ML20082R745
| ML20082R745 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 12/09/1983 |
| From: | Norton B NORTON, BURKE, BERRY & FRENCH, PACIFIC GAS & ELECTRIC CO. |
| To: | |
| References | |
| NUDOCS 8312130285 | |
| Download: ML20082R745 (100) | |
Text
.
I DOCHETED i
USNRC UNITED STATES OF AMERICA NUCLEAR REGULATORY COMISSION
,0 3 UEC i 2 fi!I :57 '
BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD c.gcp;gc my
...,.m x s. q,. -
In the Matter of
)
Docket Nos. 50-275
-.W.e
)
50-323 PACIFIC GAS AND ELECTRIC COWANY
)
)
Reopened Hearing -
(Diablo Canyon Nuclear Power
)
Design Quality Assurance Plant, Units 1 and 2)
)
)
PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW SUBMITTED BY PACIFIC GAS AND ELECTRIC COMPANY I.
INTRODUCTION By order dated April 21, 1983, the Atomic Safety and Licensing Appeal i
Board reopened the record to hear evidence concerning alleged deficiencies in the implementation of Pacific Gas and Electric Company's (PGandE) design quality assurance program as it applied to Units 1 and 2 at the Diablo Canyon Nuclear Power Plant (DCNPP). Necessarily included was the adequacy of the design verification programl established by PGandE and its contiactors to compensate for the alleged deficiencies. As stated by the Appeal Board in a prehearing conference order dated August 16, 1983:
//
- l. The verification program consisted of two elements: The Independent Design Verification Program (IDVP), which was established in response to Commission Order CLI-81-30 and the NRC Staff Letter dated November 19, 1981,d as a condition for re'nstatement of PGandE's suspended low power license, an the Internal Technical Program (ITP), which was a complementary effort by the Diablo Canyon Project (DCP) organization. The primary objectives of the ITP were to (1) provide an additional design review effort to assure the overall adequacy of the analyses and design of the plant, (2) develop data and findings, and (4)pport of the IDVP, (3) respond to IDVP open items andimp information in su arising from the IDVP and the ITP.
8312130285 831209 $DRADOCK 05000275 PDR
"The real issue in the reopened proceedings has, in effect, moved beyond the question of what deficiences existed in the applicant's Diablo Canyon design quality assurance program to the question whether the applicant can demonstrate that the IDVP and the ITP verify the correctness of the Diablo Canyon design."
(Order, at 6.)
Extensive discovery was pursued by the parties. As directed by the Appeal Board, the Governor and Joint Intervanors (JI) jointly proposed a series of contentions to be litigated in this proceeding. At the conmiencement of the hearing, there were 39 contentions remaining after certain contentions had been denied by the Board or deleted by prior stipulation. Those contentions are set forth in Appendix A attached hereto. Of the 39 remaining 2 of j
contentions, neither JI nor the Governor offered any testimony on 27 3
those contentions and as to 3, the sponsoring witness stated he no longer had any concerns. The Governor and JI did, however, address by cross examination a number of the 27 contentions on which they offered no testimony.
The reopened hearings were held at the San Luis Bay Inn, Avila Beach, 1
California, from October 31, 1983 through November 21, 1983. At the hearing prefiled direct testimony was received in the record and sponsoring witnesses were made available for cross examination. Exhibits were offered by the parties and, as indicated in Appendix B, were received in evidence.
//
//
//
- 2. la, lb, lc, le, 2a, 2b, 2d, 3f(i), 3f(11), 30, 3s, 3t, 4a, 4b, 4h, 41(1),
41(2), 4j(10, 4j(2), 4k, 41, 4q, 4r, 4s, 4t, and 4u.
3 3f(iv), 3f(v), and 3p.
//
,2-i i
~..
s i
PROPOSED FINDINGS OF FACT 4
i Contentions 1 and 2 1
1.
PGandE, the IDVP, and the NRC Staff testified on all subparts of i
Contentions 1 and 2.
The Governor and JI presented testimony only on 4
Contentions 1(c) and 2(c). PGandE presented the testimony of Richard C.
i 4
Anderson, Engineering Manager for the DCP; Gary H. Moore, Project Engineer for Unit 1; Gregory V. Cranston, Project Engineer for Unit 2; Dr. William H.
