4 and 23 of Appendix A, and Sections III and XI of Appendix B, to 10 C.F.R. Part 50. Staff guidance for meet-ing the regulatory requirements in 10 C.F.R. 5 50.49 is pro-vided in NUREG-0588 (Revision 1), " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equip-msnt," which is endorsed by 10 C.F.R. 5 50.49(k). Applicants' Introductory Testimony at 9-10; Masciantonio at 3-5.
- 56. Applicants' EQ Program for the SHNPP is described in datail in the SHNPP FSAR at Section 3.11. ESAR Appendix 3.11A compares Applicants' procedures for environmental qualification of electrical equipment with NUREG-0588. FSAR Section 3.11 and Appendix 3.11A are Applicants' Ex. 8. Applicants' Introductory Tastimony at 10.
- 2. 9A: ITT-Barton Transmitters
- 57. Eddleman Contention 9A states:
The proposed resolution and vendor's modification for ITT-Barton transmitters has not been shown to be adequate. (Ref. IE Information Notices 81-29, 82-52 and 83-72). 1 r--
For Contention 9A, Applicants' panel consisted of Mr. Prunty, Mr. Yandow and Mr. Richard B. Miller. Mr. Prunty's and Mr. Yandow's qualifications are discussed supra at 1 53. Mr. ~ i Miller is a Principal Engineer with the Nuclear Safety Depart-I msnt of Westinghouse Electric Corporation, the supplier of the Nuclear Steam Supply System ("NSSS") for the SENPP. Appli-cents' Testimony of Robert W. Prunty, Peter M. Yandow and ! Richard B. Miller in Response to Eddleman Contention 9A (ITT-Barton Transmitters), ff. Tr. 5093 (hereinafter "Appli-l cents' 9A Testimony"), at 2. Mr. Miller is lead engineer in the Nuclear Safety Department responsible for electrical equip-msnt qualification. Id. at 3.
- 58. The transmitt'ers addressed in the IE Information No-3 tices referenced in Contention 9A are pressure-type transmit-i tars. ITT-Barton pressure-type transmitters use either a Bourdon tube to measure pressure, or a bellows assembly to mea-cure differential pressure, depending on the type of transmit-ter. In both types of transmitters, pressure changes cause me-1 chanical movement of internal strain gauges, thereby varying the tension. The variation in tension causes changes in elec-j trical resistance of the strain gauges, which is converted into an electrical output by the electronic circuitry of the trans-mitters. Applicants' 9A Testimony at 4.
- 59. IE Information Notice No. 81-29 (September 24, 1981) rsported test failures which occurred during the initial 4
. - - , . - . . .-, , , , - . ,. - - - , , - - - - , --,----n , . ~ . - - - - - . - - - _ . - - . . - - - , . - - -
qualification testing of ITT-Barton transmitters performed by
' Wastinghouse. Two Model 764 differential pressure transmitters and one Model 763 pressure transmitter exhibited erratic behav-ior.(fluctuating signal or step change in the output) during portions of the test sequence. Subsequent testing and evalua-tion led to the conclusion that the erratic behavior would not occur until the product had been in use for at least five yaars. As documented in IE Information Notice No. 82-52 i
(December 21, 1982), all the failures resulted from degradation of contacts in the internal circuit connector assemblies of the transmitters. Applicants' 9A Testimony at 5; Masciantonio at 8; Tr. 5124 (Miller).
- 60. As a result of the investigation of the problem, Wastinghouse and ITT-Barton determined that it could be cor-rected by soldering the connector assemblies. The modification was then successfully retested by both Westinghouse and ITT-Barton. As indicated in IE Information Notice 82-52, i
Westinghouse submitted to the Staff a report which described
- the modification as well as the successful retesting. The
, Staff approved that test report. Applicants' 9A Testimony at 5-6; Masciantonio at 8.
- 61. Both Models 763 and 764 ITT-Barton transmitters will bn used at the SHNPP. The transmitters will be located throughout the containment building, and will be used to per-4 form various safety functions, e.g., monitoring reactor coolant l
l
pressure, steam pressure and steam generator level. .Appli-cants' 9A Testimony at 6. Pursuant to IE Information Notices 81-29 and 82-52, and on receipt of a change notice from Westinghouse, CP&L sent the safety-related Models 763 and 764 transmitters back to ITT-Barton to perform the necessary modi-fication. Applicants' 9A Testimony at 7. Any new transmitters ordered by CP&L will have the modification already made. Tr. 5159 (Prunty).
- 62. IE Informatic n Notice No. 83-72 (October 28, 1983) reported two additional problems with ITT-Barton transmitters.
One problem was a negative shift (decrease) in output during initial exposure to a constant operating pressure. This defect occurred during testing by ITT-Barton of a suppressed zero (minimum measurement greater than zero) Model 763 pressure transmitter.13/ On the basis of further testing, ITT-Barton identified the cause of the negative shift to be combined creep in the link wire (between the pressure Bourdon tube and the otrain-sensing beam) and in the material used to attach the link wire. Applicants' 9A Testimony at 8; Masciantonio at 9. 13/ ITT-Barton reported a similar negative shift problem to the NRC under 10 C.F.R. Part 21 for its Model 763 zero based transmitters. Some of these transmitters are planned for use at the SHNPP. However, the negative shift on the zero based transmitters is less than that for the suppressed zero trans-mitters, and is in fact within the plus or minus one percent drift allowance established for the SHNPP. Tr. 5102-04 (Mill-or). Thi's negative shift problem was found to have no safety signif-l icance for the SHNPP. Applicants' 9A Testimony at 8-9; Tr. 5140-44 (Miller, Yandow); Staff Ex. 7 at 3-4.14/
- 63. IE Information Notice 83-72 also reported thermal nonrepeatability failures at 320*F of Model 763 and 764 trans-mitters during testing by ITT-Barton. Thermal nonrepeatability failure is the inability of an instrument to repeat a specified output, within allowable limits, when exposed to the same tem-parature and pressure to which it was initially calibrated.
j Applicants' 9A Testimony at 6. on further investigation, ITT-Barton determined two separate causes of the thermal r 0 nonrepeatability. One cause was an incorrect procedure in
- ITT-Barton's calibration technique for temperature compensa-tion. The procedure was found to result in previously t
14/ Staff Ex. 7 -- Affidavit of Armando S. Masciantonio, Richard A. Kendall and Robert C. Jones, Jr. in Further Response to Eddleman Contention 9A (November 20, 1984) -- was filed by the Staff following the hearing, and was subsequently admitted
- into the record. Tr. (December 17, 1984 Conference
- Call). As part of the Staff's testimony on Eddleman Contention 9A, the Staff had indicated that it had not yet reviewed the t ITT-Barton and Westinghouse tests and analyses concerning the problems with ITT-Barton transmitters identified in IE Informa-tion Notice 83-72. Masciantonio at 9-10. During the cross-examination of the Staff's witness, Staff Counsel indicated to the Licensing Board and the parties that the review could be completed shortly and suggested that the Staff could submit an affidavit setting forth the results of its review. Tr.
5692-93. None of the parties had any objection to the submis-sion of the affidavit. Mr. Eddleman stated on tdun record that he did not wish to cross-examine the Staff on the contents of the affidavit. Tr. 5770-72. 4 4 4 4 _ -,- . ,-.--.---c,-,-,.---,--,--,-m- -s- -
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i l l
. unaccounted for errors at both abnormal and accident tempera-tures, resulting in an overall change in the specified accuracy daat was assumed for these transmitters. The second cause of l
l the thermal nonrepeatability was discovered to be an electrical l l leakage path through the wiper arm and shaft of the zero and t span calibration potentiometers to the instrument case. This path only created significant errors at high temperatures, and ; in only of concern during accident conditions. Applicants' 9A - Testimony at 9-10. j 64. Based on calibration data received from ITT-Barton, j Wastinghouse calcula'ted expected error deviations and evaluated the effect of any additional deviation on functions performed by the Models 763 and.764 transmitters. Westinghouse notified i j those plants, including the SENPP, where adequate margin did not exist between the safety analysis limit and the set point j ^ for trip or actuation functions, and changed the set points of l the transmitters to provide adequate margin. Applicants' 9A
- Testimony at 10-11. Additional margin will be provided by mod-l ifications to be performed by ITT-Barton to correct the thermal nonrepeatability problem. Id. at 11. Tests performed on
- modified units have established that the defect has been elimi-
- noted. ~ Staff Ex. 7 at 3. CP&L has instructed ITT-Barton to parform the modifications e,n all transmitters returned to the ,
l i fcctory for rework pursuant to IE Information Notices 81-29 and
! 82-52. Id.; Staff Ex. 7 at 3. Any new transmitters ordered l
j will have the modifications already made. Tr. 5159 (Prunty) . I s l
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. w
- 65. In summary, each of the potential safety problems with the Models 763 and 764 ITT-Barton transmitters has been
- cddressed and resolved for the SHNPP. Where necessary, modifi-cations are being made. The Staff has evaluated and approved
- the Westinghouse analysis and testing addressing each of the
- problems reported by Information Notices 81-29 and 82-52
! (Applicants' 9A Testimony at 6; Masciantonio at 8) and Informa-l tion Notice 83-72 (Staff Ex. 7 at 3-5). Mr. Eddleman's cross-l
- cxamination of Applicants' witnesses and the Staff witness !
2 i j uncovered no basis for challenging the findings of these ex-4 parts. The resolutions recommended by Westinghouse and 1 ITT-Barton for the Models 763 and 764 transmitters, which Applicants have committed to implement at the SHNPP, provide reasonable assurance that the potential safety problems with ., those transmitters identified in IE Information Notices 81-29, i j 82-52 and 83-72 have been satisfactorily resolved. I 4 3. 9B: Limitorque Valve Operators 1
- 66. Eddleman Contention 9B states:
i There is not sufficient assurance that the concerns with Limitorque valve opera-i tors identified in IE Information Notice 83-72 (except for Items C2, C5 and C7) have , been adequately addressed. i
- For Contention 9B, Applicants' panel wts composed of Mr. Prunty 1
cnd Mr. Yandow. Their qualifications are discussed supra at I 1 53. j i I
- 67. A valve operator (or actuator) is a component of a valve which causes it to open or close. Limitorque valve oper-ntors contain electrical motors which, through a series of me-chanical gears, cause the valve to change position. Appli-cants' Testimony of Robert W. Prunty and Peter M. Yandow in Response to Eddleman Contention 9B (Limitorque Valve Opera-tors), ff. Tr. 4971 (hereinafter " Applicants' 9B Testimony"),
at 3. Limitorque valve operators are used on a number of valves which perform safety-related functions at the SHNPP and are found in various locations in the reactor containment and the reactor auxiliary building. Id.
- 68. IE Information Notice 83-72 contained information on construction deficiencies related to Limitorque valve operators at Consumer Power Company's Midland Plant, Units 1 and 2 ("Mid-land"). The deficiencies referenced in Eddleman Contention 9B concern: (1) qualification and rating of terminal blocks, (2) qualification of motor insulation material, (3) installation orientation, (4) installation of drain plugs, (5) lack of l cgreement between purchase order and qualification files and I
installed components, and (6) qualification of 0-rings. Appli-cants' 9B Testimony at 3-4; Masciantonio at 10-11. -
- 69. In order to assure that none of the problems with Limitorque valve operators at Midland applies to the SHNPP,15/
15/ CP&L in.its investigation of the deficiencies identified in Information Notice 83-72 found that, for the most part, the (Continued Next Page) l
l CP&L is in the process of implementing a three-part field veri-fication program for all active,1p/ safety-related valves with Limitorque operators located in a harsh environment. Tr. l 4975-76 (Yandow). Part 1 of the program, which has been com-plated, is an inspection of the 16 active, safety-related volves with Limitorque valve operators inside containment. l Applicants' 9B Testimony at 5; Tr. 4975 (Yandow). Part 2, which also has been completed, is an inspection of the eight cuch valves installed in the main steam tunnel in the reactor r cuxiliary building. Tr. 4975, 5040 (Yandow). Part 3 of the
- program is an inspection of the approximately 120 remaining ac-tive, safety-related valves with Limitorque valve operators lo-cated in a harsh environment outside containment. Tr. 4975-76, 5040 (Yandow).
- 70. The inspections for Parts 1 and 2 of the field veri-fication program consisted of the following: (1) measuring the dimensions of the valve operator terminal blocks, including the point-to-point distances of the terminal screws, and comparing (Continued) d2ficiencies were plant specific and were the result of lack of information on the.part of Midland personnel. The field veri-fication program discussed infra is intended to provide addi-tional assurance that the Midland problems are not applicable i to the SENPP. Applicants' Testimony at 4-5. l 1p/ Passive valves are not required to actuate in order to parform their safety functions. Tr. 5029-30 (Yandow).
those measurements with vendor supplied information (Appli-cants' 9B Testimony at 7-8; Tr. 4975, 5085-86 (Yandow)), (2) verifying motor insulation type using motor nameplate ratings (Applicants' 9B Testimony at 9; Tr. 4975 (Yandow)), (3) checking installation orientation (Applicants' 9B Testimony at '10; Tr. 4975 (Yandow)), (4) checking to make sure drain plugs are properly installed (Applicants' 9B Testimony at 11; Tr. 4975 (Yandow)), (5) checking serial numbers and other valve identification data to verify that the installed valves are the ones specified (Tr. 4975 (Yandow)), and (6) verifying identifi-cation by visually inspecting internal valve operator compo-nents (Tr. 4975 (Yandow)).17/
- 71. As stated above, Applicants have completed Parts 1 and 2 of their field verification program. No deficiencies in any of the Limitorque valve operators were found. Tr. 4975 (Yandow). Applicants have not yet begun Part 3 of the program.
The scope of the inspections under Part 3 has not yet been 17/ O-rings were not visually inspected because such in-spection would require disassemblying the valve operators. Applicants' 9B Testimony at 13; Tr. 5079 (Yandow). Disassembly is unnecessary because Limitorque has a valve operator assembly control system, which includes color coding of 0-rings, to as-sure that the proper 0-rings are used in each type of valve op-erator. Information Notice 83-72 did not call into question the adequacy of this assembly control system. Id. However, if the field verification program identifies any components of an operator for which qualification appears questionable, the valve operator will be disassembled and all questionable compo-nents, including any unidentifiable 0-rings, will be replaced. Applicants' 9B Testimony at 13.
determined; however, all items under Parts 1 and 2 applicable to the safety functioning of the outside containment valves in-cpected under Part 3 will be checked. Tr. 4975-78, 5031 l (Yandow).
- 72. In conclusion, CP&L's field verification program for Limitorque valve operators addresses and resolves each of the concerns identified in IE Information Notice 83-72 and refer-cnced in Eddleman Contention 9B. Mr. Eddleman's cross-cxamination of Applicants' witnesses and the Staff witness presented no basis for questioning the adequacy of that pro-gram. Applicants' EQ Program provides reasonable assurance that the problems with Limitorque valve operators at Midland either do not apply to the SHNPP, or will be discovered and corrected prior to full load.
- 4. 9C: Thermal Aging of RTDs
- 73. Eddleman Contention 9C states:
It has not been demonstrated that the RTDs have been qualified in that the Arrhenius thermal aging methodology em-ployed is not adequate to reflect the actu-al effects of exposures to temperatures of normal operation and accidents over the times the RTDs could be exposed to those temperatures. (Ref. NUREG/CR-1466, SAND 79-1561, Predicting Life Expectancy of Com-plex Equipment Using Accelerated Aging Techniques.) On Applicants' panel for Contention 9C were Mr. Miller and Dr. Thomas W. Dakin. Mr. Miller's qualifications are discussed l l l
rupra at 1 57. Dr. Dakin is retired, but continues to serve as a consultant to Westinghouse. Applicants' Testimony of Richard D. Miller and Thomas W. Dakin in Response to Eddleman Conten-tion 9C (Thermal Aging of RTDs), ff. Tr. 4839 (hereinafter
" Applicants' 9C Testimony"), at 1. While still employed by Westinghouse, he was Department Manager in the field of elec-trical insulation at the Westinghouse Research Laboratory. Id.
at 2. Dr. Dakin has a Ph.D. in Physical Chemistry and has pub-lished extensively on the subjects of thermal aging and accel-erated life testing of electrical equipment. Id. at 2-3, At-tachment A.
- 74. An RTD, a resistance temperature detector, is an instrument used to measure temperature in which the primary el-cment, a resistance wire, has a well-defined resistance-temperature relationship. The primary element in the RTDs used at the SHNPP is a platinum wire. Signal condi-tioning equipment is used to detect and amplify changes in the resistance of the platinum element which correspond to changes in temperature. These RTD signals are used in plant instrumen-tetion systems. Applicants' 9C Testimony at 4.
- 75. The RTDs used at the SHNPP are manufactured by the RdF Corporation and supplied by Westinghouse. Eighteen Model 21204 RTDs are directly immersed in bypass lines to the reactor coolant system. There are three coolant loops at the SHNPP and these eighteen RTDs are used to measure the " hot leg" and " cold r- 'l i )
l l l leg" temperature in each loop. These RTDs are directly l immersed to provide rapid time response measurements for use in l the reactor protection and control systems. In addition, six Model 21205 RTDs are installed in wells located in the reactor coolant system piping to provide measurement of the hot and cold leg temperature in each loop for use in monitoring plant conditions. Id.
- 76. The construction of the Models 21204 and 21205 RTDs is almost identical.18/ The complete RTD assembly consists of a platinum element contained inside the tip of a sheath, and the necessary wire and supports which allow connection to a cable system through which signals are transmitted outside the containment building. A stainless steel sheath protects the element and wire over that portion inserted in the pipe. A stainless steel bellows hose protects external wires from mois-ture penetration and physical damage.19/ Id. at 5-6.
- 77. The portion of the RTD inserted in the primary system piping contains no age-sensitive materials. The organic mate-rials in the external cable and cable interface are epoxy
- potting material and silicone varnish cable coating. Only the 18/ The primary difference is in the length of the sheath in-certed into the piping system. Applicants' 9C Testimony at 5.
19/ A helium leak test assures the adequacy of the moisture barrier provided by the bellows hose. Applicants' 9C Testimony ct 6.
~
.cpoxy-potting material has a safety function, namely, mechani- i I
col support and insulation for the wires at the cable-probe interface. The silicone varnish is only used to prevent the i fiberglass insulation on the cable from fraying during the L nenufacturing process. It is not required for the RTD to per-form its safety function. Id. at 6.
- 78. 10 C.F.R. I'50.49(e)(5) requires that "[e]quipment j qualified by test must be preconditioned by natural or artifical (accelerated) aging to its end-of-installed life con- j dition." Since real time aging is not practical over the long
{ j time periods for which most electrical equipment must be envi-i l renmentally qualified for nuclear power plant application, ac-
- colerated processes have been developed to simulate a defined life over a much shorter period of time. Id. at 7. :
l
- 79. The Arrhenius methodology is used to simulate and to predict the aging effects of temperature. The Arrhenius meth-cdology is based on the premise that deterioration of materials !
i
- in service is due to chemical reactions. These reactions occur l
! internally, sometimes between components of the material and j cemetimes with compounds in the environment such as oxygen or f water vapor. Chemical reactions occur more rapidly at in-I creased temperature. Arrhenius in the last century showed i i i i [ -4s- _ _ . _ _ _ - _ - _ _ . ___ . _-___.~.._ ,.--_. _ . .._ _ _.. _ ..___., _ _ _ _ _ ._ -
\
i ! theoretically that the temperature dependence of chemical reac-- tions followed an exponential equation.20/ Id. at 7-8. Usu-cily, accelerated type tests of materials are made extending up to one or two years. The data obtained are then extrapolated d wn to expected continuous service temperatures to predict a cpecified level of deterioration for the material in particular cpplications. Id. at 8-9.
- 80. Since the epoxy is the only safety-related age sensi-tive material used in the RTDs, the activation energy for this material was selected in the application of the Arrhenius meth-cd to the RTDs to be used at the SHNPP.21/ Using the Arrhenius 2_0/ That equation is:
raten/exp (-E/kT)n/a constant / time where T is the Kelvin temperature (degrees C +273); E is the activation energy of the chemical reaction (electron volts); and k is the Boltzmann gas constant (electron volts / degrees Kelvin). The activation energy is characteristic of the material and the oignificant chemical change. Applicants' 9C Testimony at 7-8. Usually this equation is rearranged and changed to the logarithmic form as follows: Ln(Time)d(-E/kT) = (-E/k)/T cnd the logarithms of times to change are graphed versus recip-rocal Kelvin temperature. Id. at 8; Tr. 4836-37 (Dakin). 21/ An activation energy of 0.98 electron volts was chosen for the pre-aging portion of the qualification testing based on ex-(Continued Next Page) 1 cquation and the ambient temperature at the cable interface portion of the RTD assembly,22/ an aging temperature was calcu-lated which would simulate the desired life at an accelerated rate and not inadvertently degrade the material due to the high temperature alone. Following the thermal aging portion of the qualification test, calibration checks were performed at vari- I cus temperatures. Insulation resistance measurements were taken as well. No degradation of the RTDs was detected during these tests. Based on the entire qualification test, which in-cluded thermal aging, thermal cycling, irradiation aging and vibration aging, the RTDs installed in the bypass lines were qualified for a minimum 20-year life and the RTDs install.ed in the wells were qualified for a minimum lO-year life. Id. at 10-12.
- 81. The NRC Staff has approved the Arrhenius method for environmental qualification of electrical equipment in nuclear (Continued) cmination of test data on a number of similar epoxies. This choice was conservative, as it was toward the lower end of the l range of activation energies considered. Applicants' 9C Testi-many at 12;~Tr. 4918, 4954-55 (Dakin); 5652-53 (Masciantonio).
For the simulation of the post-accident period, an even more conservative activation energy of 0.5 electron volts was used cs part of Westinghouse's standard accident profile. Appli-cants' 9C Testimony at 12; Tr. 5956-57 (Miller, Dakin). 22/ This ambient temperature included the expected temperature rise at the cable interface as a result of heat transfer from the reactor coolant system. Applicants 9C Testimony at 11.
power plant applications. Section 4(4) of NUREG-0588, " Interim l Staff Position on Environmental Qualification of Safety Related Electrical Equipment," states: "The Arrhenius methodology is , considered an acceptable method of addressing accelerated l cging."' The Staff most recently endorsed the use of this meth-
; od in Regulatory Guide 1.89 (Rev. 1), " Environmental Qualifica- .
j tion of Certain Electric Equipment Important to Safety for Nu-clear Power Plants" (June 1984). In addition, the Westinghouse
- qualification methodology, which includes accelerated thermal aging techniques based on Arrhenius methodology, has been ac-cepted by the NRC. Applicants' 9C Testimony at 10-11. The Staff specifically approved the Westinghouse generic qualifica-
, tion testing program for RTDs, which envelopes the environ-a mental conditions, including temperatures, for which the SHNPP RTDs must be qualified. Id. at 13; Tr. 4957 (Miller). (
- 82. The Sandia National Laboratories report referenced in Eddleman Contention 9C ("Sandia Report")23/ discusses the use-
- j fulness of the Arrhenius relation in accelerated aging tests i
] but also discusses possible conditions which would invalidate I the use of this relation for extrapolation from accelerated i ; cging tests. The report points out the need for a single i
i 1 23/ NUREG/CR-1466, entitled " Predicting Life Expectancy and i Simulating Age of Complex Equipment using Accelerated Aging Techniques," was first published by Sandia National Laboratories in 1979 as a consultant's report to the NRC. i 1 l 1
l chemical. reaction to control the aging of the material over the whole temperature range from accelerated test temperatures down to service temperatures. If, for example, moisture diffusion were controlling at lower temperatures, this would change the clope of the Arrhenius type graph to a lower slope and predict a shorter failure time than predicted by extrapolating high temperature tests. Applicants' 9C Testimony at 13-14.
- 83. However, it was found that none of the predictive difficulties discussed in the Sandia Report applies to the spoxy used in the SHNPP RTDs. Moisture diffusion is the only potentially invalidating condition mentioned in the Sandia Re-port that could apply to the accelerated aging of RTDs. Id. at 14-15. The insulation system of the RTD connector and cable is osaled against moisture, so that diffusion of moisture is pre-vsnted.24/ Id. at 14; Tr. 4893 (Dakin); 4960-62 (Miller).
Further, the effects of moisture on epoxy resins are much smaller than the effects of moisture on the polyurethane cited in the Sandia Report. Applicants' 9C Testimony at 15; Tr. 4914-17, 4920 (Dakin). Finally, it is unlikely that degrada-tion sufficient to prevent the RTD from performing its safety function will occur because the performance requirements of the spoxy in this particular application are minimal. Tr. 4924-25, i 2f/ The seal itself also is required to be environmentally qualified. Tr. 4960-62 (Miller).
1 i
-4946-47 (Dakin). Thus, the Sandia Report does not support the
. ollegation in Eddleman Contention 9C that the Arrhenius thermal eging methodology is not adequate to reflect the actual effects of exposures to temperatures of normal operation and accidents ! cver the times the RTDs could be exposed to those temperatures. Indeed, the Sandia Report (at page 47) concludes that
"[a]ccelerated aging techniques offer the best opportunity for predicting lifetimes or simulating life of complex equipment."
Applicants' 9C Testimony at 15; Masciantonio at 14.
- 84. In any event, uncertainties that may exist in the Arrhenius methodology as applied to the SENPP RTDs will be ac-counted for by Applicants' surveillance and maintenance pro-gram.25/ Surveillance and maintenance procedures for the RTDs will include periodic calibration checks and performance tests i
which would be able to detect any significant unanticipated degradation of the epoxy in the RTDs due to thermal or other offects. Tr. 4962-63 (Miller, Dakin); 5707 (Masciantonio).
- 85. In conclusion, Applicants have shown that the
! Arrhenius method is satisfactory for simulating the thermal aging of the organic materials in the qualification of the RTDs 25/ Applicants have committed to follow Regulatory Guide 1.33, Rev. 2, " Quality Assurance Program Requirements (Operation)" in developing the surveillance and maintenance procedures for the , SENPP. The Staff will verify that an appropriate program is
- implemented prior to issuance of an operating license.
l Masciantonio at 12-14. I l l i _ _. _ _ _ _ . _ _ _ . . __ l
j s If , t , J' l 1 l_ for the SHNPP. Mr. Eddleman's cross-examination of Applicants' l witnesses and the Staff's witness did not present any basis for questioning that conclusion. The Arrhenius method, combined with Applicants' surveillance and maintenance program as ap-plied to the RIDS,-provide reasonable assurance that w unanticipated de92ndation of the RTDs is not a safety concern for the SHNPP. y a
- 5. 9D: Instrument Cables 3 86. Eddleman Contention 9D states:
, The qualification of instrument cables ^
did not include adequate consideration and analysis of leakage currents resulting from the radiation environment. These leakage currents could cause degradation of signal quality and/or spurious signr.ls in Harris instrument cables. i Applicants' panel for Contention 9D wds composed of Mr. Richard M. Bucci and Mr. Edwin J. Pagan. Mr. Bucci and Mr. Pagan both are employed by Ebasco Services, Inc. Ebasco is the architect-engineer for the SHNPP. Ebasco also supplies the balance-of-plant (" BOP") safety-related electrical equipment for the SHNPP, i.e., equipment which is not part of the NSSS. Applicants' Testimony of Ric$ard M. Bucci, Edwin J. Pagan and t Peter M. Yandow in Response to Eddleman Contention 9F (Lubri- ! cants and Seals), ff. Tr. 5441 (hereinafter " Applicants' 9F s ! Tastimony" ) , at 4. Mr. Bucci'is an Associate Consulting Engi-neer in the Corporate and Co$sulting Engineering Department of
?
A 1 - o 0.4 . _ y s]. . , -. - - - - - - - - -
'Ebasco. Applicants' Testimony of Richard M. Bucci and Edwin J.
Pagan in Response to Eddleman Contention 9D (Instrument Ca-I i bles), ff. Tr. 5166 (hereinafter " Applicants' 9D Testimony"), l nt 2. From 1979 to 1983, Mr. Bucci was lead electrical engi-n;er responsible for all electrical engineering and design ac-tivities performed by Ebasco on the SHNPP project. One of these activities was implementation of the SHNPP EQ Program. Id. at 3. Mr. Pagan is a Senior Electrical Engineer at Ebasco. Id. at 4. He is the current Equipment Qualification Task Leader for the SHNPP, responsible for developing and implementing the EQ program and supervising the work of the Ebasco EQ group. Id.
- 87. 'An instrument cable, in its simplest form, is an electrical cable constructed of a conductor, insulation, chield, drain wire, and overall jacket. More complex construc-tions include various multiples of these basic components.
Instrument cables are-designed to conduct low power electrical signals. During normal operation, instrument cables are used to conduct electrical signals containing information about plant operating conditions, such as reactor coolant system pressure, reactor coolant system temperature, and containment radiation levels. These signals are transmitted from measuring instruments throughout the plant to indicating and control dsvices in the control room and other locations. In the event of an accident, instrument cables transmit the protective !
cction signals required to achieve safe plant shutdown,'to mit-igate the consequences of the accident, and to monitor plant canditions during and after the accident. Applicants' 9D L Tastimony at 6.
- 88. There are several thousand circuits utilizing instru-ment cables in the SENPP design. The instrument cables used cre of various types, and have been purchased from several dif-ferent vendors. Because most instrument cables are routed through more than one plant area, these cables will be exposed to a variety of environmental conditions. Many cables are routed from instruments inside the containment to indicators in the control room. Applicants' 9D Testimony at 7.
- 89. Instrument cables at the SHNPP required to be envi-ronmentally qualified by 10 C.F.R. 5 50.49 were qualified by test. The test methodology employed is the one set forth in IEEE 383-1974.2p/ In the tests, instrument cables were sub-jected to thermal aging, radiation, and other design basis accident conditions (as applicable). Each type of instrument cable used at the SHNPP was qualified for its worst case loca-tion, i.e., for the most severe environmental conditions that 2s/
Field ". Splices,IEEE and Standard for Type Connections Tests ofPower for Nuclear Class Generating lE Electric Cables, Stations" (1974). IEEE 383-1974 is endorsed by NRC Regulatory Guide 1.131, " Qualification Tests of Electric Cables, Field Splices, and Connections for Light-Water-Cooled Nuclear Power Plants" (August 1977).
w l any part of a cable of that type could. experience. In addi-tion, during testing the SHNPP instrument cables were exposed to'substantially higher radiation doses than the most severe doses to which they actually could be exposed under normal and i cccident conditions. Following the tests described above, the instrument cables were required to pass a voltage withstand
'tsst, which subjected the cables to additional electrical and
, muchanical stresses beyond those they will experience in ser-vice. The voltage withstand test indicated th'at margin still eKisted in the integrity of the insulation after qualification tasting. , Applicants' 9D Testimony at 8.
- 90. Leakage current is that portion of an electrical sig-nnl carried by a cable which is conducted through the insula-tion to ground. Insulation resistance is the resistance of the cable insulation to the flow of leakage current. Leakage cur-rent and insulation resistance are inversely proportional.
That is, as insulation resistancs decreases, leakage current increases (provided voltage remains constant). This relation-ship is described by Ohm's Law, which is a fundamental concept in electrical engineering. Leakage currents occur when insula-tion resistance is too low, for example, when organic cable in-sulation has degraded as a result of environmental stresses. Dspending on the sensitivity of the particular instrument to which the cable is connected, a leakage current could affect the accuracy of transmitted information. Applicants' 9D Testimony at B-9.
l I l
- 91. During qualification testing of instrument cables p uced at the SHNPP, leakage current was sensed by a measurement dsvice and converted by the device to an insulation resistance value, which was recorded. At a minimum, insulation resistance wns measured prior to testing, after irradiation, and at fre-quant intervals during the remainder of the design basis acci-dnnt testing (e.g., pressure, temperature, humidity, chemical spray). Applicants' 9D Testimony at 9-10; Masciantonio at 16.
Ebasco has reviewed the insulation resistance values following irradiation for each type of instrument cable used at the SHNPP. In no case did irradiation of instrument cables during { qualification testing result in a significant decrease in insu-lation resistance. Insulation resistance values of the magni-tudes obtained indicate negligible leakage currents in the cir-cuit. Applicants' 9D Testimony at 11; Masciantonio at 16; Tr. 5207, 5225-30 (Bucci).
- 92. In conclusion, environmental qualification testing of instrument cables was conducted according to the applicable standards. Insulation resistance measurements were taken on irradiated test samples. These insulation resistance values '
have been reviewed to ensure that there will be no adverse ef-l l fact on the safety functions performed by SHNPP instrument ca-bles as a result of leakage currents caused by radiation. Mr. Eddleman's cross-examination of Applicants' witnesses and the Staff's witness presented no basis for questioning the
l i conclusion that such ieakage current effects will be negligi- ! l ble. There is reasonable assurance that the qualification of l instrument cables for the SHNPP does include adequate consider- ' ction of leakage currents resulting from the radiation environ-l ment. 1
- 6. 9E: Physical Orientation of Equipment
- 93. Eddleman Contention 9E states:
There is not sufficient assurance that the physical orientation of equipment in testing is the same as the physical orien-tation of equipment installed. For Contention 9E, Applicants' panel consisted of Mr. Bucci, Mr. Pagan and Mr. Edward M. McLean. The qualifications of Mr. Bucci and Mr. Pagan are discussed supra at V 86. Mr. McLean is employed by CP&L as a Project Mechanical Engineer. Applicants' Testimony of Richard M. Bucci, Edwin J. Pagan and Edward M. McLean in Response to Eddleman Contention 9E (Physi-cal Orientation of Equipment), ff. Tr. 5234 (hereinafter
" Applicants' 9E Testimony"), at 2. Mr. McLean was, until re-cently, supervisor of a group responsible for providing engi-neering support for the installation of equipment, including electrical equipment, at the SENPP. Id. at 3-4; Tr. 5368-5370 (McLean).
- 94. Physical orientation of equipment refers to the mounting location with respect to a set of rectangular coordi- 1 1
notes, its angular position, its location with respect to other l l
. I
i items in the plant and installation into.I.ces. Applicants' 9E Testimony at 4. Physical orientation of electrical equipment generally does not affect environmental qualification. For most electrical. equipment, environmental conditions are identi-cal regardless of the orientation. Physical orientation is ! more likely to be related either to seismic qualification or to i
- operability of the equipment. However, there are circumstances in which physical orientation of electrical equipment could af-fact environmental qualification. For example, if an electro-hydraulic valve operator.were installed upside down, i
hydraulic fluid could potentially leak onto the cable termina-tions -- possibly causing corrosion of the electrical connec-tions. Id. at S.
- 95. The environmental qualification test reports, which 3
are provided by vendors of electrical equipment qualified by testing, describe and/or provide sketches or photographs of the test set-up, including physical orientation of the test equip-> , ment. Orientation is addressed in a variety of ways. The most common method is for the vendor to test the equipment in the i ! most limiting orientation, i.e., the orientation determined by j cngineering analysis to result in the most severe environmental l conditions. In that case, the equipment would be environ-mentally qualified for any physical orientation. The vendor may instead test in a single orientation which is not the most i limiting condition, and either qualify the equipment by
- - - - - = . . . . - . - - - - - . -. . - . - -. --
L I 1 l cnalysis for other orientations or simply specify the test ori- I l cntation as the only permissible orientation. Or, finally, the vendor may test.the equipment in several orientations. Vendors also are required to provide technical manuals containing in-otallation and maintenance instructions. Finally, the vendor i provides mounting drawings which include specific instructions l for-orientation. Applicants' 9E Testimony at 5-6; Tr. 5395 4 (Bucci). This information is sent by the vendor to the respon-sible design organization. Applicants' 9E Testimony at 6; Tr. 5399 (Bucci).
- 96. The design organization reviews the test orientation 4
or orientations against the design drawings which have been j prepared for installation of the equipment at the SHNPP. Ori-i antation during the testing must either be identical to the in-j stallation shown on the design drawings, or the equipment must be able to be qualified by analysis for a different orienta-1 l tion. In addition, the design organization reviews the vendor mounting drawings and technical manuals to make sure that they 1' cre consistent with the qualification test set-up. If there
; cre any discrepancies, inconsistencies or ambiguities concern-4 ing physical orientation of the equipment, further information
, is requested from the vendor as necessary. Any concerns re-
- sulting from this review are documented in the qualification i
rsview package as outstanding items which require resolution prior to considering the equipment environmentally qualified. j I l
i Should resolution of a concern require a change to the instal-lation drawing, a design change notice ("DCN") must be issued. Applicants' 9E Testimony at 6-8; Tr. 5399-400 (Bucci). . 97. CP&L assures that safety-related electrical equipment is installed according to the installation drawings through de-t
, tailed procedures for control of design documents, preparation of installation work packages based on design documentation, installation performed in accordance with work packages and i
work procedures, and quality inspections to verify proper in-otallation. Applicants' 9E Testimony at 8-9. Installation de-cign drawings and documents are transmitted to CP&L's Document Control Center ("DCC"). The construction engineer, following , i written engineering procedures, then obtains the drawing from
; the DCC. The DCC will automatically issue subsequent revi-f sions, DCNs, and field change requests ("FCRs"), to holders of I i controlled drawings. Id. at 9.
! 98. In preparing for the installation of equipment at the i
j SHNPP, the construction engineer prepares a work package that i j generally includes installation design drawings, vendor draw-ings, vendor manuals, process control sheets, and design
! changes in the form of FCRs and DCNs. Id. at 9. Preparation i
) of the work package is governed by a procedure which specifies i that equipment orientation is one of the items which must be i , ] checked. Tr. 5370-74 (McLean). The work package is given to 1 l the field superintendent responsible for installing each piece i ) ! 4 e R
...-...--.4 .-.,,w-...w., . - - . - w., .<w-r,.-.-%,- n-,.,,-.s ------m_ #- --.--m ..-..i-..--.- %.,,, , - - - - - , -..-=- ,r.--., ---,-&v, ,-
I cf equipment. The field superintendent ensures the equipment j is installed according to the design documents. Applicants' 9E Testimony at 9.
- 99. CP&L's quality. inspection / verification program for SHNPP also helps to assure proper installation orientation of ;
cafety-related electrical equipment. Inspection points are '
- cpecified on the process control sheets in the work package.