~
White, Assistant Project Engineer in Charge of Seismic Analysis; Larry E.
i Shipley, Assistant Chief Plant Design Engineer; and Dr. Stanley Kaplan, a consultant specializing in probabilistic and decision theory. The IDVP l
presented testimony of Dr. William E. Cooper, a consulting engineer for 5
l Teledyne Engineering Services (TES), who managed the efforts of TES as Program Manager for the IDVP; Dr. Robert L. Cloud, principal of Robert L. Cloud Associates, Inc. (RLCA), which verified the seismic, structural, and l
mechanical aspects of the design process for Unit 1; John E. Krechting of i
Stone and Webster Engineering Company (SWEC), who, as Project Engineer, t
j managed the technical efforts of SWEC in verifying the safety system and I
safety analysis aspects of the design process; and Roger F. Reedy, principal l
of the firm of R. F. Reedy, Inc. (RFR), which performed the design QA audits and reviews and the design office verification for the IDVP. The Governor 6
presented testimony of Dr. George Apostolakis, a Professor in the School of Engineering and Applied Science at the University of California, Los Angeles, i
who has done research in risk assessment, including probability theory, decision theory and statistics, reliability analysis, and nuclear i
j engineering. The JI presented testimony of Dr. Francisco J. Samaniego, a Professor of Statistics at the University of California, Davis. The NRC Staf f 4.
presented the testimony of James P. Knight, Assistant Director for Components and Structures Engineering, Division of Engineering; Hartmut E. Schierling, Senior Project Manager, Division of Licensing in the Office of Nuclear Reactor Regulation; and Jared S. Wermiel, a Section Leader in the Auxiliary Systems Branch in the Office of Nuclear Reactor Regulation.
2.
The design verification program at DCNPP consisted of the IDVF and the ITP. The program included in-depth review and analyses of plant design to assure compliance with criteria and the implementation of modifications to meet criteria. Anderson et al., ff. Tr. D-224, at 3.
Scope of the Independent Design Verification Program 3.
The IDVP was to be an independent, qualified, technical contractor who would select samples of design activities for Unit 1 and, by review of existing documentation and, where considered appropriate, by independent analyses, draw conclusions as to whether or not the previously approved licensing criteria had been met. Knight et al., ff. Tr. D-2649, at 5.
4.
The IDVP, which provided the independent design review of the adequacy of the design of Unit 1, verified both the technical plant design which was done by PGandE and its service-related contractors or the DCP, and the quality assurance program applied to the design process. Anderson et al., ff. Tr.
D-224, at 7, 8.
5.
The IDVP was conducted under the management of TES. In addition to TES, the principal participants in the IDVP were RLCA, RFR, and SWEC. The IDVP also utilized the following consultants: Hansen, Holly and Biggs, General Dynamics, Alexander Kusko, Inc., Foster-Miller Associates, and J. M. Wheaton.
Anderson et al., ff. Tr. D-224, at 3, 8, 9.
6_.
The goals of the IDVP were to detect all failures to comply with.
I
.., _.. - _ _ _ _ _ _. _ ~. _ -. _ _. _... _ _
licensing criteria in the sampled systems (Krechting Tr. D-1538) as well as to assure that generic errors did not exist in the unreviewed design. Cooper, Tr. D-1537-38. This was accomplished by means of indeptw. Ant reviews and the development of a comprehensive understanding of the engineering process and of the results obtained through that process. Cooper et al., ff. Tr. D-1459, at j
1/2-13-16,1/2-21-25.
7.
The IDVP program efforts, the methodology and procedures applied to the program and the criteria for determining adequacy of design were fully documented by the IDVP Program Management Plans, the Interim Technical Reports (ITRs), and tne IDVP Final Report. Cooper et al., ff. Tr. D-1459, at 1/2-35; Knight et al., ff. Tr. D-2649, at 11.
8_.
Throughout the course of the design verification effort the NRC Staff met with the IDVP and the DCP in open public meetings to discuss the progress of the verification effort and to ensure that the program met the objectives set forth in the Commission Order and NRC Staff Letter. Knight et al., ff. Tr.