The quality inspector prepares inspection documents corre-cponding to the process control sheets developed by the con-
- otruction engineers. The inspectors refer to the work packages l when they make their inspections. Physical orientation is one 1
- of the required inspections. Until the inspection points for a i
j piece of equipment are accepted, the installation is not ! nceeptable and the procedural requirements are not satisfied. If there is a discrepancy a nonconformance report is written cnd a " hold tag" is placed on the equipment, which may limit the work that can be performed. Each nonconformance report re-l quires a specific disposition, i.e., rework, repair, scrap, or l cceept-as-is, which requires design engineering approval. Id. j j ct 9-12. I 100. In summary, physical orientation of safety-related i i electrical equipment at the SENPP is controlled from qualifica-I tion testing of the equipment, to installation design, to phys- , ical installation of the equipment in the plant. Mr. j Eddleman's cross-examination of Applicants' witnesses and the j ! 4
-- ,,.-e-.--r ~..m--- -,,-,--e .r.-,_,-ww-
. - -e.--. e- w w ,,-,,,m,----------,...w, . - , , , . , ----,.-n....,,e,-,-, -,vy-,- , - - - - . . . - - - ---
l Staff's witness presented no basis for questioning the adequacy-j of that process. There is reasonable assurance that safety-rolated electrical equipment is installed so that physical ori-cntation of the equipment does not prevent the equipment from b31ng environmentally qualified. 4
- 7. 9F: Lubricants and Seals 101. Eddleman Contention 9F states:
+ The effects of radiation on lubricants
- and seals have not been adequately
. addressed in the environmental qualifica-tion program.
; Applicants' panel for Contention 9F included Mr. Bucci, Mr. Pagan and Mr. Yandow. Mr. Yandow's qualifications are dis-f l cussed supra at 1 53. The qualifications of Mr. Bucci and 1 Mr. Pagan are discussed supra at 1 86.
102. A lubricant is an oily or greasy substance which pro-1 vides a near-frictionless film on two or more surfaces which roll, rub or rotate against each other. Motors, valve opera-f tors and pumps are three examples of safety-related electrical : squipment which.use lubricants. A seal is a device -- static
; or dynamic; metallic or organic -- that prevents foreign sub-i stances from entering equipment or retains a required substance
- within the equipment. Transmitters, valve operators, pumps and RTDs are examples of safety-related electrical equipment which
. have seals. Applicants' 9F Testimony at 3.
.. -. -, . - , . _ - . - - - - . _ . - . , . - . - - .L
103. All BOP (Ebasco supplied) safety-related electrical cquipment for SHNPP which is located in a harsh environment is qualified by test. Equipment which normally contains lubri-cants or seals is tested with those components as part of the cquipment. Qualification testing consists of accelerated ther-mal aging, irradiation, and a design basis accident simulation 1 (if applicable). During the irradiation portion of the testing ; ^ program, electrical equipment is irradiated as a whole, l j including any seals or lubricants. The qualification test re-ports identify the radiation dose to which the equipment is ex-posed. In every case, the radiation exposure of the electrical i squipment during testing exceeds the maximum total integrated i radiation dose to which the equipment could be exposed over its
- qualified life.22/ The required radiation exposure is based on
- normal operating conditions, design basis accident conditions
{ (if applicable), and post-accident conditions (if applicable). (Not all safety-related electrical equipment is located in l areas of the plant which will be~ subjected to accident and/or } post-accident conditions.) Applicants' 9F Testimony at 4-5; Masciantonio at 19. 1 1 J 22/ Although lubricants are required to be qualified for the life of the particular equipment in which they are used, lubri- , cants will be routinely checked at more frequent intervals as ' part of Applicants surveillance and maintenance program. Tr. 5303-04 (Yandow). 1
104. To assure that the lubricants and seals tested are of the same type as the lubricants and seals supplied or recom-tended by the vendor for BOP equipment, Ebasco reviews the ven-dor test reports to identify organic components of the tested cquipment, including lubricants and seals. Ebasco compares the lubricants and seals identified in the test report to the lu-bricants and seals supplied or recommended by the vendor in order to verify that they are the same type. Applicants' 9F Testimony at 5. If there is a discrepancy, ambiguity or omis-sion concerning the identification of a lubricant or seal which was tested, supplied or recommended by the vendor, Ebasco then attempts to resolve the open item by requesting additional information from the vendor. If the vendor cannot demonstrate that the lubricant or seal supplied or recommended is the same type as that tested, corrective action is required to qualify the different components. Any corrective actions must be docu-mented in the environmental qualification package for the equipment. Id.; Tr. 5488-94 (Bucci, Pagan). 105. Some of the NSSS (Westinghouse supplied) safety-related electrical equipment for the SHNPP also uses lubricants cnd seals. With respect to seals, either metallic seals, which are not degraded by the environmental conditions for which olectrical equipment must be qualified, or organic seals, which cre qualified au part of the equipment tested, are used. Applicants' 9F Testimony at 6.
106. Westinghouse, unlike Ebasco, does not identify the cpecific lubricants used during testing. Rather, Westinghouse recommends a general type of lubricant and provides the speci-fications the lubricant must meet to assure operability of the cquipment. Therefore, CP&L has contracted for and received a lubrication study performed for the SHNPP by the Mobil Oil Com-pany, a leading lubricant vendor. The purpose of the study is to identify, for each piece of electrical equipment which re-quires lubrication, the specific brands of lubricants which can be used with that equipment. In the study, the results of ra-diation stability testing are provided. This radiation stabil-ity testing includes standard performance tests which were con-ducted both before and during irradiation to measure the offects of radiation. CP&L currently is reviewing the adequacy of the Mobil study. For each lubricant to be used in a piece of NSSS electrical equipment, the radiation dose received dur-ing lubricant testing will be compared to the total integrated done which the equipment must be qualified to receive at the SHNPP. The radiation dose received during testing must be higher than the dose for which the equipment is required to be qualified.2g/ In addition, the performance of the lubricant during testing will be reviewed to verify that the equipment manufacturer's lubricant performance specifications have been 2p/ See n.27, supra.
I i met. CP&L will develop an environmental qualification package which will document the tests described in the lubricant study,
.cs well as the analysis which apply the test results to specif-ic electrical equipment at the SHNPP. Id. at 6-7.
107. In summary, lubricants and seals in Ebasco supplied BOP safety-related electrical equipment have been evaluated for radiation exposure during qualification tests as components of the electrical equipment tested. Ebasco verifies that the coals and lubricants supplied with safety-related electrical cquipment are of the same type as those tested. Seals in NSSS cafety-related electrical equipment are either metallic seals, which need not be qualified, or organic seals, which are quali-fied as part of the equipment tested. Qualification of lubri-cants to be used in NSSS safety-related electrical equipment, including qualification for radiation effects, is addressed by the Mobil lubricant study. Mr. Eddleman's cross-examination of Applicants' witnesses and the Staff witness presented no basis for questioning the adequacy of Applicants' EQ Program as it , relates to qualification of lubricants and seals for radiation offects. Therefore, there is reasonable assurance that the ef-facts of radiation on lubricants and seals have been adequately cddressed for the SHNPP.
0-l t -
- 8. 90: Type Test Reporting '
[ 108. Eddleman contention 9G states: There is inadequate assurance that failure to report all results of environ-mental qualification tests, including fail-ures, has been brought to light in connec-tion with electrical equipment installed in Harris. This includes past-test failures i of equipment which subsequently passes an
, EQ test and test failures of equipment l which is said to be qualified by similari-l ty. (Ref. Item 2, Page 5, L. D. Bustard et i al., Annual Report: Equipment Qualification ! Inspection Program, Sandia National
! Laboratories, FY83.) For Contention 9G, Applicants' panel consisted of Mr. Prunty, l Mr. Bucci, Mr. Pagan and Mr. Kumar V. Hate. Mr. Prunty's qual-
! ifications are discussed supra at 1 53. The qualifications of I
h Mr. Bucci and Mr. Pagan are discussed supra at 1 86. Mr. Hate 1 j 10 Principal QA Engineer for the S!!NPP in the CP&L Corporate i i j Quality Assurance Department, which has overall responsibility i jl for vendor qualification. Applicants' Testimony of Robert W. l l Prunty, Richard M. Bucci, Edwin J. Pagan and Kumar V. Hate in l I Response to Eddleman Contention 9G (Type Test Reporting), ff. j ~ l ) Tr. 5515 (hereinafter " Applicants' 9G Testimony"), at 2; Tr. ,
- 5551-52 (Hate). Prior to being assigned to the onsite Quality Acsurance ("QA") organization at the SHNPP, he was responsible j for directing a team of QA engineers who, among other things, i
j ccnducted audits of vendors and assisted in program planning of j v ndor qualification and inspection activities. Applicants' 90 { Testimony at 3. 4
- i 1
. ~ _. _ , . - _ - _ . . _ _ _ . _ , _ . . - . . _ . _ . _ _ . , . , . _ _ - . _ _ _ , , . . - . - . ..,-,,.....,_.._.._,._.,_.._..,y.__
l l [ 109. The " Annual Report: Equipment Qualification In- ) j cpection Program" prepared by L. D. Bustard, et al., Sandia Na-l tional Laboratories (FY1983), referenced in Contention 9G,was
- transmitted in a memorandum dated February 2, 1984 from William i
J. Dircks, NRC Executive Director for Operations, to the Com- ' nissioners. Item 2, page 5 of this Annual Report 29/ states as l l follows: i
; Another company started to qualify a prod- ) uct line by testing five different products in that line. By completion of the test program, four of the products had substan-tially degraded. A qualification report ; was written describing only the successful ; qualification of the one product that did i not degrade. A second qualification report I was then generated arguing that other mem-I bers of the product line were qualified by i similarity. The degradation observed dur-
! ing testing for four members of the product i line was never discussed in the similarity j- report. Interestingly, the one product
- that successfully performed throughout this test had substantially degraded during pre-
- vious qualification attempts. These previ- t i
ous efforts were never mentioned in the ! qualification report. The qualification { test parameters had been successively changed until qualification success was achieved. ] Applicants' Contention 9G Testimony at 4-5. 110. An attachment to the Dircks memorandum, entitled
" Discussion of Sandia Items," identifies a number of NRC in-l cpection reports potentially relevant to " item 2." Examination 1
29/ The scope of Contention 9G is limited to this particular i item in the Sandia Annual Report. Tr. 5662-64. l t i l
l l L cf the inspection reports reveals that " item 2" in based on In-l
- cpection Report 99900277/83-02, which documents the results of an inspection of the Rockbestos Company conducted on June ,
20-23, 1983. The inspection report questions the use of Rockbestos environmental qualification test report QR 2806 to i qualify Rockbestos' entire 100 series line of coaxial, triaxial cnd twinax cables. The inspection report notes that QR 2806 o enly demonstrates qualification (by test) for RSS-6-104 coaxial cables. Furthermore, during the same test used to show quali-fication of RSS-6-104 cables, other cables (namely, RSS-6-100A, i i RSS-6-109 RSS-6-llO and RSS-6-ll2) failed electrically. This
, fact is not mentioned in Rockbestos similarity discussions for other cables. Applicants' 9G Testimony at 5.
111. As a result of the above-described deficiency and l other deficiencies identified during inspections of Rockbestos, the NRC Staff issued IE Information Notice No. 84-44 (June 8, 1984), which notified licensees of potential generic problems regarding Rockbestos environmental qualification testing of f cafety-related electrical cables. Applicants' 9G Testimony at 5-6; Masciantonio at 21. The Staff, in Information Notice 84-44, suggested several possible ways in which qualification ! of Rockbestos cables could be demonstrated: a) Perform a valid qualification test of the installed Rockbestos cables.
- b) obtain documentation from other avail-able qualification tests already per-
- formed and' determine its applicability to the installed cables.
i f 1
.._ __ L,_ . _ . _ .. . _ _ _ _ . . - . , . _ . _ , _ _ _ _ . , _ _ _ . . . . _ . . . - . _ , _ _ .
l c) Perform analyses of existing qualifi-l cation reports applicable to the in-stalled cables to ensure that the doc-umentation relied upon to demonstrate environmental qualification supports such a conclusion. Masciantonio at 21, Attachment 3 at 2. 112. The following types of vendor-supplied Rockbestos cable 30/ are used in safety-related applications at the SENPP: RSS-6-104/LD coaxial cable, RSS-6-105/LD coaxial cable, RSS-6-108/LD triaxial cable, Firewall III insulated thermocouple cable, and Firewall III insulated control cable, i (The SHNPP does not use any of the Rockbestos cables identified in the inspection report as having failed qualification tests.) The RSS-6-105, RSS-6-lO8 and Firewall III insulated thermocouple cables all arc used as pigtails, which are approx-I imately three feet long, in electrical containment penetra-tions. The control cable is used as jumper wirec, each only a fcw inches in length, in the limit switch compartments of Lim-itorque valve operators. Only the RSS-6-104/LD RMS cable is 4 installed in the SHNPP cable raceway system. Applicants' 9G Tcstimony at 6; Applicants' Supplemental Testimony of Robert W. i Prunty, Richard M. Bucci, Edwin J. Pagan and Kumar V. Hate in R3sponse to Eddleman Contention 9G (Type Test Reporting), ff. Tr. 5515 (hereinafter " Applicants' Supplemental 9G Testimony"), ! ct 4. 30/ Rockbestos is not a direct cable vendor at the SENPP. Applicants' 9G Testimony at 6.
c l 113. Applicants have chosen the second alternative i specified in Information Notice 84-44 in order to demonstrate cnvironmental qualification of the Rockbestos cables used at the SHNPP, i.e., to obtain documentation from qualification tests performed on Rockbestos cables by vendors or test laboratories other than Rockbestos. Applicants have obtained two test reports, IPS-1053 and IPS-1054, from Conax Corporation ("Conax")31/ which describe environmental qualification testing cf electrical penetration module assemblies, including R:ckbestos RSS-6-105/LD coaxial cables. Applicants have re-viewed these reports and have determined that the qualification t0st parameters envelope applicable SHNPP parameters for the warst case location through which Rockbestos coaxial and triaxial cables are routed. Applicants' Supplemental 9G Testi-osny at 3. , 114. The qualification testing of RSS-6-105/LD cables is cpplicable to the other coaxial and the triaxial cables used at the SENPP as well. The RSS-6-104/LD and RS-6-105/LD are both ceaxial cables, have the same electrical, physical and environ-COntal properties, and are of identical construction. Their cenductor', s insulation, shields and jackets are the same 31/ related conduit seals, is a directConax, vendor as a supplier at the SENPP.of safetys Conax quality assurance ("QA") program has been reviewed by CP&L and has been found accept-able. Ebasco also has reviewed Conax's QA program and found it cceeptable. Applicants 90 Testimony at 4. G
. . .. . _ . . _ _ ~ _ . - - - __- ---
f. 1 i ' materials. The only difference is that the RSS-6-105/LD has an 1 inert coating applied between the shield and the insulation to *'
! improve electrical noise reduction properties. This coating is cpplied after the insulation has been extruded on the conductor '
i , cnd does not affect the properties of the insulation material. i The RSS-6-108/LD is a triaxial cable which also uses the same , caterials, and which has two shields instead of one, as is the case with the two coaxial cables identified above. Since it is I also of concentric construction, the arrangement of the compo- ! nents is sufficiently similar to that of the coaxial cable to l l permit its qualification. Further, the dimensions of the ingu-lation and Jackets of the RSS-6-108/LD are greater than those ! i ]* of the RSS-6-105/LD. For qualification purposes and for a ' l given cable type, a thinner insulation and jacket thickness can l 1 j .be used to qualify a thicker insulation and jacket thickness of f i l the same materials. As such, the RSS-6-105/LD can be used to i qualify the RSS-6-108/LD. In short, the minor differences 1
/ among these cable types do not affect qualification. Appli-(
j cants' 9G Testimony at 6-7; Applicants' Supplemental 90 Testi- ; 4 cony at 3-4, ] 4 1
! 115. Applicants also have obta'ned two test reports which ,
i l describe environmental qualification research tests by Sandia \ ' j National Laboratories on Rockbestos Firewall III insulated con-trol cable. The Rockbestos control cable used at the SHNPP was i i cne of tha cable types tested. Those test reports are: i
NUREG/CR-2932, 1 of 2, " Equipment Qualification Research Test
. cf Electric Cable with Factory Splices and Insulation Rework Test No. 2" (September 1982); and NUREG/CR-3588, "The Effect of LOCA Simulation Procedures on Cross-Linked Polyolefin Cable's Performance" (April 1984). Applicants have reviewed these re-ports and have determined that the qualification test parameters, in each test, envelope applicable SENPP parameters for the worst case location for both the control cable and thermocouple cable. Further, the control cable is representa-tive of the thermocouple cable for qualification purposes, cince the insulation materials and all other construction fea-tures significant to environmental qualification are the same.
The thickness of the insulation material on the thermocouple cable is 25 mils compared to 30 mils on the control cable. However, the thermocouple cable wires are covered by a metallic chield and Hypalon overall jacket which more than compensate for this minor difference in thickness. Applicants' supplemen-tal 9G Testimony at 5-6. In addition, the control cable was cubjected to an extremely high voltage (480 volts) during testing, compared to the low voltage to which the thermocouple cable will actually be subjected in service (millivolts). Tr. 5660-61 (Pagan, Bucci). 116. In conclusion, Applicants have qualification test data independent of Rockbestos which demonstrate the environ-mental qualification of the Rockbestos cables to be used in the A
l l' ! 8HNFF. This approach is,, consistent with the Staff's sugges-tions set forth in IE Information Notice 84-44. Tr. 5585 l l (Masciantonio). There is reasonable assurance that the Rockbestos cables used in the SHNPP are environmentally quali- l fled to perform their safety functions. E. Eddleman Contention 11: Polyethylene Cable Insulation [ Board to insert decision granting summary disposition. See Tr. 2167; Applicants' Motion for Summary Disposition of ; Eddleman Contention 11 (May 25, 1984); NRC Staff Response in ! Support of Applicants' Motion for Summary Disposition of Eddleman's Contention 11 (June 18, 1984); Eddleman's Response to Summary Disposition on Eddleman Contention 11 (Cable Insula-tion Degradation) (June 29, 1984).] F. Eddleman contention 41: Pipe Hanger Welding i 117. Eddleman Contention 41 states: Applicants' QA/QC program fails to assure that safety-related equipment is properly . inspected (e.g. the "OK" tagging of defec-tive pipe hanger welds at SHNPP). This contention was admitted by the Board in its Memorandum and ! Order (Reflecting Decisions Made Following Prehearing Confer-Cnce) of September 22, 1982. LBP-82-119A, 16 N.R.C. 2069 ! (1982). In ruling on this contention, the Board rejected the view that the entire QA/QC program be subject to litigation l
-73 !