D-2649, at 11.
9.
The IDVP program first identified PGandE's service-related design contractors whose efforts affected the final design. Cooper et al., ff. Tr.
D-1459, at 1/2-9,1/2-10.
It then reviewed the quality assurance program of PGandE and those service-related contractors and audited the implementation of their QA programs.
Initial and followup samples of their design activities were verified along with DCP actions, including reanalysis and modifications.
Anderson et al., ff. Tr. D-224, at 8.
- 10. The QA Audits and Reviews were performed by the IDVP to evaluate both the formal QA program imposed for the work and the implementation of that They also provided background information which might have impacted program..
I t
7 l
the extent of design process verification. Based on experience gained in the seismic review, an additional step was added for the non-seismic review. For the non-seismic review if the reviewed organization did not have a formal QA program, or if its formal QA program was not properly implemented, its actual design control practices were evaluated and reported as a part of the QA Audit and Review Report. Additional sampling was considered if negative results were obtained from the QA Audit and Review of an organization whose work was not included in the initial sample. Similarly, additional verification was considered when the organization's work was included in the initial sample, but the initial sample did not include the negative aspect. Cooper et al.,
ff. Tr. D-1459, at 1/2-11-12.
IDVP - Seismic H. The IDVP reviewed the seismic design of the safety-related contractors who performed design work and whose work contributed significantly to t..e seismic design or qualification of the plant. The IDVP also reviewed the internal seismic design performed by PGandE. Anderson et al., ff. Tr. D-224, at 24.
R. The entire seismic design process, including the making of drawings, development of response spectra, and the transmittal of information, was reviewed by the IDVP. Ar.derson et al., ff. Tr. D-224, at 24.
_13,.
In its verification of seismic design, the IDVP performed a complete independent analysis of the initial sample and additional sample / verification.
In its verification of the DCP activities, the IDVP used independent calculations on a selected basis as part of the design verification process. In every aspect of the IDVP's seismic work, the verification process consisted of much more than merely checking data of
inputs to models used by PGandE. Cooper et al., ff. Tr. D-1459, at 1/2-34-35.
H. In its verification of the Unit I seismic design, the IDVP sampled all Class I (safety related) structures and those additional structures containing Class I equipment. The IDVP also sampled various size piping and related supports, as well as HVAC ducts, electrical raceways, instrument tubing and related supports. For major activities of the ITP's 100% seismic design verification, such as reanalysis of a building or of all large-bore piping, the IDVP carefully reviewed the complete design methodology. Anderson et al.,
ff. Tr. D-224, at 9.
IDVP - Non-Seismic
,15. In its verification of the non-seismic design, the IDVP performed independent calculations or analyses, and/or independent review of PGandE calculations and analyses. The majority of the Phase II non-seismic verification consisted of the performance by the IDVP of independent calculations or analyses. The independent calculations and analyses performed by the IDVP used independent models developed by IDVP i.;. /or different computer programs.
In its additional verification of DCP-performed activities, the IDVP used independent calculations, analyses, and/or field verification for essentially all of the verification effort.
In every aspect of the IDVP's non-seismic work, the verification process consisted of much more than merely checking data of inputs to models used by PGandE. Cooper et al., ff. Tr. D-1459, at 1/2-35.
- 3. In its verification of non-seismic design, the IDVP Program Plan provided for selecting a broad sample of various types of engineering design work.
Among the criteria used to select samples were:
(1) diversity (water, air, and electrical), (2) importance to safety. (3) interfaces between PGandE.
t
internal design organizations and between PGandE and external service-related contractors, and (4) interfaces with the design criteria of the NSSS vendor.
Anderson et al., ff. Tr. D-224, at 10.
- 17. For the non-seismic review, the IDVP selected three systems: the auxiliary feedwater (AFW) system, the control room ventilation and
)
pressurization (CRYP) system, and the safety-related portion of the 4160V electrical distribution system (4160V). The IDVP conducted a comprehensive review of all safety-related aspects of these systems, including a review of the systems design requirements. The AFW and CRYP systems were reviewed for effects of high energy and moderate energy line breaks. Analyses used to f
define radiation, pressure-temperature, and humidity environments for such systems were also reviewed for the AFW and CRYP. Anderson et al., ff. Tr.
D-224, at 10, 11. The design work and methodology reviewed by the IDVP in these specific systems is generic to all safety-related systems of the plant.
Cooper et al., ff. Tr. D-1459, at 1/2-24.
- 18. The AFW system was selected because its design represents an interrelationship of several design criteria and interfaces. Specifically, it involves interface with NSSS vendor criteria, with containment design criteria, interfaces of PGandE internal design organizations, and the methodology of determining a water system's mechanical, electrical, and control component design criteria. In addition, AFW systems often appear in the dominant accident sequences in various probabilistic risk assessment programs. Cooper et al., ff. Tr. D-1459, at 1/2-22.
- 3. The CRYP system was selected because it too represents an interrelationship of several design criteria and interfaces. Specifically, it involves interfaces with a service-related contractor, interfaces of PGandE i
i
i internal design organizations, and interface with the control room habitability criteria.
It also represents a contrast of design methods since it is an air system rather than a water system. Cooper et al., ff. Tr.
D-1459, at 1/2-22.
- 20. The 4160V system was selected because it is the basic power supply system for safety-related electrical equipment. It also represents an i
interrelationship of several design criteria and involves the interfaces a'nong several PGandE internal design organizations. Cooper et al., ff. Tr. D-1459, at 1/2-22.
- 21. The three systems selected were designed by three different engineering L
l groups within PGandE which performed design work in other systems in the plant. All of the electric systems were designed by PGandE. Anderson et al.,
ff. Tr. D-224, at 25,
- 22. The IDVP verified the fire protection provided for the sample systems, including the separation, fire barriers, suppression and detection systems provided in areas of the plant containing say le system components. Cooper et al., ff. Tr. D-1459 at 1/2-23.
- 23. The IDVP also verified that the AFW and CRVP system components were adequately protected from the affects of internally generated missiles.
Cooper et al., ff. Tr. D-1459 at 1/2-23.
- 24. The IDVP not only performed very detailed and comprehensive reviews of three sample systems which included all the PGandE internal design groups responsible for non-seismic safety-related system design, but the IDVP verification also included work by the service-related contractor who provided the most significant input into the safety-related system design. In addition, the IDVP performed many verifications of analysis and design
_g.
t
functions that are generic to the design or design methodology of all safety-related systems. The latter reviews included work from the various PGandE design groups as well as from all service-related contractors performing significant non-seismic design analysis. Cooper et al., ff. Tr.
D-1459, at 1/2-25.
- 25. Samples were added to the program if the IDVP's evaluation suggested the 5
existence of a potential generic concern. Cooper et al., ff. Tr. D-1459, at 1/2-24,1/2-31-32.
- 26. There was no restriction upon the IDVP's authority to expand any sample if it was deemed necessary by the IDVP. Krechting. Tr. D-1711. Additional sampling was perforned if significant deficiencies in the QA program or its implementation were identified during the QA audit of any organization not included in the initial sample or if additional information was needed to explain the discrepancies that had been found during the course of the design process verification. Anderson et al., ff. Tr. D-224, at 11.
E. If a generic concern was identified, its effect on other safety-related structures and components was evaluated also. Anderson et al., ff. Tr. D-224, at 11.
- 28. Rather than using random sampling in the selection of samples for review and verification, the IDVP employed engineering judgment based on its extensive experience and a detailed knowledge of the important safety considerations and the design steps and interfaces involved in the engineering design process. From this base of knowledge, those combinations of structures, systems, components and design activities which were judged to represent the breadth and depth necessary to draw conclusions regarding design adequacy were selected and reviewed. The IDVP developed procedures and.
1
methodology to identify possible generic safety concerns and to expand the samples both vertically and horizontally in response to such concerns.
Anderson et al., ff. Tr. D-224, at 15; Cooper et al., ff. Tr. D-1459, at 1/2-22.
Scope of the Internal Technical Program 29_. The other portion of the verification program, the ITP, was conducted by the DCP organization.