4 l l
r i under the ambit of Contention 41, and limited the contention to I. address Mr. Eddleman's specified concern "that there exist de-factive hanger welds that have been improperly inspected and e cpproved." M . at 2097. 118. Applicants' direct testimony in response to conten- } -tion 41 was presented by a panel of four witnesses. See Appli-cents' Testimony of James F. Nevill, Alexander G. Fuller, David R. Timberlake and Kumar V. Hate in Response to Eddleman Conten-l. 4 tion 41 (Pipe Hanger Welding), ff. Tr. 6663 (hereinafter "Nev-j ill et al."). Mr. Nevill is a Principal Engineer - Civil in ? l the Harris Plant Engineering Section and is responsible for two Civil sub-units which perform the following functions with re-I cpect to pipe-hangers: resolution of identified field prob-Icms, design of new pipe supports due to pipe / system changes,
- cnd stress analysis evaluations associated with field changes.
] Mr. Fuller is a Principal Engineer - Mechanical (Hanger Engi- ) n ering) in the Harris Plant Construction Section. From Jcnuary 1981, through September 1983, he was the lead in the
- H:nger Engineering group responsible for the technical support I of pipe hanger installation and was also responsible for the i revision of procedures for hanger installation and the resolu-i tion of nonconformances involving hanger installations. Since October 1983, when the Hanger Engineering group was reorganised, Mr. Fuller has been responsible for technical sup-part of hanger installation. Mr. Timberlake is also employed 1
in the Harris Plant Construction Section as a Senior Engineer - , i Metallurgy / Welding (Welding Engineering). As such, he has been ' responsible for the review of pipe hanger sketches from the Ctandpoint of welding requirements for field fabrication and for assigning welding procedures, filler metal and mandatory inspection holdpoints on Seismic Weld Data Reports, as well as cupplying additional welding instructions as needed. Mr. Tim-berlake has also been responsible for resolving field-related welding problems, and has provided training to Quality Control-Welding inspection personnel and craft personnel. Finally, Mr. Timberlake is responsible for maintaining and, as needed, revising the field welding procedure for pipe hangers. Mr. Hate is a Principal QA Engineer in the QA/QC Harris Plant Sec-tion. This section is responsible for performing the following functions to assure that the hanger program is adequate and complies with regulatory requirements: review of construction cpecifications, procedures and documentation; weld inspections; QA surveillances; and nonconformance identification / resolution. Nevill et al. at 1-4. 119. The NRC Staff also sponsored a panel of witnesses in response to Eddleman Contention 41. See NRC Staff Testimony of Paul R. Bemis, George A. Hallstrom and Jerome J. Blake on Eddleman Contention Number 41, Pipe Hanger Welds, ff. Tr. 7217 (hereinafter "Bemis, Ha11strom and Blake"). Mr. Blake is the Section Chief of the Materials and Processes Section in
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-.------r- -
Region II. In this position he is responsible for coordinating , and overseeing engineering inspections and technical evalua-tions in the areas of welding, metallurgical engineering, - l nondestructiva examination, failure analyses, mechanical engi-neering and design, and Inservice Inspection and Testing of Re-cctor Plant systems and components. Mr. Hallstrom is a Reactor
- Inspector in the Region II Materials and Processes Section.
- His duties have primarily involved inspections related to fab-I rication, inspection and testing of nuclear components and sys-1 l tems with particular attention directed to welding'and nondestructive examinations. As a specialist, Mr. Hallstrom provides assistance to other members of the NRC Staff concern-
- ing conditions arising during construction, inservice in-1 I
cpection, or operation of nuclear facilities which require a knowledge of welding and/or destructive examination. Until re-cently, Mr. Bemis was the Section Chief, Projects, Section 1C at Region II and had direct responsibility for the inspection cnd enforcement program at the Harris plant. Bemis, Hallstrom cnd Blake at 1-6; see also, Tr. 7216-17 (Bemis). L 120. In addition to the witnesses sponsored by Applicants cnd the Staff, three employees of Carolina Power & Light Com- ! pany at the Harris site (Kenneth A. Douglas, William H. Pere ) l cnd Gene G. Tingen) and the NRC Resident Inspector at the r Herris site (George A. Maxwell) appeared voluntarily in re-cponse to the Board's grant of Mr. Eddleman's August 17, 1984 l l
.. . -- . - . .- _ . - - - _ _ ~ _ - - _ _ . _ . - -
L request that these individuals be subpoenaed to testify at the hearing. Mr. Douglas is employed in the QA/QC Harris Plant Section as a Quality Assurance Specialist, responsible for per- l l forming QA surveillances on pipe hangers. Tr. 7086 (Douglas). Mr. Pere is a QA/QC Specialist in the QC Structural Welding In-cpection Unit and has been performing QC inspections of pipe - l hanger welding for over four years. Tr. 7087-88 (Pere). Mr. Tingen is also employed by the QA/QC Harris Plant Secti'on and was previously responsible for performing QC inspections of pipe hanger welding. Tr. 7088 (Tingen). Mr. Maxwell, present-j ly the Operations Resident Inspector, was previously the Con-ctruction Resident Inspector at the Harris site and, as such, observed welding inspections of pipe hangers. Tr. 7234 1 i (Maxwell). f 121. Prior to addressing the merits of Contention 41, a brief discussion regarding the scope of the direct testimony j and the testimony adduced on cross-examination in in order. As I
- noted by counsel for Applicants in his opening statement on j
i this contention, Applicants' pre-filed direct testimony is pro-f grammatic in nature, identifying problems encountered over the j years relating to pipe hanger welding and the corrective and
- preventive actions taken in response.32/ It was necessary for .
i s j
- 32/ While Applicants' testimony did not delve deeply into the '
j details of the pipe hanger welding problems, it is clear that I (Continued Next Page) i i E
,c,..___,..,,n--,n.c . . . , . . - <
-, ..-._,_..--m.- ,_ ___. _
,m _c-*_ ,,m_,,____._.,__.,,, ,,,,-.,,,_w ,,,.% .
= - _. -- - - --- - - . . - -
I Applicants to present this type of testimony due to Mr., i L 'Eddleman's failure to identify, in response to discovery by l
- Applicants, the specific concerns he had with the pipe hanger welding program. See Tr. 6652-57 (Baxter); Applicants' Inter-f~
rogatories and Request for Production of Documents to Wells Eddlemanf(Contention 41), dated April 2, 1984 (which Mr.
- Eddleman did not answer).33/ Further, it soon became' clear 1
that Mr. Eddleman's cross-examination suffered from a similar j lack of focus, leading the Board to require that Mr..Eddleman provide an explanation of what he viewed as the problems with
~
the pipe hanger program. Tr. 6788-90, 6826-27 (Kelley). f (Continued) Applicants' witnesses were intimately involved in the pipe i hanger program and were qualified to address the wide range of 4 pipe hanger welding issues, both specific and general, raised by Mr. Eddleman on cross-examination. See generally Nevill et nl. at 1-4; Tr. 6671-7057. i 33/ At the contention pleading stage, intervenors need not i cdvance detailed evidence to meet the bases with specificity i requirements of 10 C.F.R. I 2.714(b). Houston Lighting & Power i Company (Allens Creek Nuclear Generating Station, Unit 1), 3 ALAB-590, 11 N.R.C. 542, 547-48 (1980). However, the Appeal Board has held that, while applicants carry the ultimate burden l of proof, intervenors also bear evidentiary responsibilities. ! Further, discovery requests, such as those filed by Applicants, i which seek to discover what (if any) evidence underlies an in-tervenor's own contentions are proper and, indeed, necessary if ' Applicants are to be in a position to effectively discharge their burden of proof. Pennsylvania Power & Light Company l (Susquehanna Steam Electric Station, Urits 1 and 2), ALAB-613, i 12 N.R.C. 317, 338, 340 (1980). , i i I I i.
4 a 122. We turn now to a consideration \of the issues raised by/C$ntention 41. ' A Seismic Class I pip'e hanger is a component rQ \ , or etructural assembly designe,d and installed to support or re-6 f
'ctrain a section of pipe subjected to a combination of loads, s cnd which protehts the pipe from stresses that could impair the pipe's ability to perform its function. The design of pipe hangers (a based upon a stress analysis of the operating, ther-mal and seissic' loads imposed upon aspiping arrangement. The hangers are then designed to counteract these combined loads end to prevent the pipe from being overstressed. For the Harris Plant, Bergen-Paterson was the primary design and fabri-cation organization for the pipe supportcx, Nevill et al. at s-6.
123. Pip [hangerweldsaresubjectedtothreebasictypes of weld quality. inspections:34/ in-process and final inspection
^
) > e is performed on shop welds by both Bergen-Paterson inspectors cnd Ebasco Vendor QA representatives prior to shipment from Bergen-Paterson; shop welds are also receipt inspected by the
/
CP&L QC organization when th hangers are received on site;35/ 34/ The QC Section performs welding inspections of pipe hang-era at the Harris site. Pipe hangers are also subjected to in-Epections of non-welding attribu tes, such as location, materi-cl, etc., whics are performed by the CP&L Construction Inspection (CI)' organization. Consideration of these CI in-spections was deemed outside the scope of Contention 41, which focused on hanger welding. See, e.g., Tr. 7263-68. 35/ Prior to June 1982, only a limited number of shop welds (i.e., those made by Bergen-Paterson) on pipe hangers were in-i '( s (Continued Next Page) g, u k-.% 4 8 .b 9 -- . . .. .. -. . - - - . - . -. . - - . -
cnd, field welds receive a final QC welding inspection after the hangers have been welded in place.3s/ See Nevill et al. at 6-11. These various inspections will be discussed in greater i detail below as they relate to the pipe hanger welding problems I encountered at the Harris Plant and the corrective actions in-stituted by Applicants in response to those problems. 124. Nonconforming conditions identified by these in-spections (or by any other method) are required to be docu-mented and tracked for resolution. Documentation of deficiencies, such as rejectable weld conditions, is governed by procedure CQA-3. In brief, isolated or less significant deficiencies would be documented in a " subordinate" nonconformance report, such as a Seismic Weld Data Report. Significant adverse conditions are typically documented in a formal Nonconformance Report ("NCR," formerly referred to as a DR, NCR or DDR), c'o pies of which are routed to upper site (Continued) cpected upon receipt. Due to the discovery of a large number of deficiencies, Applicants instituted a 100% inspection of chop welds which was retrofit to include hangers received on cite prior to June 1982. Nevill et al. at 13. t 3s/ The initial hanger erection program involved a preliminary l (Phase I) inspection and a final (Phase II) inspection. Phase I inspections were conducted when the hanger installation was I not yet complete. Phase I inspections provided early indica-tions of potential problems, as evidenced by the 1980 and 1982 l reinspection programs, both of which involved findings from l Phase I inspections. See Nevill et al. at 7-8. l l l l
m2nagement. Tr. 6678-80 (Hate), 6684-87 (Hate, Fuller), 6690-93 (Hate), 7031-32 (Hate). 125. Installation of pipe hangers at the Harris Plant l bzgan in early 1979; QC welding inspections of the hangers were instituted shortly thereafter. Tr. 6673 (Fuller); see n.36, Eupra. In September 1980 during a routine inspection, the NRC Rcsident Inspector identified several Bergen-Paterson pipe hnnger drawings with unclear and incorrect weld symbols, as wall as several cases in which the field weld was different from the drawing requirements without the discrepancy having bnen identified by QC.32/ The results of this inspection were cited as an infraction in an inspection report dated November 3, 1980 (and was subsequently closed out by the Staff on Snptember 14, 1981). Nevill et al. at 15-16; Tr. 7050-52 (Fuller). 126. Based upon the identification of these deficiencies, Applicants conducted a site investigation of a selected sample of pipe hanger drawings and of installed or partially installed hengers.31/ This investigation identified a wide scope of daficiencies, beyond those identified by the NRC, and led to a 32/ The NRC inspector did not identify any rejectable weld daficiencies during this inspection. Tr. 7051 (Fuller). 33/ Contrary to Mr. Eddleman's implications, the Resident In-cpector did not force CP&L to undertake this investigation. As Mr. Maxwell explained, he encouraged the investigation, but "not very much" prompting was needed. Tr. 7248 (Maxwell). I. l I 100% reinspection of hangers released to the field for instal-l L lation, as well as a 100% in-house review of hanger design drawings released to the field for installation. Nevill et al. at 14-15; Tr. 7050-52 (Fuller), 7245-47 (Maxwell); Eddleman Ex. 58, Attachment at 1, 3. The results of this reinspection ef-fort were documented and reported to the NRC in accordance with the provisions of 10 C.F.R. 6 50.55(e). Nevill et al. at 15; nee, generally, Eddleman Ex. 58. The reinspection identified numerous problems with the pipe hanger design drawings involving unclear, missing and incorrect weld symbols. These problems were reported to Bergen-Paterson and resulted in the issuance of new (corrected) drawing revisions. The reinspection also documented welding defects on a number of pipe hangers. These defects were.either reworked or accepted based upon engineering analyses performed by the design organi-zation.39/ Eddleman Ex. 58, Attachment at Exhibit 2; Nevill et nl. at 16. 127. In response to the problems identified by this reinspection, Applicants undertook a number of corrective ac-tions to remedy the causes of the identified discrepancies. i l { 39/ In most cases, Applicants chose to rework defective welds l rather than undertaking an engineering analysis to accept them l es-is. This decision was based on the fact that, generally, t raworking a weld is easier and more cost-effective then per-forming an engineering analysis. Tr. 7330-31 (Bemis); see also, Nevill et al. at 22-23. _ _ _ . . _ . _ _ ~ . . _ , _ _
c- a With respect to drawing errors, Hanger Engineering and Welding l L ' Engineering personnel began reviewing pipe hanger design draw-i l ings prior to their issuance to the field. Drawings with' weld i-l cymbol errors were thus identified and returned to l Ebasco/Bergen-Paterson for correction before they could cause i potential problems in the field. Nevill et al. at 16. Fur-
.ther, Bergen-Paterson also revised their drawing review proce-
.dures such that all design drawings were routed through a sin-gle office to assure review consistency. Id.
128. Applicants also conducted additional training classes for both craft and. inspection personnel (welders and craft supervisors and QC welding inspectors). These classes con-l sisted of training on AWS standard weld symbol nomenclature and emphasized the importance of welding and inspecting the hangers in accordance with the hanger design drawings. Both craft and 9 QC personnel were also instructed to return design drawings to Henger Engineering if they discovered drawing errors or if the hengers could not be installed as designed.40/ Nevill et al. 40/ On cross-examination, Mr. Eddleman attempted to prove his allegation that retraining classes in general were inadequate
- due to the length of time that training was conducted and be-ccuse the same subjects had been covered during the normal, on--
going training classes. See Tr. 6858, 6934-37 (Eddleman). However, the witnesses' testimony is clear that while on-going l training would cover the same general subjects, the special 1 training conducted after discovery of problems emphasized the nature of the problems identified and what should be done to prevent future occurrences, and that sufficient time was de-voted to the retraining. See Tr. 7124, 7129, 7193 (Pere, Tingen), 7052-53 (Timberlake); see also, Nevill et al. at 17.
~83-
i et 17; Eddleman Ex. 58, Attachment at 2. 129. A second pipe hanger reinspection program was insti-tuted in 1982 upon the discovery by Applicants of three differ-ent problems: documentation errors and field weld defects; de-L ficient shop welds made by Bergen-Paterson;41/ and, the use of an improper technique for the measurement of skewed tee fillet walds.42/ Nevill et al. at 18. These three problems were dis-tinct from those identified in 1980, were not included in the scope of the 1980 reinspection effort, were not covered by the 1980 corrective actions (as discussed in greater detail below), and thus led to the implementation of a second' reinspection of installed hangers.43/ The reinspection found: (1) missing and undersized shop and field welds; (2) minor shop and field weld 4 defects; and (3) inaccurate and incomplete QC weld documenta-
~
tion. Id. 130. The problem with improper measurement of skewed tee fillet welds was identified in 1982 by CP&L site personnel due 41/ The deficient shop welds were discovered by CP&L QC per-sonnel during receipt inspections. See Eddleman Ex. 46, At-tschment at 1; Nevill et al. at 18. 42/ Each of these items was reported to the NRC in accordance with 10 C.F.R. 5 50.55(e). See Eddleman Exs. 22, 41, 46 and j 47. 43/ The 1982 reinspection consisted of inspecting all welds (chop and field) on the entire hanger, not just the types of walds (i.e., skewed tee fillet welds and shop welds) for which l daficiencies had been identified. Tr. 7126 (Pere).
l l l l to reporta received regarding undersized skewed tee fillet' i welds at other plants. As explained by Applicants' witness l Timberlake, clear criteria for the measurement of skewed tee l l fillet welds were not included in the 1975 edition of AWS D-1.1 L utilized at the Harris site. These criteria were added in a subsequent revision to the Code. Tr. 6945-48 (Timberlake). After identifying this problem, additional training was pro-vided to craft and QC welding inspection personnel which uti-lized a figure from the 1981 edition of AWS D-1.1 that clearly outlined the proper techniques for measurement of skewed tee fillet welds. When this technique was implemented by CP&L, un-dersized field welds were also identified. Tr. 6947 (Tim-1 berlake); see also, Nevill et al. at 18-20; Tr. 7125 (Pere, Tingen and Douglas). 131. The second problem involved the identification by CP&L personnel of defects on shop welds made by Bergen-1 Paterson. The cause of this problem was determined to be the , use of different weld acceptance criteria by Bergen-Paterson than were utilized at the Harris site.44/ Nevill et al. at 17-20. In order to remedy this situation, Applicants imple-i l mented a revised set of weld acceptance criteria -- based in 44/ Bergen-Paterson, as a supplier to many nuclear utilities, utilized ASME Section III, Subsection NF as the basis for their acceptance criteria, while Applicants utilized AWS D-1.1 as their acceptance criteria. Tr. 7034 (Timberlake). i
i l l part on the criteria used by Bergen-Paterson -- to be used for l the inspection of both shop and field welds in order to assure I consistency of inspection.45/ Additional training on the re-l l vised weld acceptance criteria was provided to QC weld in-l
- cpectors. Id. at 20; Eddleman Ex. 46, Attachment at 2; Tr.
7033-34 (Nevill, Hate, Timberlake). Although this revision modified, to some extent, the visual acceptance criteria set forth in AWS D-1.1, the general provisions of that Code relating to workmanship, technique and other non-destructive examination criteria were maintained and the modification was compatible with Applicants' previous regulatory commitments. , Eddleman Ex. 30; Tr. 7036 (Nevill, Hate). Additionally, the Ebasco Vendor QA program at the Bergen-Paterson facility was significantly upgraded to include both in process and 100% final visual weld inspections by Ebasco Vendor QA representa-tives. Applicants also instituted a 100% receipt inspection of shop welds on pipe hangers. Nevill et al. at 20; Tr. 7027-29 i (Hate). The cause of the minor field weld defects identified was the reliance placed on the qualifications of QC inspectors and their acceptance of welds as the final word, without .i 45/ The revised inspection criteria were set forth in Field Change Request (FCR)-H-979 (Eddleman Ex. 30). See also, Eddleman Ex. 31 (Justification for FCR-H-979). FCR-H-979 has since been superceded under the enhanced program (see 11 133-138, infra) by CAR 2165-A-003. See Bemis, Hallstrom and Blake at Exhibit 1; see also proposed Applicants' Exa. 27 and 28. l l
. . - . . . , _ . _ , . . _ _ . . _ . . _ . , . _ , _ . . _ , _ _ . . _ . . ._,.m
routine checks or surveillance of their work. To resolve these problems, procedure NDEP-605 (Eddleman Ex. 39) was issued to provide specific guidance to QC welding inspectors on condi-tions governing pipe hanger weld inspections.4s/ Tr. 6964-66 (Hate, Timberlake); Eddleman Ex. 41, Attachment at 3. Finally,
.a program of routine audits of each QC inspector's field work by QC supervisors was implemented in order to provide a double-check on the quality of inspections. Nevill et al. at 20; Tr. 7056 (Hate).
132. The third area of concern identified in 1982 related to documentation errors caused by minimal review of weld records and the absence of a procedure for standardizing the requirements for completing documentation records. Nevill et 3 1. at 19. To resolve this problem, procedure QCI-19.3, Seismic Pipe Hanger Documentation System (Eddleman Ex. 25), was issued to provide consistent guidance for completion and review of pipe hanger weld inspection documentation. Nevill et al. at 20; Tr. 6773-74 (Hate), 6969-70 (Fuller); see also, Eddleman Ex. 41, Attachment at 2; proposed Applicants' Ex. 28, Attach-ment at 2. l 4p/ Previously, visual inspection of hangers was governed by
- NDEP-601, which was written primarily to cover weld inspection criteria for ASME-Code class piping, rather than structur-al/ hanger welding. Tr. 6964-65 (Timberlake).
l _ _. - . _
~
133. Despite the efforts undertaken by Applicants, d3ficiencies related to pipe hanger erection continued to be identified. A stop work order was issued in July 1983 as the l rasult of a QA surveillance; this stop work order required that final inspections of pipe hangers by both CI and QC be discon-4 tinued.47/ During this time, site management reviewed the problems that were identified by that surveillance, by subse-quent Hanger and Welding Engineering surveillances, past NRC inspection reports and previous nonconformances. The need to hsve a system that woul'd stand up to constant scrutiny was also considered. In December 1983 a completely restructured pipe henger program was implemented.4B/ This new pipe hanger pro-gram, referred to as the " enhanced program," has proven to be quite effective in limiting the number of hanger welding
! dnficiencies. See 1 140, infra.
134. The enhanced program for pipe hangers includes new, and revised, procedures intended to clarify the installation and inspection requirements; further, the enhanced program pro-l vides for engineering / technical support to the craft before and i during the installation process in order to resolve potential j 47/ Although final inspections / approvals were not undertaken until the new program was in place, QC welding inspections con-tinued. Tr. 7327 (Bemis). 4g/ Contrary to Mr. Eddleman's expectations, there was no par-ticular NRC finding which prompted this restructuring. Tr. 6790-91 (Fuller). i i t I __ - .-- - , - - . - , . , , , . . . - ,. ..,..-...,_e ..%- _.w .._-..,-m , _ . , , - . _ , _ , , _ _ - . . -
problems prior to turning the hangers over for final in-cpection. See, generally, Nevill et al. at 23-25; Bemis, , 1 Hallstrom and Blake at 11-12. The major improvements of the cnhanced program as they relate to pipe hanger welding are sum-marized below. 135. First, pipe hanger work packages are reviewed by a work package group prior to issuance to the field. At this time, the hanger design drawing is " weld mapped," i.e., each joint to be welded is given a specific identifying number, thereby precluding the possibility that a joint would not be identified or would be confused with another joint during the inspection process. This weld mapping process has been ret-rofit to previously inspected hangers as well as in-process hangers. Nevill et al. at 23; see also, Tr. 6915 (Fuller), 7189 (Douglas). 136. Generic engineering documents are, for the most part, no longer used as solutions to common problems. Instead, field modifications are written for each hanger detailing necessary changes due to these problems. This has greatly reduced the potential for misinterpretation and subsequent misapplication of construction requirements. Nevill et al. at 24; Tr. 6792 (Fuller), 7189 (Douglas). 137. Additional engineering and technical resources have been provided to support the craft. A field hanger engineering support unit has been developed whose purpose is to support the l l craft during hanger installation. These. Hanger Engineering . j personnel remain in the field throughout the hanger's construc-tion and identify and resolve installation problems. These ef-l forts produce additional confidence that the design organiza-i tion's intent is being met during construction. Welding , Engineering personnel examine hanger welds (both shop and i field) prior to submitting the hanger package to QC for final weld inspection. ' Finally, in addition to the QC final review, a Hanger Engineering final review group has been formed to re-view seismic hanger packages prior to final turnover to the permanent QA records vault. Hanger package documentation is l thus verified as being complete and accurate. Nevill et.al. at l 24-25; Tr. 7038 (Hate), 7358-59 (Blake). 138. Although not formally a part of the enhanced program,
. in 1983 the visual weld acceptance criteria were also revised; by providing well-defined criteria for inspecting weld attributes, the potential for conflicting judgments based upon j personal interpretation has been lessened. Tr. 7159-60 (Douglas), 7324-27 (Maxwell, Bemis). The revised acceptance criteria have been reviewed and accepted by the Staff.49/
Bemis, Hallstrom and Blake at 14-16; Tr. 7334-36 (Hallstrom). i 1 l I 49/ The revised criteria are being used for weld inspections under the enhanced program. Tr. 6792-93 (Fuller), 7320-21 (Bemis, Blake) l
- . . - - . = . .
l 139. It was the consensus view of the witnesses who testified in response to Contention 41 that the enhanced pro-j- gram is successful and has provided the required degree of as-l nurance that pipe hanger welding is being properly performed I and inspected. See, aenerally, Nevill et al. at 26; Bemis, Hallstrom and Blake at 13, 17; Tr. 7038-43 (Hate, Fuller), 7188-90 (Douglas), 7356-58 (Bemis). Indeed, the Staff witness-es testified that they had no doubts about the efficacy of the enhanced program and that what, in the Staff's view, was the root cause of the historical problems with pipe hanger welding has now been addressed. Tr. 7322-26 (Hallstrom, Blake, Maxwell, Bemis), 7357 (Bemis). 140. These opinions regarding the effectiveness of the en-hanced program appear to have been borne out, based upon the quality of the work being presented for inspection, results of CP&L QA surveillance audits and NRC inspections. For example, in the second quarter of 1984, approximately 93% of the work presented by the craft to QC for weld inspection was found to ba acceptable. Nevill et al. at 25. Further, QA surveillance of-QC final accepted hangers during the period from January. through October 1984 indicated an acceptance rate of 98.78% for l QC-inspected weld attributes.50/ Tr. 6670-71, 7041-42 (Hate); l 50/ Any rejectable deficiencies identified during QA surveil-lances have been minor in nature and have no safety signifi-cance, but are being reworked. Tr. 7027 (Fuller), 7041 (Hate). 4 91-
. - - --------g -m-gw .-iemu .w,-,e sm-- y- ---*we- w - mg- e- - -~ - e---'
l l See also Nevill et al. at 25. 141. The QA surveillance of pipe hangers is undertaken cnly after a hanger has been final accepted by QC. Tr. 7192 (Douglas). Unlike QC welding inspections, which are performed on all hangers, surveillances are done on a sampling basis. This sample size (approximately 500 hangers), based on the guidance of Mil. Standard 105-D, will provide the necessary confidence level to assure the inspection program is func-tioning properly. Tr. 7039-40, 7047 (Hate). Recent NRC in-spections have reinforced these QA surveillance findings. Bemis et al. at 13; Tr. 7218-19 (Hallstrom), 7357-58 (Bemis). i At present, approximately 150 hangers have been subjected to QA
- i. surveillances. However, this QA surveillance program will re-main in force until fuel load (when all pipe hangers have been completely installed) and the complete sample surveilled.51/
Should any adverse trends be detected, appropriate actions will be taken to assure that the hangers are brought into conformance with requirements. Tr. 7043 (Fuller), 7190 (Douglas), 7356-58 (Bemis). 142. In conclusion, it is clear that the inspection of 2 pipe hanger welding at the Harris Plant has suffered from nu-merous deficiencies in the past. Equally obvious, however, is I i 51/ Ultimately, about 19,000 seismic pipe hangers will be in-stalled. Tr. 7039-40 (Hate). l. l y-T-* .--. , + - - . - , ~ _ , , --- ., -._,--,,.-.~._w--,,-...-----c,..--.----w.- .--e,~~ - - , , -vw--'
l [ the fact that Applicants recognized these deficiencies and, on their own initiative, implemented a thorough program to assure the quality of the pipe hangers when installed and in-2 l Epected.52/ Based upon the results of the program to date, there exists reasonable assurance that the pipe hangers are ca-pable of performing their intended function. Thus, contrary to the wording of the contention, there is no basis to find that, upon completion of the enhanced program, there will exist de-factive hanger welds that have been improperly inspected and J spproved. G. Eddleman Contention 65: Concrete Containment Structure [ Reserved for reply, if necessary. See Tr. 7366-69.] i H. Eddleman Contention 116: Fire Protection 143. The version of Eddleman Contention 116 which was lit-igated in this proceeding.can be stated as follows:53/ 52/ As the Staff pointed out, the enhanced program, as implemented, exceeds regulatory requirements and industry stan-dards. Bemis, Hallstrom and Blake at 11, 17. 53/ See Memorandum and Order (Ruling on Various Safety and r Procedural Questions) at 4-5 (July 27, 1984)(adopting the pro- t I posed text in Applicants' Motion to Amend Eddleman Contention 116, dated July 16, 1984). A sixth subissue -- whether the plant fire-fighting capability for simultaneous fires is inade-quate -- was rejected by the Board at the hearing for failure , to state a litigable issue. Tr. 4351-52; 4831-32 (Kelley). i i i
- - _ , - - . - . . ..- ---__ - ~ _
('
.(1) The fire hazard analysis of section 9.5A
-(Appendix) in the FSAR'does not address the availability of control and power to the l safety equipment.
! (2) In establishing fire resistance ratings of fire barriers with respect to fires in L cable trays, Applicants have not established L i that qualification tests represent actual plant conditions or comparable conditions. (3) Another vague statement is that barriers are used "where practical" without defining practical or stating the criteria to decide , where a fire barrier is or is not practical
-(and what type of fire barrier is or is not practical). 9.5.1.1.1.
j (4) The " analysis" of Appendix 9.5A does not demonstrate, as 9.5.1.1.1 claims it will, the , adequacy of other fire protection measures in all cases. Rather,'it estimates the BTU of combustible material, smoke generation and removal rate from the area, gives usually a , qualitative description of some measures to mitigate or reduce fire effects, and assumes that the fire will be promptly detected (usu-1 ally, no analysis of location of detection
- instruments, etc.) and the fire brigade will respond rapidly and put out the fire, or the automatic equipment will work. These asser-1 tions are made despite the time it takes to get people into the containment and to the i fire (not well analyzed). Further, the
" analysis" of what happens if the fire i spreads is generally a rationalization that i it can't spread much, not an analysis. See, ,
- e.g. " Analysis of Effects of postulated !
l fires". 3 (5) The effect of a fire in a Fire Area or Fire Zone with a combustible loading greater i than 240,000 BTU /sq. ft. doesn't get dealt
- j. with in realistic terms. 1 144. The purpose of the fire protection program is to en-
, sure the capability safely to shutdown the reactor, maintain it l i
l in a safe shutdown condition, and to limit the radioactive re-i lease to the environment in the event of fire. The SHNPP fire
- protection program consists of design features, personnel, l
l cquipment, and procedures to provide " defense-in-depth" protec-tion of the public health and safety. The program is imple-mented through plant system and facility design, fire preven-tion, fire detection, annunciation, confinement, fire suppression, administrative controls, fire brigade organiza-tion, inspection and maintenance, training, quality assurance, and testing. Applicants' Ex. 6 at 9.5.1-1. The applicable NRC regulations and regulatory guidance for the SENPP fire protec-tion program are: 10 C.F.R. Part 50, Appendix A, General Design Criterion 3 " Fire Protection"; l'1 C.F.R. 5 50.48, " Fire Protec-tion"; 10 C.F.R. Part 50, Appen/dx R, " Fire Protection Program For Nuclear Power Facilities Operating Prior to January 1, 1979";54/ Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 3; NUREG-0800, " Standard Review Plan," Section 9.5 Fire Protection; and Branch Technical Position (BTP) - Chemical En-gineering Branch (CMEB) 9.5-1, " Guidelines for Fire Protection for Nuclear Power Plants," dated July 1981. Applicants' 54/ Although technically the SHNPP is not governed by the ! requirements of Appendix R, Applicants have committed to a fire protection program which meets Appendix R. Tr. 4598-60 (O'Neill); 4640 (Ferguson).
4 'I y Testimony of Margareta A. Serbanescu in Response to Eddleman l Contention 116 (Fire Protection), ff. Tr. 4256 (hereinafter
~
"Serbanescu"), at.5; see a1so, NRC Staff Testimony of Randall Eberly and Robert L. Ferguson Concerning Eddleman Contention
. 116, ff. Tr. 4626 (hereinafter "Eberly and Ferguson"), at 7-8. 145. Applicants' Fire Protection Program is described in FSAR 5 9.5.1 and Appendix 9.5A and in the Safe Shutdown Analy-sis ("SSA") in Case of Fire for the SHNPP. Applicants' Ex. 6; Applicants' Ex. 7 (Summary of the SSA). Mrs. Margareta A. i Serbanescu, a' Principal Engineer with EBASCO, and Mr. David B. Waters, a Principal Engineer - Operations for CP&L, testified for Applicants in response to Contention 116. Mrs. Serbanescu has 19 years of mechanical engineering. experience, including 11 ! . years of fire protection engineering for both nuclear and fos-i sil power generating stations. She has been involved in the
- design and implementation of the SHNPP fire protection program since 1978. Serbanescu at 1-3. Mr. Waters is responsible for i administration of the Harris Plant fire protection program dur-ing the operational phase. He has considerable working knowl-cdge of nuclear plant fire protection programs through his ex-perience with CP&L's H.B. Robinson and Brunswick Nuclear Plants beginning as early as May, 1976. Applicants' Testimony of David B. Waters in response to Eddleman Contention 116 (Fire js Protection), ff. Tr. 4250 (hereinafter " Waters"), at 2-3.
9 i e I
-- - - - - - - - _.---l- - - - - -. --,--.n . - .,_,_..,e--, , . - - - . - . , , , - , - . . - . . , , . , .
.. - - ._ _ .-. . ~ . _ _ . . - - . _ =- -
146. Mr. Randall Eberly and Mr. Robert L. Ferguson [ s testified on behalf of the Staff. Until recently, Mr. Eberly j l l was a Fire Protection Engineer in the Chemical Engineering Branch, NRR, and was the Staff fire protection reviewer for the l SHNPP. Mr. Ferguson is the Fire Protection Section Leader in the Division of Engineering and is responsible for supervising
- the Staff's review of safety considerations associated with fire protection programs at nuclear power generating stations.
Eberly and Ferguson at 1, 3. The testimony of all four wit-nesses rejected the unsupported allegations found in Eddleman 4 Contention 116. l 1. Control and Power to Safety Equipment j 147. Eddleman Contention 116 first alleges that the Fire ! Hazards Analysis in FSAR Appendix 9.5A "does not address avail-ebility of control and power to safety equipment." It is true that the Fire Hazards Analysis in FSAR Appendix 9.5A does not ;
- directly address availability of control and power cables to safety-related equipment; instead, this is addressed in consid-4 erable detail in FSAR Subsection 9.5.1.2.2, " Fire Protection of Cables and Circuitry," FSAR Section 8.3, "Onsite Power Sys- r tems," and in Applicants' SSA. Serbanescu at 6; Tr. 4570-72 (Serbanescu); 4653-55 (Eberly); see, generally, Applicants' Ex.
- 7. Cross-examination of the witnesses during the proceeding by Mr. Eddleman failed to identify any specific deficiency in the
{ FSAR and SSA analyses regarding the availability of control and i
l . . I power to safety. equipment. This allegation is rejected as sim-i . ply off-base. I
- 2. Qualification Tests for Cable Tray Fire Barriers i
l 148. The second issue raised by Eddleman Contention 116 is
- - . cn allegation that "in establishing fire resistance ratings of fire barriers with respect to fires in cable trays, Applicants
, have not established that qualification tests represent actual plant conditions or comparable conditions." A fire barrier is a component of construction, rated by testing laboratories in hours of resistance to fire, which is used to prevent the spread of fire. Each Fire Area in the SHNPP is enclosed within three-hour fire resistance rated barriers or equivalent. In Eddition, certain cable trays within a Fire Area are protected by three-hour or one-hour fire resistance rated enclosures (en-r velopes), as identified in the SSA at Table 9.5B-3. Where a cable tray penetrates a fire barrier, penetration fire stop i seals -- having a minimum fire resistance rating at least squivalent to the rating required of the fire barrier -- are installed as described in FSAR Subsection 9.5.1.2.2.
i Serbanescu at 8. - l 149. The fire resistance ratings of fire barriers are es-tablished by standard qualification tests.!5/ The i t i
}5/ ASTM E-119, " Standard Test Method for Fire Test of Build-ing Construction and Materials"; NFPA-251, " Standard Methods of
! (Continued Next Page) 1 e.--
l l l l qualification tests for determining the fire resistance rating l of a fire barrier are based on an exposure fire represented by 4 the " standard time-temperature curve." The standard time-temperature curve has been determined empirically to rep-resent a common " worst case" exposure fire. Serbanescu at 10-11; Tr. 4526 (Serbanescu); Tr. 4656-58, 4666-68 (Ferguson). The NRC Staff has determined that standard industry qualifica-tion tests provide conservative conditions which envelope actu-al plant configurations. Eberly and Ferguson at 10-11. Fires in mock-ups of rooms representative of a nuclear power plant I have not approached the temperatures of a fire represented by 1 the standard time-temperature curve. Tr. 4668 (Ferguson). A j fire barrier tested to withstand a fire based on the standard time-temperature curve will resist a fire from the maximum cal-culated combustible loading in any Fire Area in the SHNPP power i
; block. Serbanescu at 11.
For each fire barrier for cable trays that will be used in the SHNPP, a qualification test will be performed on a " generic 1 1 1 1 (Continued) Fire Tests of Building Construction and Materials"; Nuclear Mu-tual Limited (NML), " Property Loss Prevention Standards for Nu-clear Generating Stations," Appendix A-14; Underwriter Laboratories (UL) 263, " Fire Tests of Building Construction and l Materials"; and American Nuclear Insurers Bulletin No. 5 "Stan-i dard Fire Endurance Test Method to Qualify a Protective Enve-lope for Class IE Electrical Circuits." Serbanescu at 8-9, At-tachment B; Eberly and Ferguson at 10. l
- r i
i
. 1 i
l casombly" of that fire barrier by an independent laboratory to catablish its fire rating. Serbanescu at 12; Eberly an'd Ferguson at 11. Installation of fire barriers at the SHNPP l will-be-in accordance.with the testing laboratory recommenda-i tions to ensure that the actual installed fire barrier conforms j to the configuration of the tested assembly. Serbanescu at 12. Thus, Applicants have committed to install only those fire bar-riers which meet qualification tests that envelope the worst i ! case fire event to,which they could be exposed at the SHNPP.