It carried out its own comprehensive design l
verification and provided a controlled, coordinated interface with the IDVP.
Anderson et al., ff. Tr. D-224, at 3.
_30. The ITP provided information requested by the IDVP, responded to the findings of the IDVP, and under the Corrective Action Program (CAP) implemented all corrective actions in response to the findings of the IDVP and the ITP. Anderson et al., ff. Tr. D-224, at 4.
l
- 31. The design verification efforts of the ITP consisted of a review of both the seismic and non-seismic design of DCNPP. All safety-related design work I
performed by the ITP was conducted in accordance with an NRC-approved quality l
assurance program which met the criteria of 10 CFR 50, Appendix B.
Anderson et al., ff. Tr. D-224, at 5, 7.
l ITP - Seismic l
l
- 32. The ITP's seismic review included a 1005 review of the seismic design for all safety-related structures, systems, and components designed by PGandE and its service-related contractors. Anderson et al., ff. Tr. D-224, at 6.
- 33. Non-seismic loads were included in the load combinations utilized in the
(
seismic review of structures, piping, electrical raceways, HVAC ducts and instrument tubing. Anderson et al., ff. Tr. D-224, at E.
- 34. All other non-seismic attributes of the PGandE-designed piping, such as I
~..
thermal expansion and dead weight load analysis, were incorporated into the ITP review and were reviewed along with the seismic attributes by the IDVP.
Shipley, Tr. D-298.
ITP - Non-Seismic
- 35. The ITP conducted a functional design review of all PGandE-designed, safety-related, mechanical systems:
(a) the component cooling water (CCW) system; (b) the diesel engine-generator (DEG) systems; (c) the auxiliary feedwater (AFW) system; (d) the auxiliary saltwater (ASW) system; (e) the fire protection system (portions); (f) containment hydrogen venting system; and (g) containment spray system. Anderson et al., ff. Tr. D-224, at 17.
- 36. A separate instrumentation and controls (I&C) review of all the systems referred to immediately above and the containment isolation system included reviews of the ability of control valves to meet system requirements, classification of instrumentation, and environmental and seismic qualification of instruments. The environmental qualification files were reviewed for changes resulting from reviews of pressure, temperature, and radiation.
Anderson et al., ff. Tr. D-224, at 17.
- 37. Functional design reviews were performed for all safety-related electrical systems:
(a) The 4kV system; (b) 480V ac system; (c) 115V ac l
system; (d) 124V de system; (e) electrical heat trace for the boric acid system; (f) electrical raceways; and (g) electrical penetrations. Anderson et al., ff. Tr. D-224, at 17, 18.
- 38. A functional design review of all the Class I HVAC systems was l
performed. These systems included:
(a) control room ventilation and pressurization; (b) auxiliary building ventilation; (c) fuel handling building ventilation; (d) DC/480 volt switchgear ventilation; (e) 4kV switchgear O
i
= __-
(
ventilation; (f) forced draft shutter for reactor vessel cavity; (g) diesel generator compartment ductwork; and (h) auxiliary saltwater ventilation Also included were the containment fan cooler units designed by the system.
NSSS vendor. Anderson et al., ff. Tr. D-224, at 18.
- 39. In addition to the detailed review of each system itself, each safety-related system was also reviewed by the DCP to assure that it was adequately protected from the various postulated hazards that the NRC regulatio:ts required to be considered. Anderson et al., ff. Tr. D-224, at 18, 19.
- 40. When the IDVP identified concerns that were potentially generic, another eview was performed by the DCP for that specific concern for all PGandE-designed safety-related systems and was verified by the IDVP. These reviews and verifications were performed in all areas of analyses of pressure, temperature, and humidity due to high energy line break (HELB) outside containment; selection of system design pressure and temperature; selection of differential pressure across power-operated valves; redundancy of power supplies for shared systems; separation and single failure criteria for mutually redundant circuits; and jet impingement effects of HELB inside I
containment. Cooper et al., ff. Tr. D-1459, at 1/2-24; App. Ex. 139, 140.
_41.
If an item was determined not to have met licensing criteria by the IDVP, that item was reported to DCP for corrective action. The IDVP then performed verifications of DCP corrective actions. When IDVP verification of a corrective action indicated that the corrected item met licensing criteria, f
the item was considered closed. If verification indicated that the corrective action did not meet licensing criteria, the item was again reported to the DCP l
for continuation of correction action. Cooper et al., ff. Tr. D-1459, at l i
o 1/2-26-27.
Conclusions of the Verification Program 4_2. Together, the IDVP and ITP reviewed 100% of the seismic design (Anderson 2
et al., ff. Tr. D-224, at 6; Cooper et al., ff. Tr. D-1459, at 1/2-13) and 755 to 80% of the non-seismic design originally performed by PGandE and its service-related contractors. Anderson, Tr. D-1420 D-1441.
43_. As a result of its extensive review, the IDYF, in its collective judgment, determined that while it is possible, indeed likely, that there remain errors in the sense of a failure to exactly meet each and every licensing criteria in the corrective action program work, it is highly unlikely that any such errors are significant. Cloud, Tr. D-1543. The IDVP also determined that it was highly unlikely that any significant undetected errors exist in the unreviewed design work.' Cooper et al., ff. Tr. D-1459, at 1/2-32; Cooper, Tr. D-1539; App. Ex. 90; IDVP Final Report, Volume 3. Page 6.2.5-2.
No significant errors were found by the ITP in its review effort.
Anderson, Tr. D-1420.
- 44. A significant error was considered to be a violation of a licensing criteria which resulted in a generic concern or posed a substantial safety hazard as defined in 10 CFR 50, Part 21. Cooper, Tr. D-1539.
45.
It also was the considered engineering judgement of qualified representatives of the Applicant that none of the findings of the ITP or IDVP, even considering a reduction of margin, would have resulted in the failure of any structure or system to perform its intended safety function or constituted an errer of safety significance as defined in 10 CFR Part 21. Anderson, Tr.
D-345, D-346; Anderson et al., ff. Tr. D-224, at 14. The Staff also testified that none of the deficiencies discovered by the verification program was l.
I significant or constituted a substantial safety hazard. Knight, Tr.
i D-2696-697; D-2819.
- 46. The verification program has taken place over a two-year period and has
~
InvolvedtheeffortsofhundredsofengineersfromPGandE,BechtelPower Corporation, and the IDVP. Over one million aanhours have been expended in carrying out the program. Sixty-eight in-depth technical reports and a 1
comprehensive Final Report have been issued by the IDVP. Over 50 public meetings, mostly technical, were held during the course of the program.
Anderson, et al., ff. Tr. D-224, at 4.
Through a semi-monthly reporting system, the NRC continuously was appraised of the findings and progress of the program. App. Ex. 88, 89.
- 47. Because of this extensive effort and the satisfactory resolution of all Ioncernsthatwereraised,theBoardfindsthatthereisreasonableassurance that the design of DCNPP is in compliance with licensing criteria and the rules and regulations of the Commission and that the activities authorized by an operating license can be conducted without endangering the health and safety of the public.
Contentions 1(a) and 2(a)
- 48. Contentions 1(a) and 2(a) maintain that the scope of both the seismic and non-seismic aspects of the design of safety-related systems, structures, and components was too narrow in that the neither the IDVP (la) nor the ITP (2a) verified samples from each design activity.
- 49. Pursuant to requirements of Comission Memorandum and Order CLI-81-30 (App. Ex. 86) and the Staff Letter of November 19, 1981 (App. Ex. 87), the verification program plans (App. Ex. 88 and 89) were submitted and approved by the Commission. Knight et al., ff. Tr. D-2649, at 10, 11; App. Ex. 156, 157,
- k
(
O s
158, 159.
- 50. For non-seismic design, the Staff originally had required a review of "a suitable sample of the activities performed by DCNPP Unit 1 by each service-related contractor that was completed subsequent to January 1,1978."
App. Ex. 87, at 3.
- 51. The IDVP reviewed the work scope of all contractors and included those who had made a significant contribution to the final design work. Cooper et al., ff. Tr. D-1459, at 1/2-7-11; Cooper, Tr. D-1470; Knight Tr. D-2740, D-2742; App. Ex. 101. The IDVP's approach was subsequently accepted by Comission approval of SECY 82-414. App. Ex. 157, 159; Knight Tr. D-2746.