- 3. Fire Barriers "Where Practical" l 150. The third issue raised by Eddleman Contention 116 is 4
that FSAR $ 9.5.1.1.1 contains the " vague statement" that
"[ fire] barriers are used "where practical" without defining
" practical" or stating the criteria to decide where a fire bar-rier is or is not practical (and what type of fire barrier chould be used)." FSAR 5 9.5.1.1.1 only makes a general state-
- msnt regarding the fire barriers; whereas, the use of fire bar-i riers in the SHNPP is described in detail in FSAR 5 9.5.1.2.2 and Appendix 9.5A. Applicants' Ex. 6. The specific fire bar-rier locations and qualifications are contained in FSAR Appen-dix 9.5A and Applicants' SSA. Serbanescu at 13; Eberly and Ferguson at 11; Applicants' Exa. 6 and 7.
151. Fire barriers are used to separate Fire Areas to re-l l duce the possibility of fire-related damage to redundant I safety-related trains of equipment and to isolate
-100-I l ,, _ ..__.-_ -_ -._._.._,,_._- , _.. _ __~.. .~ - _ . . . _ - . _ _ _ - . _ - _ - , ,
l l cafety-related systems from hazards in nonsafety-related areas. l
. Fire Areas in the power block are bounded by barriers with-con-l 3 otruction that provide a minimum three-hour fire rating or r
L cquivalent, regardless of the combustible loading. One excep-n tion is the Emergency Diesel denerator Rooms which have large cir. intake openings required for diesel operation. Serbanescu
~
- ct 13. The air exhaust and intakes at exterior wall stacks and i roofs of the outside structure of the power block and remote buildings (i.e., Diesel Generator-Building and Emergency Ser-vice Water-Intake Structure) do not have fire dampers because these walls and roofs are not contiguous with other Fire Areas.
Certain special doors (i.e., tornado, wind and missile doors), bullet resistant doors and air tight doors have not been fire tested, but provide equivalent protection. Applicants' Supple-mental Testimony of Margareta A. Serbanescu in Response to l Eddleman Contention 116 (Fire Protection), ff. Tr. 4256 (here-inafter "Serbanescu II"), at 7. 152. Applicants use the guidance of the NRC's Standard Re-view Plan 9.5-1 55 C.5 and C.7 to determine where fire barriers chould be located. The Staff accepts alternatives to or devia-
- tions from locating fire barriers in accordance with the Stan-dard Review Plan as long an equivalent level of protection is l
provided. Eberly and Ferguson at 12. The fact that "practi-i cal" was not specifically defined in the FSAR is of no conse-
- quence to the NRC reviewer in determining whether the location i
-101-l l
l' 1 l
i l l I of fire barriers in the SHNPP is proper. Tr. 4670-72 (Kelley, Eberly); see also, Eberly and Ferguson at 11. ) l . 153. There was considerable discussion during the hearings regarding the use of doors that have not been fire tested. See, e.g., Tr. 4413-42 (Serbanescu); 4713, 4783-85 (Eberly, Ferguson); 4785-806 (Board, Parties, Eberly). The Staff noted , that the qualification of fire doors was an "open item" with respect to the Staff's review of the adequacy of the SENPP fire
- protection program.5p/ Eberly and Ferguson at 21. Both Staff i
- and Applicants argued, however, that the qualification of fire doors was not relevant to Eddleman Contention 116. Staff coun-
~
i l cel noted that the open item was mentioned in the Staff testi-4 mony as part of its responsibility to keep the Board and the parties informed, not because it was relevant to the conten-
! tion. Tr. 4788 (Moore). Applicants stated that they under-I stood this aspect of Eddleman Contention 116 to question the t
i l Sp/ Subsequent to the presentation of the Staff's case on l Eddleman Contention 116, but prior to the close of the hearing, i the Staff completed its review of the fire doors and found i " Applicants' specialty fire doors are an acceptable deviation from Section C.S.A of the NRC Staff guidelines . . . ." The Staff no longer considers the fire doors an "open item." Joint
- Affidavit of Randall Eberly and Dennis J. Kubicki concerning SER Open Item 8 (Acceptability of Fire Doors), dated November 9, 1984. See Tr. 6908-13. The Board admitted the Joint Affidavit into evidence as Staff Ex. 8 after providing Mr. Eddleman an opportunity to examine Mr. Kubicki. Conference l Call, December 17, 1984. The Board also admitted into evidence I
during the December 17, 1984 Conference Call Applicants' Nsvember 8, 1984 submittal providing additional information to the Staff regarding fire doors. Eddleman Ex. 61. I
-102-l .___________ ___._ _. _____.___._ _.. _ ~ _ , _ . . _ _ _ _
l j i l decision-making process by which it was determined where fire barriers will be installed, not to challenge the qualification 1 of fire doors or any other barrier.}7/ Tr. 4791-92 (O'Neill). l Nevertheless, all special doors are designed and constructed to i provide protection equivalent to a three-hour rating. The man-ufacturer is required to provide a letter of certification to that effect. Tr. 4437-39 (Serbanescu, Waters). Because of the size and weight of the special doors, they cannot be tested in standard test facilities. Tr. 4811-16 (Eberly). However, the , .cdditional strength in construction required of the special l doors generall.y assures that the doors will provide at least an { squivalent degree of protection as rated fire doors. Tr. 4417 (Serbanescu). These doors are constructed of heavy weight, re-
- j. inforced steel plates many times thicker than those used in tested fire doors. Furthermore, the locations within the SHNPP
! where the twenty special doors are installed do not contain !, significant amounts of combustibles that would cause a fire ex-j posure capable of causing failure of the doors.lg/ Staff Ex. 8
}7/ Mr. Eddleman first argued that the qualification of fire doors was relevant to qualification testing of fire barriers i for cable trays. Tr. 4789 (Eddleman). He also argued that it went to the aspect of the contention that alleges that Appli-cants have not analyzed the effect of a fire spreading. Tr.
4798 (Eddleman). The Board noted that the text of the conten-tion certainly did not specifically mention fire doors. Id. (Kelley). The Board did not at the hearing rule on the rele-vance of the fire doors to Contention 116. i
}}/ The actual combustible loadings on either side of the twenty special doors are provided in Eddleman Ex. 61. The l (Continued Next Page)
-103- l I
i
t at 2-3. Finally, the majority of the special doors open to the exterior, where a fire barrier is less important.59/ Id.; Tr. 4418'(Serbanescu). Thus, the evidence in the record demon-ctrates that the special doors will provide adequate fire pro-tection. 154. Applicants have demonstrated that Fire Areas are bounded by rated fire barriers or the equivalent, and the ques-tioning of the use of the term "where practical" by Mr. Eddleman is not of any consequence. 1 .
- 4. Fire Hazards Analysis 155. The fourth issue raised by Eddleman Contention 116 is t.
a generalized criticism of Appendix 9.5A of the FSAR, claiming the Applicants have not demonstrated "the adequacy of fire pro-tection measures in all cases." Contention 116 finds fault with the " estimates" of the BTU content of combustible materi-al, smoke generation and removal rates, measures to reduce or mitigate fire effects, detection capability and fire brigade j response and effectiveness. In response to this aspect of l i l (Continued) Staff also took into account fire detection and suppression systems in the vicinity of the special doors. Tr. (December 17, 1984 Conference Call). 59/ There is no safety-related equipment on the outside of an j exterior door that is threatened by a fire from the inside. Nor is there a combustible load that could create a fire hazard from the exterior to threaten safety-related equipment inside the door. See Eddleman Ex. 61. 1
- 104-
~
_i___ ____ __._ __ __ _____ ___ .._.. m__-.-m_m .. _ ,_. .__ _ . ,__ _ . _ . . _ _ . , _ . _ . _ , , _ , - , . _ . , . _ , _
l Contention 116, Applicants described in considerable detail their Fire Hazards Analysis and measures to prevent, detect, cnd suppress fires and mitigate fire effects. Serbanescu at l 14-28. The Staff witnesses described the applicable guidelines in Standard Review Plan 9.5-1, against which Applicants' Fire Hazards Analysis and fire protection program is measured. Eberly and Ferguson at 12-19. While the Staff will make a site visit at'a very late stage of construction at the SHNPP -- when the majority of the fire protection systems have been installed , -- prior to completing its technical review, the Staff has not identified to date any deficiencies with Applicants' fire pro-tection program that. relate to the issues raised in Contention 116. Eberly and Ferguson at 19-21. 156. To facilitate the Fire Hazards Analysis, Fire Areas were established in the SHNPP based on the nature and occupancies of the plant space, the amount and distribution of combustible materials within the area, and the location of safety-related systems and equipment. Each Fire Area is bound-ed by barriers with construction that provide a minimum three-hour fire resistance rating or an equivalent (with the excep-
-tion of certain air exhausts and intakes described previously).
( For each designated Fire Area, the Fire Hazards Analysi evalu-ates separately the occupancy, boundaries, combustible loading, control of hazards, fire detection, access and initial re-l sponse, fire suppression systems, Fire Area fire fighting i 105-4
, - - , - .ys , . _ - . ,_ - . _ . . _ _ . , _ - - --
l l cquipment, and the effects of postulated fires. Serbanescu at 16; Applicants' Ex. 6. Mr. Eddleman's questioning of the wit- ) l nesses during the hearing on the various aspects of the Fire l Hazards Analysis did not point to any specific inadequacy. Consequently, we need not do more than briefly summarize the considerable evidence which supports our findings. 157. As part of the Fire Hazards Analysis, a detailed in-ventory of combustibles in each fire area is catalogued, including an inventory of " transient" combustibles which might realistically be introduced into areas. The calculated combus-- tible loading of a Fire Area is used to compare the fire hazard relative to those of.other Fire Areas, to judge the' adequacy of I . the area boundary fire barriers, and to verify the proper se-i lection of adequate fire control and suppression systems and equipment. A number of conservatisms are built into this ana-lytical process. A fire barrier hourly rating is then selected to provide a rated boundary to the Fire Area in excess of the l calculated combustible loading. As discussed infra the calcu-lated combustible loading in all Fire Areas -- with the excep-l tion of the four enclosures for diesel oil tanks -- is less than 240,000 BTU /sq. ft. or less than a three-hour fire rating. Sarbanescu at 16-20; Serbanescu II at 2-5; Applicants' Ex. 6. 158. Three different types of fire detectors will be used at the SHNPP based on the type of fire expected'from the type
' ~
of combustibles present in each area: ionization detectors,
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~
l l
/)
v th,ermal detectors and ultraviolet flame detectors. Ionization Ttype smoke detectors were selected as the principal detection cystem becaus,e these detectors respond to the first traces of fire in the form of visible smoke or invisible products of com-bustion. Serbanescu at 22-23. The detection systems selected s l' . are designed to conform to NatConal Fire Protection Association
, +
("NFPA") Codes. Eberly and Ferguson at 16.
,1 159. Automatic fire suppression systems, also designed to i ? conform to NFPA codes, have been provided at the SHNPP. Id.
The specific types of sprinkler systems were described in de-tail by Mrs. Serbanescu. ' Serbanescu at 24-27. The selection
\
of the particular fire suppression system, mode of operation and performance criteria is based'on the fire hazards found in the area, th'e realistic fire expected and the overall fire con-trol approach utilized for containment of the fire. Serbanescu o at 27; Appl'icants' Ex. 6. 160. A trained fire brigade will be available on each chift at f, the SHNPP to respond to any fire event. A minimum of five persons on each shift, who will.have been trained pursuant to the requirements described in F8.4 dection 13.2, plus at least one fire protection tecb34 tt .de to provide expert ad-vice and assistance, will constitute the fire brigade.60/ d, ' J4 60/ Additihnal plant personnel are 'available to fight a fire ' supplementing the designated fire brigade. Tr. 4639-02 (Wa-ters). y N
/'
.j s -107-
,I
- i
Waters at 9-10, Attachment B. A fire brigade response time of [ cpproximately;five to fifteen minutes is expected for most fire i Svents within-the power block. Id. at 5. In reviewing the ad-l cquacy of Applicants' fire protection program, the Staff as-1cumes conservatively that the fire brigade does not respond for thirty minutes. Eberly and Ferguson at 19. Applicants' wit-ness Mr. Waters described in detail fire brigade training, squipment available for fighting fires, fire fighting drills cnd actual practice sessions at a fire-fighting training facility, the availability of a fire engine housed onsite, and arrangements with off-site fire companies to. assist in re-sponding to fires if considered necessary. Waters at 6-11. Importantly, Applicants have established administrative con-trols for flammable liquids and combustible materials'and an cggressive housekeeping program at the Harris Plant to ensure that there is a low probability that a fire which could affect plant safety will occur. Id. at 9. Mr. Waters' testimony con-4 vinced the Board that CP&L's management has fully supported and encouraged the development of an aggressive fire protection program and properly crained fire protection staff at the Harris Plant. 161. Mrs. Serbanescu described a recent change in the
.cmoke removal philosophy for the SHNPP fire protection program.
The supply and exhaust ventilation systems are being provided with fire dampers in ducts which pass through three-hour or l-l
-108-
l two-hour fire rated barriers to maintain the integrity of the l fire barriers which enclose Fire Areas. These ducts, through which the normal ventilation systems are capable of automat-ically removing smoke generated by fire, will now be subject to l l fire damper closure when the fusible link of the fire damper is subjected to a pre-determined temperature. As individual damp-ors close,.the initial smoke removal capability diminishes. In addition, air duct smoke detectors automatically stop certain ! fans in the ventilation system. Serbanescu II at 5-6. How-ever, these ventilation systems can be restored to a smoke re-moval mode by manual actuation from the plant control room. In addition, the automatic shutdown features can be overridden by the plant operator. The fire brigade has at its disposal por-tchle smoke ejection equipment as well as self-contained breathing apparatus for negating the adverse effects of smoke on personnel responding to a fire condition. This change re-flects the fire fighting principle of " bottling up" an area and removing the continuing source of available oxygen to sustain the fire.61/ However, it allows the fire brigade to make a de-termination that smoke removal is necessary in order manually to fight the fire. Id. at 6. l 61/ Mr. Eberly testified that this philosophy is similar to that adopted by other operating nuclear plants which have been approved by the NRC _* rom a fire protection standpoint. The Standard Review Plan does not contain specific criteria regard-ing the adequacy of smoke removal. The Staff relies on indus-try guides. Tr. 4677-83 (Eberly). i 109-
- a't- - - y- m, , *- c-r- <-----Tr--r-- 4 T%' -'7 -" P fs -w---e7*--T * - *e9 9 r--S-7------'W'M 8-mM7'8+- - r--- m--
162. In summary, Applicants' Fire Hazards Analysis verifies the effectiveness of the fire protection program by ovaluation of fire hazards, postulation of realistic potential l l fires, assessment of plant response to a fire and the effects I' of fires in Fire Areas throughout the plant.
- 5. Postulated Fire .in a Diesel Oil Tank Enclosure i
j 163. The fifth issue raised by Eddleman Contention 116 is en allegation that "the effect of a fire in a fire area or fire zone with a combustible loading greater than 240,000 BTU /sq.ft. , doesn't get dealt with in realistic terms." There are four Fire Areas at the SHNPP with a combustible loading greater than 240,000 BTU /sq. ft.: two diesel generator fuel oil day tank en-closures (Fire Areas 1-D-DTA and 1-D-DTB), each having a capac-ity of 3,000 gallons of diesel oil; and diesel fuel oil storage tanks A and B (Fire Areas 12-D-TA and 12-D-TB), each having a total capacity of 175,000 gallons of diesel oil. Serbanescu at 28; Eberly and Ferguson at 20-21.
- 164. The diesel fuel oil day tank enclosures are each iso-lated from other Fire Areas by three-hour rated, reinforced concrete fire barriers. Although the calculated combustible i loading of the enclosures are greater than 240,000 BTU /sq. ft.,
this calculated loading is extremely conservative since it is l based on combustion of the total volume of oil in each enclo-l l cure. The only realistic way to postulate combustion of any significant volume of oil in the fuel oil day tank is attendant
-110-i l
l
l to a rupture of the tank. The diesel fuel oil day tank is a cafety class 3, seismic category I component which is designed ! to. remain functional after a Safe Shutdown Earthquake. Serbanescu at 28-29; Tr. 4737-38 (Eberly). ' l 165. NRC regulatory guidance in Standard Review Plan 9.5-1
$C.1.b provides.that " worst case" fires need not be postulated to be simultaneous with nonfire-related failures in safety sys-tems, plant accidents, or the most severe natural phenomena.
Even in the. highly unlikely event of a rupture of the diesel fuel oil day tank followed by combustion, only a thin layer of oil would actually be ignited in a fire. Furthermore in the cvent of fire, an' automatic multi-cycle sprinkler system would be actuated by thermal detectors to cool the oil below the ig-nition point. If the thermal detectors or the valve automatic release failed to operate, the sprinkler system could be actu-l ated manually. Finally, automatic fusible link fire dampers are provided to the diesel fuel oil day tank enclosures to contain the fire and limit the amount of air available to sup-port continued combustion. All of these design features in combination provide assurance that in the highly unlikely event of a fire in a diesel fuel oil day tank enclosure, the fire l l will be quickly contained, suppressed and extinguished. Sarbanescu at 29; Eberly and Ferguson at 20-21. 166. Diesel fuel oil storage tanks A and B are installed underground in the yard area of the SHNPP, over 175 feet from
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I i principal plant structures. The tanks are constructed of rein-forced concrete designed to' Seismic Category I requirements and
. ore lined with steel. The only access to the tanks is by a re-
! inforced concrete hatch. Each tank vent is provided with a l
- j. flame arrestor to prevent flash-back of a flame into the tank.
Yard hydrants are located adjacent to these tanks to facilitate fighting a fire.- For the reasons discussed above with respect to the diesel fuel oil day tanks, a fire in the diesel fuel oil storage tanks is extremely remote. However, in the unlikely event of a fire, the physical location of the tanks away from plant structures preclude any potential impact to safety relat-l ad eystems. Serbanescu at 30; Eberly and Ferguson at 21. 167. 'Ehe redundant measures designed to protect the SENPP f in the unlikely event of a fire in either a diesel fuel oil day tank enclosure or a diesel fuel oil storage tank are adequate to ensure protection of plant safety systems and safety-related j systems. j 6. Conclusion 168. Both Applicants and the Staff described the SHNPP fire protection program as one based on the " defense-in-depth" concept. The fire protection program is designed to (1) pre-l l vant fires; (2) promptly detect any fire or incipient fire con-1 dition; (3) suppress fires to limit consequent damage; (4) con-fine fires to their areas of initiation; and (5) separate , rcdundant safety-related equipment to maintain operational
-112-
capability under postulated fire conditions. Serbanescu at 31; Eberly and Ferguson at 17-18. No one echelon can or need be
. perfect or complete by itself; each aspect of the fire protec-
, 1 l tion program must meet certain minimum requirements. Strengths l in one part of the program, however, can compensate for weak-nesses, known or unknown, in the others. Eberly and Ferguson at 17. Eddleman Contention 116 was a scattershot attack on certain aspects of Applicants' fire protection program without any articulated thesis of overall inadequacy that somehow re-lates to public health and safety. Applicants have demon-strated that each of the pieces subject to assertions of inade-quacy in the Contention do indeed meet at least the minimum rsquirements contained in NRC regulations and the Staff's guid-ence. More importantly, Applicants have demonstrated that the pieces fit together and that they have developed a coherent fire protection program that provides defense-in-depth to pro-tact public health and safety. I. Eddleman Contention 132C(II): Control Room Design [ Board to insert decision granting summary disposition. See Tr. 2167; Applicants' Motion for Summary Disposition of Eddleman Contention 132C(II) (May 9, 1984); NRC Staff Response in Support of Applicants' Motion for Summary Disposition of Eddleman's Contention 132C(II) (May 29, 1984); Eddleman's Re-aponse to Summary Disposition on Contention 132(c)(2) (June 5, 1984).]