- 52. The Board finds that the sampling process employed by the IDVP and ITP meet the requirements of the Commission Order and Staff Letter, and is of sufficient scope to provide reasonable assurance that Units 1 and 2 of DCNPP are designed in accordance with licensing criteria.
Contentions 1(b) and 2(b)
- 53. Contentions 1(b) and 2(b) maintain that the scope of the verification program of both the seismic and non-seismic aspects of the design of the safety-related systems, structures, and components was too narrow in that, in the design activities that were reviewed, neither the IDVP (Ib) nor the ITP (2b) verified samples from each of the design groups in the design chain performing the design activity.
- 54. The verification program which received Commission approval (Knight et al., ff. Tr. D-2649, at 10, 11) implemented the Order and the Staff Letter, each of which required that "a suitable number and type of sample calculations related to design" be selected for the purpose of verifying the design process.
- 55. The reanalysis and verification of all of the seismic design (Anderson et.
t
al., ff. Tr. D-224, at 6; Cooper et al., ff. Tr. D-1459, at 1/2-13) and the extensive amount of the non-seismic design originally performed by PGandE and its service-related contractors (Anderson, Tr. D-1420 D-1441) constitute a suitable nunber and type of sample for the purpose of verifying the design process.
- 56. The Board finds that the sampling process of the IDVP and ITP meet the requirements of the Comission Order and Staff Letter, and is of sufficient scope to provide reasonable assurance that Units 1 and 2 of DCNPP are designed in accordance with licensing criteria.
Contentions 1(c) and 2(c)
- 57. Contentions 1(c) and 2(c) maintain that the verification program was too narrow in that it did not have statistically valid samples from which to draw conclusions.
- 58. In approving the IDVP the Comission ordered that the program include
" consideration of" the use of a statistician (App. Ex. 158). There was no Comission requirement that the IDVP be based on or employ statistical sampling methods. Knight et al., ff. Tr. D-2649, at 21; Cooper et al., ff.
Tr. D-1459, at 1/2-32.
- 59. In the program management plans, the IDVP comitted that "TES will arrange for an evaluation of the entire Phase II program by an expert in the applicability of statistics to engineered systems." App. Ex. 88, 89, App. C.
- 60. Contained within the Staff approval of the program was the Staff coment hat " Rigorous statistical techniques are largely inappropriate for design verification programs." App. Ex. 157, App. 11. The Staff recomendation was l
l approved by the Comission on December 9,1982. App. Ex. 159.
- 61. The IDVP thereafter did not retain an expert in the applicability of l 1 L
t
statistics to engineered systems because it did not consider its commitment in the proposed program plan for Phase II to be operative. Cooper Tr. D-1515.
- 62. Consistent with the Order of the Commission, the IDVP considered and dismissed the use of " formal" statistics in its proposed verification program. Cooper et al., ff. Tr. D-1459, at 1/2-32; Cooper, Tr. D-1513, D-1515; App. Ex. 90, Section 3.5.
- 63. Dr. Samaniego, a professor of statistics, testified on behalf of the JI that because the IDVP employed judgment sampling in their review and not random sampling, there is an absence of a " scientifically rigorous, systematic methodology" to justify the IDVP conclusion about the general characteristics of DCNPP and that the IDVP statements on conformance of the design of the plant have no " scientific validity." From that, he concluded that the IDVP's finding of reasonable assurance is without a reliable, objective basis.
Samaneigo, Tr. D-2392, at 7.
- 64. On examination, Dr. Samaniego testified that without the use of random sampling techniques, there is no way to assess validity, and that any conclusion drawn would be out of the range of science and in the range of opinion. The accuracy of such conclusions would depend upon the skill of the person rendering the opinion.
Samaniego Tr. D-2429. On further examination, Dr. Samaniego admitted that when in his testimony, he referred to " science" or l
the " scientific method," he was only speaking in the sense of " statistical science." Samaniego, Tr. D-2430-31.