-113-
L . III.- CONCLUSIONS OF LAW f 169. This is a contested proceeding on an application for-I cn operating license for a utilization facility. In issuing l l
- this decision, the Board will have made findings of fact and conclusions of law on all safety matters put into controversy by the parties to the proceeding, except for the emergency pre-paredness contentions which remain pending before the Board.
{ The Board has not determined-that a serious safety, environ-mantal, or common defense and security matter exists. See 10 C.F.R. 5 2.760a. Hother findings required to be made prior to the issuance of an operating license, except for the remaining ! matters in controversy, are to be made by the Director of Nu-clear Reactor Regulation. See id. and 10 C.F.R. $ 50.57. 170. In reaching this decision, the Board should have con-cidered all the evidence submitted by the parties and the en-tire record of this proceeding, consisting of the Commission's ] Notice of Hearing, the pleadings filed by the parties, the transcripts of the hearing and the exhibits received into evi-l dance. All issues and proposed findings presented by the par-ties, and not addressed in the Board's decision, are deemed to ba without merit or unnecessary to the decision. The Board's findings of fact are supported by reliable, probative and sub-stantial evidence in the record.
-114-I'
p-l l l 171. If the Board, in its two partial initial decisions, decides all matters in controversy in favor of authorizing cperation of the facility, it should conclude that, as to the matters resolved in those decisions, the Director of Nuclear
- Reactor Regulation would be authorized, upon making the requi-cite findings with respect to matters not resolved in those de-cisions, and subject.to the Board's resolution of outstanding matters in controversy, to issue to CP&L a license to operate.
the Shearon Harris Nuclear Power Plant. Such authorization by the Board would not be deemed granted, however, until the Board resolves the outstanding matters in controversy or issues a
; further order to the contrary.
, IV. ORDER 172. WHEREFORE, THE BOARD SHOULD ORDER, in accordance with 4 10 C.F.R. 59 2.760(a) and 2.762, that its Partial Initial Deci-l sion on Safety Matters shall constitute the final action of the 4 Commission thirty (30) days after the date of its issuance, un-less an appeal is taken in accordance with section 2.762 or the
! Commission directs that the record be certified to it for final decision. Any Notice of Appeal from the decision must be filed within ten (10) days after service of the decision. A brief in cupport of the appeal must be filed within thirty (30) days j (forty (40) days in the case of the NRC Staff) after filing the Notice of Appeal. Any party which is not an appellant may file l
I
-115-I i
l
._ __ . _ . - , _ . . . . ~ . . - _ _ _ _ . . _ . _ _ . _ - - . . ~ . . . -. . . - - _ - . _ _ _ __ _._ _ _..
c brief in support of or in opposition to the appeal within i thirty (30) days (forty (40) days in the case of the NRC Staff) after the period has expired for the filing and service of the briefs of all appellants. j Respectfully submitted,
=:== < :
Thomas A. B a x t e r , P .'C . John H. O'Neill, Jr., P.C. Pamela H. Anderson Michael A. Swiger
, .SHAW, PITTMAN, POTTS & TROWBRIDGE 1800 M Street, N.W.
Washington, D.C. 20036 4 (202) 822-1000 Richard E. Jones Samantha Francis Flynn CAROLINA POWER & LIGHT COMPANY ! P.O. Box 1551 Raleigh, North Carolina 27602 (919) 836-7707 Edgar M. Roach, Jr. HUNTON & WILLIAMS P.O. Box 109 Raleigh, North Carolina 27602
- (919) 828-9371 l Counsel for Applicants Dated
- December 21, 1984 l
l l ! -116-
4 Brordlove, Russell L. Resume,' Inspector-Level III (QC Inspector II-ANSI) 6227 Browne, Stephen A.
" Applicants'. Testimony of Stephen A. Browne in Response to Joint Contention IV l (Thermoluminescent Dosimeters)" 6407
- Bucci, Richard M.
l " Applicants'. Testimony of Richard M. Bucci and Edwin J. Pagan in Response to Eddleman Contention 9D (Instrument Cable)" 5166
" Applicants' Testimony of Richard M. Bucci, Edwin J. Pagan and Edward M. McLean in Response to Eddleman Contention 9E (Physical i Orientation of Equipment)" 5234
" Applicants' Testimony of Richard M. Bucci,
; Edwin J. Pagan and Peter M. Yandow in Response
] to Eddleman Contention 9F (Lubricants and Seals)" 5441
" Applicants' Testimony of Robert W. Prunty,
- Richard M. Bucci, Edwin J. Pagan and Kuoar V.
- Hate in Response to Eddleman Contentioa 9G (Type Test Reporting)" 5515
" Applicants' Supplemental Testimony of Robert W.
1 Prunty, Richard M. Bucci, Edwin J. Pagan and Kumar V. Hate in Response to Eddleman Contention 9G (Type Test Reporting)" 5515 Conrad, Herbert F.
"NRC Staff Testimony of Ledyard B. Marsh and Herbert F. Conrad Regarding Joint
, Contention VII Part (4)" 4176 Cucimano, John P. l "NRC Staff Testimony of John P. Cusimano and Seymour Block Concerning Joint Contention IV" 6560 Dakin, Thomas W.
" Applicants' Testimony of Richard B. Miller and Thomas W. Dakin in Response to Eddleman Contention 9C (Thermal Aging of RTDs)" 4839 A-2 i
,-.,,,-.n,,_.,----,----,-,---n-----, ---w, ,,----,--ern,n ,-,,,w-ew,, ,----,--,,,-,,<n,w, ,rr,, , - . -.--,
----r- - ,,n.--.-.- .
D'vis,. James M., Jr.
" Applicants' Joint Testimony of James M.
Davis, Jr. and A. Wayne Powell on Joint Intervenors' Contention I (Management Capability)" 3399
~Dintz, C.R.
" Applicants' Joint Testimony of Patrick W.
Howe-and C.R. Dietz on Joint Intervenors' Contention I (Management Capability)" 3124-IDeuglas, Kenneth A. Resume, QA/QC Specialist 7089 Ebtrly,~Randall "NRC Staff Testimony of Randall Eberly and Robert L. Ferguson Concerning Eddleman Contention 116" 4626
- Elleman, Thomas S.
I " Applicants' Joint Testimony of E.E. Utley, M.A. McDuffie, Dr. Thomas S. Elleman j and Harold R. Banks on Joint Intervenors' Contention I (Management Capability)" 2452 Forquson,. Robert L.
"NRC Staff Testimony of Randall Eberly and Robert L. Ferguson Concerning Eddleman Contention 116" 4626 French, Charles S.
Resume, Senior Engineer 6227 Fuller, Alexander G. i
" Applicants' Testimony of James F. Nevill,
, Alexander G. Fuller, David R. Timberlake l' and Kumar V. Hate in Response to Eddleman ! Contention 41 (Pipe Hanger Welding)" 6663 4 Garner, Larry F.
" Applicants' Testimony of George A. Kanakaris, Roland M. Parsons and Larry F. Garner in Response to Eddleman Contention 65 (Concrete Containment Structure)" 5764 I
Hallstrom, George A.
"NRC Staff Testimony of Paul R. Bemis, George A. Hallstrom and Jerome J. Blake on Eddleman Contention Number 41, Pipe Hanger Welds" 7217 A-3
Herris, John R.
"NRC Staff Testimony of John R. Harris, Joseph J. Lenahan and Paul R. Bemis on Eddleman Contention Number 65, Concrete 6320 Placement" Hnto', Kumar V.
" Applicants' Testimony of Robert W. Prunty, Richard M. Bucci, Edwin J. Pagan and Kumar V.
Hate in Response to Eddleman Contention 9G 5515 (Type Test Reporting)"
" Applicants' Supplemental Testimony of Robert W. Prunty, Richard M. Bucci, Edwin J.
Pagan and Kumar V. Hate in Response to Eddleman 5515 Contention 9G (Type Test Reporting)"
" Applicants' Testimony of James F. Nevill, Alexander G. Fuller, David R. Timberlake and Kumar V. Hate in Response to Eddleman 6663 Contention 41 (Pipe Hanger Welding)"
Hitchler, Michael J.
" Applicants' Testimony of Michael J.
Hitchler in Response to Joint Intervenors Contention VII(4)(Steam Generator Tube 4012 Rupture Analysis)" Howe, Patrick W.
" Applicants' Joint Testimony of Patrick W.
Howe and C.R. Dietz on Joint Intervenors' 3124 contention I (Management Capability)" Kanakaris, George A.
" Applicants' Testimony of George A. Kanakaris, Roland M. Parsons and Larry F. Garner in Response to Eddleman Contention 65 (Concrete 5764 Containment Structure)"
Lenahan, Joseph J.
"NRC Staff Testimony of John R. Harris, Joseph J. Lenahan and Paul R. Bemis on Eddleman Contention Number 65, Concrete 6320 Placement" Marsh, Ledyard B.
"NRC Staff Testimony of Ledyard B. Marsh and Herbert F. Conrad Regarding Joint 4176 Contention VII Part (4)"
A-4
M^^cientonio, Armando "NRC Staff Testimony of Armando Masciantonio cn Eddleman Contention 9" 5567 'McDuffie, M.A.
" Applicants' Joint Testimony of E. E. Utley, M.A. McDuffie, Dr. Thomas S. Elleman and Harold R. Banks on Joint Intervenors' Contention I (Management Capability)" 2452 McLnan, Edward M.
" Applicants' Testimony of Richard M. Bucci, Edwin J. Pagan and Edward M. McLean in Response to Eddleman Contention 9E (Physical Orientation of Equipment)" 5234 Millor, Richard B.
" Applicants' Testimony of Richard B. Miller cnd Thomas W. Dakin in Response to Eddleman Contention 9C (Thermal Aging of RTDs)" 4839
" Applicants' Testimony of Robert W. Prunty, Peter M. Yandow and Richard B. Miller in Response to Eddleman Contention 9A (ITT-Barton Transmitters)" 5093 MarqRn, Richard E.
" Applicants' Joint Testimony of Guy P. Beatty Jr. and Richard E. Morgan on Joint Intervenors' Contention I (Management Capability)" 3120 Nnvill, James F.
" Applicants' Testimony of James F. Nevill, !
Alexander G. Fuller, David R. Timberlake cnd Kumar V. Bate in Response to Eddleman Contention 41 (Pipe Hanger Welding)" 6663 Prgen, Edwin J.
" Applicants' Testimony of Richard M. Bucci cnd Edwin J. Pagan in Response to Eddleman Contention 9D (Instrument Cable)" 5166
' Applicants' Testimony of Richard M. Bucci, Edwin J. Pagan and Edward M. McLean in Response to Eddleman Contention 9E (Physical Orientation of Equipment)" 5234
" Applicants' Testimony of Richard M. Bucci, Edwin J. Pagan and Peter M. Yandow in Response to Eddleman Contention 9F (Lubricants and Seals)" 5441 A-5 I
" Applicants' Testimony of Robert W. Prunty, Richard M. Bucci, Edwin J. Pagan and Kumar V.
Hate in Response to Eddleman' Contention 9G (Type Test Reporting)" 5515
" Applicants' Supplemental Testimony of Robert W. Prunty, Richard M. Bucci, Edwin J.
Pagan and Kumar V. Hate in Response to Eddleman Contention 9G (Type Test Reporting)" 5515 Pnrnens, Roland M.
" Applicants' Testimony of George A. Kanakaris, Roland M. Parsons and Larry'F. Garner in Response to Eddleman Contention 65 (Concrete Containment structure)" 5764 Parn, William H.
Resume, QA/QC Specialist 7089 Pinto, Phillip A. Professional Qualifications 6561 P6wnll, A. Wayne
" Applicants' Joint Testimony of James M.
Davis, Jr. and A. Wayne Powell on Joint Intervenors' Contention I.(Management Capability)" 3399 Prunty, Robert W.
" Applicants' Testimony of Robert W. Prunty and Peter M. Yandow in Response to Eddleman Contention 9 (Environmental Qualification of Electrical Equipment)" 4971
" Applicants' Testimony of Robert W. Prunty, Peter M. Yandow and Richard B. Miller in Response to Eddleman Contention 9A (ITT-Barton Transmitters)" 5093
" Applicants' Testimony of Robert W. Prunty, Richard M. Bucci, Edwin J. Pagan and Kumar V.
Hate in Response to Eddleman Contention 9G (Type Test Reporting)" 5515
" Applicants' Supplemental Testimony of Robert W. Prunty, Richard M. Bucci, Edwin J.
Pagan and Kumar V. Hate in Response to Eddleman Contention 9G (Type Test Reporting)" 5515 Sonley, Mac D. Resume, Senior Construction Specialist 6227 A-6
T 89tbanescu, Margareta A.
" Applicants' Supplemental Testimony of
- Margareta A. Serbanescu in Response.to Eddleman Contention 116 (Fire Protection)" 4256
" Applicants' Testimony of Margareta A.
Serbanescu in Response to Eddleman Contention F 116 (Fire Protection)" 4256 Stekes, Charles C.
" Resume - Supplemental Fact Sheet on Experience" 6177
" Resume" 6177 Auburn University Transcript 6177 j " Affidavit" dated June 13, 1984 6177 Strickland, Ricky D.
Resume, Quality Inspector III 6227 Tirb7tlake, David R.
" Applicants' Testimony of James F. Nevill, Alexander G. Fuller, David R. Timberlake and Kumar V. Hate in Response to Eddleman Contention 41 (Pipe Hanger Welding)" 6663 l Tingan, Gene Gibson Resume, QA/QC Control Technician I 7089
! Utlny, E.E.
" Applicants' Joint Testimony of E. E. Utley,
! M.A. McDuffie, Dr. Thomas S. Elleman and ! Harold R. Banks on Joint Intervenors' Contention I (Management Capability)" 2452 i Wntors, David B. l " Applicants' Testimony of David B. Waters in Response to Eddleman Contention 116 (Fire Protection)" 4250 W9tnen, R.A. , " Applicants' Joint Testimony of R.A. Watson and J.L. Willis on Joint Intervenors' Contention I (Management Capability)" 3390 Willis, J.L.
" Applicants' Joint Testimony of R.A. Watson and J.L. Willis on Joint Intervenors' Contention I'(Management Capability)" 3390 A-7
. _ _. ._..._-___._.,_.__-___.._______m
Woltz, C;ndy J. R:sume, QA/QC Technician II 6227 Y^ndow, Peter M.
" Applicants' Testimony of Robert W. Prunty and POter M. Yandow in Response to Eddleman C ntention 9 (Environmental Qualification of Electrical Equipment)" 4971
- Applicants' Testimony of Robert W. Prunty, P0ter M. Yandow and Richard B. Miller in R sponse to Eddleman Contention 9A (ITT-Barton Transmitters)" 5093
' Applicants' Testimony of Richard M. Bucci, Edwin J. Pagan and Peter M. Yandow in Response to Eddleman Contention 9F (Lubricants and Seals)" 5441 i
DOCUMENTS INCORPORATED INTO RECORD Following 'D^ecription Transcript Page LottGr dated December 15, 1982 from Croc p, McCormick and Paget to the North Carolina Utilities Commission 2792 SyctC20 tic Assessment of Licensee Performance datcd August 21, 1984 from NRC Region II to E.E. Utley (Carolina Power & Light Co.) 3660 ApplicCnts' Order of Testimony Presentation -- Oct bar 10, 1984 Hearing 3987 lcTha Poisson Distribution," Chapter 11 of Prrb-bility and Statistics by Allen L. Edwards 4143 Rpplicants' Proposed Schedule l(N3vcCber 14, 1984) 6840 l PEddicran 56," weld symbols drawing 7149 A-8
l APPENDIX B EXHIBITS Exhibit Identified At Admitted At Number Description Transcript Page Transcript Page BoOrd Ex. 1 Ebasco Drawing 4400 CAR-2168 G-115, Revision 5 Board Ex. 2 NUREG/CR-2891, 6501 6527
" Performance Testing of Personnel Dosimetry Services," pages A.45-47 App. Ex. 1 FSAR 55 13.0-13.1.3.2 and 2638 2638 13.4.1-13.5.2.2 App. Ex. 2 Executive Summary from ,
3107 3107 CP&L Second Semi-Annual Report to the North Carolina Utilities Commission, dated June 29, 1984 App. Ex. 3 Management Audit Recommen- 3107 3107 dation Action Plan from CP&L June 1983 Report to the North Carolina Utilities Commission App. Ex. 4 FSAR 55 14.2.2.2, 14.2.3- 3391 3391 14.2.12 App. Ex. 5 FSAR 5 13.2 3398 3399 App. Ex. 6 FSAR 5 9.5.1 and 4245 4273 Appendix 9.5A App. Ex. 7 Safe Shutdown Analysis - 4246 4280 Summary and Description - Fire Protection System B-1
t
' Exhibit Identified At Admitted At Number Description Transcript Page Transcript Pane App. Ex. 8 FSAR $ 3.11 and Appendix 4834 -4972 3.11A -App. Ex. 9 FSAR S 3.8.1 5755 5765 '
. App. Ex. 10 Pour package 1CBXW219001 5756 5765
' App. Ex. 11 Pour Package ICBXW242001 5757 5765 App. Ex. 12 Pour Package ICBXW256004 5757 5765 App. Ex. 13 Pour Package ICBXW276002 5757 5765 App. Ex. 14 Pour Package ICBXW290001 5757 5765 App. Ex. 15 Pour Package ICBXW308001 5757 5765 App. Ex. 16 Pour Package ICBXW336003 5758 5765
! App. Ex. 17 Pour Package 1CBXW386001 5758 5765 App. Ex. 18 Pour Package ICBXW396002 5758 5765 App. Ex. 19 Pour Package ICBXW425001 5758 5765 App. Ex. 20 Pour Package ICBXW444001 5758 5765 App. Ex. 21 Pour Package ICBSL216001 5759 5765 App. Ex. 22 Pour Package ICBXL216002 5759 5765 l App. Ex. 23 Regulatory Guide 1.136 6190 62141/ l App. Ex. 24 Regulatory Guide 1.142 6190 REJECTED, 62141/ , App. Ex. 25 Specifications of TL 6404 6407 5 Badge and TL Badge Hanger i. 1/ Admission of these documents was moved by Mr. Eddleman. App. Ex. 23 was j subsequently identified and admitted as Eddleman Ex. 19. See Tr. 6217. ' l i B-2 -t .
r Exhibit. Identified At Admitted At .
. Number Description Transcript Pane Transcript Pane
~ App. Ex. 262/ CP&L letter dated June 1,
.1982 from H. R. Banks to 7055 7055 J. P. O'Reilly (NRC),
response to Inspection
, Report 82-03, with attachments -
App. Ex. 27 CP&L letter (NRC-291) See Applicants' Motion to dated November 30, 1984 Receive Additional with attached " Final Evidence (Eddleman i Report - Shop Welding Contention 41), Deficiencies in Seismic December 11, 1984 i I Pipe Hangers Supplied by Bergen-Paterson, Item 95; Undersized Skewed Tee i Fillet Welds on Seismic I Pipe Hangers, Item 72" , App". Ex. 28 CP&L letter (NRC-292) See Applicants' Motion to e dated November 30, 1984 Receive Additional l with attached " Final Evidence (Eddleman Report - Pipe Hangers Contention 41), Previously Accepted by December 11, 1984
; QC Welding Inspectors, j Item 96; Undersized Skewed
- Tee Fillet Welds on Seismic i
I Pipe Hangers, Item 72" 1 i j Stcff Ex. 4 SER 2392 2392
! Stcff Ex. 5 SER Supplement 1 2392 2392 2
Stcff Ex. 63/ Standard Review Plan, 4627 4628
! Section 9.5.1 i
i \ l 2/ This exhibit was mistakenly identified as Applicants' Ex. 25. See Appli- , l cants' Proposed Transcript Corrections, December 20, 1984. 3/ This exhibit was initially numbered as Staff Exhibit 7; the identification i eas corrected at Tr. 5026 B-3 i _ . - ~ , . _ . _ ~ . . . - - . , - - . . . . _ - - , - . , - - _ - - . , - - - , . _ - - . _ - _ . - - - _ - _ _ _ _ _ - , _ - - - - - _ . - - - , - , -
,-Eshibit . Identified At Admitted At Number Description Transcript Page Transcript Page Stnff Ex. 7 Affidavit of Armando S. December 17,1984 Conference Call Masciantonio, Richard A. Kendall and Robert C. ' Jones, Jr. in Further Response to Eddleman Contention 9A Staff Ex. 8 Joint Affidavit of Randall December 17, 1984 Conference Call Eberly and Dennis J. Kubick; concerning SER Open Item 8 (Acceptability of Fire Doors) WE Ex. 2 NFPA-31, pages 4386 49004/ 2-5, 24-31 WE Ex. 3 NFPA-30, pages 4-7 4386 4900 b/ WE Ex. 4 NFPA-30, pages 4386 4900 b/ 8, 9, 12-15 WE Ex. 5 NFPA-30, pages 16-19, 4388 4900 b/ 30-35, 38-45 WE Ex. 6 NFPA-30, pages 68-75 4390 4900 b/ WE Ex. 7 NFPA-30, pages 78-79 4390 4900 b/ WE Ex. 8 NFPA-30, pages 88-89, 4391 4900 b/ 106-107 WE Ex. 9 NFPA-30, pages 126-133 4391 4900 b/ WE Ex. 10 Ebasco Specification 5893 5943 CAR-SH-CH-6, " Concrete," (Revision 11) WE Ex. 11 Technical Procedure TP-15 5922
" Concrete Placement Inspection" (Revision 11) 4/ See Tr. 4899-4900: Eddleman Exs. 2 through 9 admitted for limited purpose of demonstrating existence of Code, not to prove any technical issue on the merits.
I B-4 i l _ ___ - a
Exhibit Identified At Admitted At Number Description Transcript Page Transcript Page WE Ex. 12 CQC-13, Concrete Control 5923 WE Ex. 13 QCI-13.1, Concrete 5923 Compressive Strength Testing WE Fx. 14 QCI-13.2, Batch Plant 5924 Inspection WE Ex. 15 QCI-13.5, Sieve Analysi. 5924 of Fine and Coarse Aggregate i WE Ex. 16 Work Procedure WP-04, 5925 Concrete Production and Delivery (Revision 10) WE Ex. 17 Applicants' Supplemental 5925 Response to Wells Eddleman's Request for Production of Documents (Contention 65), May 25, 1984
- WE Ex. 18 Testimony of Charles C. 6165 Stokes, dated October 30, 3 1984 WE Ex. 19 Regulatory Guide 1.136 6217 6217 WE Ex. 20 " Notebook 05, Eddleman 6651 partially
- 41," Seismic Weld Data admitted 5/
Reports 6892 WE Ex. 21 Document List on 6748 Eddleman Contention 41, , August 10, 1984 I WE Ex. 22 CP&L letter dated 3/24/82 6749 6749 from Chiangi to O'Reilly (NRC), Item 72; CP&L letter dated 9/13/82 from Chiangi to O'Reilly (NRC), i Item 96 l l l l 5/ Only those portions of document which were subject of cross-examination ! were admitted. l B-5
s. 4_ l Exhibit Identified At Admitted At Number Description Transcript Page Transcript Page WE Ex. 23 Figure 1.8-la (attached 6754 to 12/12/83 letter from McDuffie to Denton) WE Ex. 24 Figure 1.8-1 (attached to 6754 12/12/83 letter from McDuffie to Denton) WE Ex. 25 .QCI-19.3, Seismic Pipe 6770 6783 Hanger Inspection Documen-tation System WE Ex. 26 WP-110, Installation of 6783 6794 Seismic Pipe Hangers and Supports for Seismically Analyzed Pipe WE Ex. 27 CP&L memo (MQAH-83-321) 6809 6813 dated 10/24/83 from Chiangi to Parsons with attachments . WE Ex. 28 Pipe Hanger Problem reports 6821 WE Ex. 29 FCR-H-286 6898 WE Ex. 30 FCR-H-979, Rev. 3 6903 6907 WE Ex. 31 FCR-H-979, Rev. 3, 6903 6907 Justification WE Ex. 32 WP-139, Pipe Hanger 6918 6918 Work Package Preparation
' WE Ex. 33 Memo dated December 9, 6923 6927 1980 from D. Timberlake to R. Hanford, Welder Craft Training Bimonthly Update WE Ex. 34 Group of training documents, 6928 6940 4
cover page dated 4/15/84 entitled "QA/QC Harris l Plant Personnel Training" B-6
Exhibit Identified At Admitted At Number Description Transcript Page Transcript Page ; WE Ex. 35 _ Group of training documents, 6943 6957 cover page dated 10/17/80, entitled " Daniel Contruction Company, Shearon Harris Nuclear Power Plant, Training Record" WE Ex. 36 Group of training documents, 6945 6957 cover sheet dated June 18-21, 1979, entitled "Shearon Harris Power Plant QC Receiving Training Program" WE Ex. 37 Group of training documents, 6953 6957 cover sheet dated 9/12/80, entitled " Construction QA
, Section Personnel Training" WE Ex. 38 CQA-1, " Personnel 6957 6958 Training and Qualification" WE Ex. 39 NDEP-605,." Visual Examina- 6959 6966 i tion of Seismic I, Structural and Hanger Welds (SHNPP)"
WE Ex. 40 CQC-19, " Weld Control" 6959 6966 WE Ex. 41 CP&L letter (NRC-127) 6968 6970 dated 10/3/83 from Parsons to O'Reilly a WE Ex. 42 FCR-AS-4294 6968 6972 WE Ex. 43 NRC IE Report No. 6973 6975 , 50-400/82-03, cover letter dated April 28, 1982
- WE Ex. 44 MP-08, " General Welding 6978 6992 l Procedure for Structural l
Steel (Seismic, Non-Seismic) and Hangers" WE Ex. 45 MP-10, " Repair of Base 6978 6981 Materials and Weldments" B-7
e Exhibit . Identified At Admitted At Number- Description Transcript Page Transcript Page
~WE Ex. 46 CP&L letter (NRC-126) 6993 7002 dated October 3, 1983 from Parsons to O'Reilly WE Ex. 47 CP&L ltr (NRC-196) 6997 7001 dated March 21, 1984 from Parsons to O'Reilly WE Ex. 48 Corrective Action and 6998 Disposition of Hanger Reinspection, no. HR-21 WE Ex. 49 NRC Inspection Report Nos. 7004 50-400/83-25 and 50-401/
83-25, cover letter dated October 19, 1983 WE Ex. 50 Group of Seismic Weld 7091 partially , Data Reports, first report admitted i dated December 23, 1982, (see n.5, Hanger PD-H-3 supra) 7201-7205 WE Ex. 51 Group of Seismic Weld Data 7092 partially Reports, first report admitted dated May 17, 1982, (see n.5, Hanger PD-H-121 supra) 7201-7205 WE Ex. 52 Seismic Weld Data Reports 7092 for Hangers PM-H-227, PM-H-229, PM-H-231 and PM-H-441 WE Ex. 53 Seismic Wald Data Reports 7092 for Hanger WL-H-2343 i WE Ex. 54 Seismic Weld Data Report 7093 for Hanger BR-H-2182 i WE Ex. 55 Seismic Wald Data Reports 7093 for Hangers WL-H-2369 and WL-H-2355 WE Ex. 56 Weld Symbol Drawing [Not identified formally or admitted as evidence; incorporated into record ff. Tr. 7149] f B-8 l i t
. . Exhibit Identified At- Admitted At Number Description Transcript Page Transcript Page WE Ex. 57 Attachment to CP&L 7223 Letter of Response to ' NRC Report RII: JWY 50-400/83-20 WE Ex. 58 CP&L. letter dated 6/11/81 7223 partially from Chiangi to O'Reilly admitted with attached Final 7251-72526/ Report, Wald Symbol Errors and Misapplication of Weld on Bergen-Paterson Pipe Hangers WE Ex. 59 Group of Seismic Weld - 7224 partially Data Reports, cover sheet admitted entitled "E-41, Weld (see n.5-Inspection Reports supra) Produced in 1984" 7279 WE Ex. 60 Affidavit of Chan Van Vo, 7362 dated October 6, 1984 i WE Ex. 61 CP&L letter dated November December 17, 1984 Conference Call 8, 1984 from S.R. Zimmerman j to H.R. Denton, response to SER Open Item 8 i 4 JI Ex. 1 Joint Contention I from 2455 2456
" Admitted Contentions"
- dated January 21, 1983 JI Ex. 2 Excerpt from FERC Form 2466 2497
, No. 1, Annual Report of . CP&L dated December 31, 1983 JI Ex. 3 ACRS letter 2516 REJECTED, 2623 JI Ex. 4 CP&L letter NRC-111 from 2516 2749 R.M. Parsons to J.P.
O'Reilly (NRC) in response to IE Report 50-400/83-22-02 i 6/ Portions of attached report dealing with HVAC, Cable Tray and Conduit Sup-
- ports (pp. 3, 4 and Exhibits 4 and 5) were not admitted.
, B-9 4
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Exhibit Identified At Admitted At Number Description Transcript Page Transcript Page
'JI Ex. 5 CP&L letter NRC-130 from 2516 REJECTED, 2846 R.M. Parsons to J.P.
O'Reilly (NRC) in response ; to IE Report 50-400/83-26-01 l l JI Ex. 6 NRC IE Report 50-400/ 2516 2889 83-25 j JI Ex. 7 CP&L letter NRC-165 from 2516 REJECTED, 2846 ! R.M. Parsons to J.P. O'Reilly (NRC) in response to IE Report 50-400/83-33 JI Ex. 8 Excerpts from FOIA material 2516 requested by T. Vaden
" Details of Investigation -
CP&L Harris Plant - December 14, 1981 - January 22, 1982" JI Ex. 9 CP&L letter NRC-194 from 2516 2763 R.M. Parsons to J.P. O'Reilly (NRC) in response to IE Report 50-400/84-02 JI Ex. 10 CP&L letter NRC-184 from 2516 2770 R.M. Parsons to J.P. O'Reilly (NRC) in response to IE Report 50-400/83-37-02 i JI Ex. 11 CP&L letter NRC-216 from 2516 2778 R.M. Parsons to J.P. O'Reilly (NRC) in response to IE Report 50-400/84-06 JI Ex. 12 Abnormal Occurrence Report, 2625 2749 47 Fed. Reg. 21653-56, dated May 19 1982 JI Ex. 13 Comparison of Productivity See 3029 Admitted only Rates iHarris Unit 1, NCUC as an offer of Docket E-2, Sub 481, proof. See McDuffie Ex. 1 Tr. 3051-54. B-10
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- . . - - . -~
t Exhibit Identified At Admitted At Number Description Transcript Page Transcript Page JI Ex. 14 CP&L Second Semi-Annual See 2793-94 2811 Report to' North Carolina Utilities Commission re status of implementation of Cresap recommendations, dated June 29, 1984 JI Ex. 15 Excerpt from Applicants' 2895 2894 May 1, 1984 responses to Joint Intervenors' Interro-gatories on Joint Contention I, p. 27 JI Ex. 16 Excerpt from Applicants' 2895 3295 May 1, 1984 responses to Joint Intervenors' Interro-I' gatories on Joint Contention I, Attachment I-24 4 JI Ex. 17 Excerpt from Applicants' 2895 2958 May 1, 1984 responses to Joint Intervenors' Interro-gatories on Joint Contention I, Attachment I-39(r) JI Ex. 18 $600,000 Civil Penalty 2895 2958 materials transmitted by C.A. Barth, March 10, 1983: NRC IE Report 50-324, 325/82-28, February 10, 1983 letter from NRC to CP&L with i ' Notice of Violation and ,. Proposed Imposition of Civil Penalties, NRC Confirmatory Order EA-82-106 JI Ex. 19 Systematic Assessment of 2895 3003 Licensee Performance dated January 15, 1981 from NRC Region II (SALP I) l JI Ex. 20 Systematic Assessment of 2895 3003 Licensee Performance dated May 21, 1982 from NRC Region II (SALP II) i )
. B-11 i
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s Exhibit r Identified At Admitted At Number Description Transcript Page Transcript Page 2 m JIEx.$1 g Systematic Assessment of 2895 3003 j Licensee Performance , j from NRC Region II
^
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dated June 14, 1983 (SALP III) _a JI Ex. 2?. CP&L Quality Check Report, 2895 3006 == form 80251 m W - JI Ex. 23 Investigation ,- CP&L 3349 REJECTED, 3381 - Brunswick Plant, dated a February, 1982 by A.R. '
===
Jacobstein for North Carolina Utilities f - Commission i ' 9 JI Ex, 24 Excerpt from Applicants' 3248 3248 -'_ May 1, 1984 responses to " Joint Intervenors' Interro- i gatories on Joint Contention - I, Attachment I-16 JI En. 25 Excerpt from Applicants' 3248 3248 May 1, 1984 responses to , 4 Joint Intervenors' Interro- _ gatories on Joint Contention I, Attachment I-18 M Z - JI En. 26 " Nuclear Plant Capccity 3248 withdrawn, 3248 Factors 1978-1983"
]
.3iiii JI Ex. 27 " Capacity Factorts of the 3248 3248 --
Brunswick and Robinson E Reactors", as modified d JI EU 28 CP&L Brunswick Plant 3264 2 3264 - Outage Schedule, NCUC Docket E-2, Sub 481,
-3 CP&L Ex. PWH-1 <
j of En. 29 NRC IE Information Notice 3316 E 3 3325 84-59 dated August 6, 1984 ; 1 g
'l -
JI En. 30 NRC letter to S. Harris ' 3615 3622 _- (CP&L) dated September 23, ' 1975 with Notice of Proposed
]
_= Imposition of Civil
, Penalties ($7,000) y B-12 2
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7;, y . ; :' :.,~ -: . JI Ex. 31 NRC letter dated August 1, 3615 3622 3, .. '.e 1980 to J.A. Jones, with % SJ. < l Notice or Violation and - ' .".1.'~- Proposed Impositions of 'y" Civil Penalties ($89,000); -,. i.- NRC IE Report 50-325/80-18 $1J Jf 1, and 50-324/80-15; CP&L 6J *\ -
# ' 9' response to Notice of Violation, dated August 27, 1980 f.j[*.. M '
?: .4 4.- .+.e .: . .> :.
.y.
JI Ex. 32 NRC IE Report 50-325/80-12 3615 3622 {.7..-;ll and 50-324/80-11; NRC J.' ,# ." . letter dated June 11, 1980 f:sc; ' , - to J.A. Jones with Notice M. l i~ of Violation and Proposed N* I Imposition of Civil Penalties ($42,000); CP&L } >J ."_ '..'f",: response to Notice of " [ [ , .' 3 Violation dated July 3, 1980 ^ ' f
~.'.
- . . . ; 2.
JI En. 33 NRC letter dated November 23, 3615 3622 l H j 7' 1981 to E.E. Utley; NRC letter . . ;- " dated October 1,1981 to J. A. ' . 7 d- c Jones with Notice of Violation 1 ; ~ .' .' . . .' ' and Proposed Imposition of .
. f0f _ . . .
Civil Penalties ($40,000); -. -- Y J.l
'.f CP&L letters of response to Mf ".
Notice of Violation dated ~ October 30, 1981 and f. December 23, 1981; NRC IE ^ ~' ' Report 50-325, 324/81-16
- f. ,. ;,.D _$[#.. .
1 JI Ex. 34 NRC letter dated July 16, 3615 3622 ? ' ' . ^'f . . .
'I '
1982 to J.A. Jones, with .c' i } , / f . Notice of Violation and 'c J N j. ' Proposed Imposition of ( .V . U.'4 R Civil Penalties ($120,000); NRC IE Report 50-324, 325/ -9$%12-;. 1 l 82-02; CP&L response to M.6:?-[i I. Notice of Violation dated . ' ? :. . . * . % ~ August 16, 1982 1.: : .M T#
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I Exhibit Identified At Admitted At Number Description Transcript Page Transcript Page JI Ex. 35 NRC letter dated May 12, 3615 3622 1981 to J.A. Jones, with Notice of Violation and Proposed Imposition of Civil Penalties ($40,000); NRC IE Report 50-261/81-10; CP&L responses to Notice of Violation dated June 17, 1981 and June 30, 1981 JI Ex. 36 NRC letter dated December 1, 3615 3622 1981 to J.A. Jones, with Notice of Violation and Proposed Imposition of Civil Penalties ($50,000); CP&L response to Notice of Violation dated January 5, 1982; NRC IE Report 50-261/ 81-24 JI Ex. 37 NRC letter dated March 13, 3615 3622 1984 imposing $30,000 Civil Penalty; NRC IE Report 50-261/84-07; CP&L response to Notice of Violation dated June 15, 1984 JI En. 38 Preliminary Assessment of 3671 3852 the Organization and Management of CP&L for operation of the Harris Plant, SECY-81-617, Enclosure 4 JI Ex. 39 Charts prepared by Joint 3803 3852 Intervenors showing
" selected functional '
area comparison" of SALP ratings for CP&L nuclear plants JI Ex. 40 Excerpts from Critical Mass 3647 3647 Energy Report by J. Clewett a B-14
?
G Exhibit Identified At Admitted At - Number Description Transcript Page Transcript Page " w" JI Ex. 41 NRC letter dated November 15, 3880 3881 1983 to E.E. Utley imposing ! civil penalty ($20,000); - CP&L response to Notice of == Violation, dated December 12, a 1983 JI Ex. 42 NRC IE Report 50-324, 3880 3881 325/83-11; NRC letter dated January 10, 1984 to E.E. Utley, with Notice of - Violation and Proposed ! Imposition of Civil Penalty - ($40,000); CP&L responses to - Notice of Violation dated February 9, 1984 and March 30, E 1984 i JI Ex. 43 NUREG/CR-2891, 6596 6596
" Performance Testing '
of Personnel Dosimetry : Services," pages D.24, - D.25
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