- 65. Dr. Samaniego testified that he was not a nuclear engineer or an engineer of any sort and that his testimony was directed toward statistical t
methodology, the methodology of samp'ing. Samaniego Tr. D-2407. He was unable to give any references in statistical literature where statistical,
i
sampling techniques previously had been used to verify design. Samaniego Tr.
D-2408-09 D-2452.
- 66. Dr. Apostolakis, the Governor's witness, testified that where one is assessing probabilistics, statistical evidence (information concerning the frequency with which a given attribute is observed in a specified population) can be most useful. Apostolakis, ff. Tr. D-2313, at 4, 7.
Dr. Apostolakis testified that, assuming that there existed well defined events, it would be possible to estimate the frequency of design errors in a nuclear power plant using random sampling techniques. Aposto1akis, ff. Tr. D-2313, at 13-14. Dr.
Apostolakis further testified that he did not believe that a sample drawn non-randomly could validly be used to generalize about the frequency of error (error rate) in the unsampled portion of the plant. Aposto1akis, ff. Tr.
D-2313 at 16, 19.
- 67. On cross examination, Dr. Apostolakis acknowledged that in a design verification program there would be considerable difficulty in designing a random sampling program and estimating an error rate. Apostolakis, Tr.
D-2334. Dr. Apostolakis had made no attempt to delineate the universe of design decisions to be sampled and had no evidence that any such listing of design dec.isions had ever been done for any nuclear power plant in the world.
Apostolakis, Tr. 0-2337.
6_8.
Dr. Apostolakis conceded that there was no generally accepted definition of design error, and that he had never talked to knowledgeable people such as l
the IDVP as to the feasibility of devising a workable or realistic definition. Apostolakis, Tr. D-2342.
- 69. Dr. Apostolakis acknowledged that after the difficulties of defining the population of design decisions and defining what constituted an error, there.
would be further difficulties in using the error rate for decision making. He admitted that he wouldn't know how he or a member of the Appeal Board would use such an error rate in decision making. Apostolakis, Tr. D-2330, D-2354.
- 70. Dr. Apostolakis testified that where random sampling techniques were not 0
employed but engineering judgment was used, one could have reasonable assurance of the presence or absence of something. Apostolakis, Tr. D-2330.
- 71. PGandE presented the testimony of Dr. Stanley Kaplan, an expert in probablistics and decision theory who holds a Ph.D. in mechanical engineering and applied mathematics. Kaplan, ff. Tr. D-ll61, Affidavit of Professional Qualifications; Kaplan, Tr. D-1347. D-1354.
- 72. Dr. Kaplan testified that the use of statistics is inappropriate for design verification. Kaplan, ff. Tr. D-1161, at 5-6.
In applying statistical techniques to a design verification program the result obtained is a frequency i
I or the fraction of design elements which is in error. Kaplan, ff. Tr. D-1161, at 17; D-1282.
- 73. The Staff also testified that a statistical compilation of various categories of design deficiencies is of little use in making a final determination regarding plant safety. Knight et al., ff. fr. D-2649, at 23.
4.
Dr. Kaplan believed the question before this Board to be whether the verification program has provided reasonable assurance that no design errors exist in the system design of the plant which would endanger public health and safety. Kaplan, ff. Tr. D-1161, at 67; D-1188-89.
- 75. Dr. Kaplan testified that the issue, then, is not the frequency of errors, i.e., violations of criteria, but rather the quality of design and whether there are any safety-significant errors remaining in the plant.
Kaplan, D-1191, D-1325. Dr. Kaplan testified that, based on his review and t,
i
analysis, he was highly confident that there were no errors in the system design that would constitute a threat to public health and safety or result in outage of the plant. Kaplan, ff. Tr. D-1161, at 63.
H. The Board finds that use of statistical methodology would not have greatly enhanced the design verification process as carried out by the verification program, and that the procedures used do provide reasonable assurance that Units 1 and 2 of DCNPP are designed in accordance with licensing criteria.
Contention 1(d)
H. Contention 1(d) maintains that the IDVP review of the seismic and non-seismic design and safety-related systems, structures, and components, was too narrow in that it merely checked data of inputs to models used by PGandE and failed to verify the ana7 sis independently.
/
H. The Governor and JI elicited agreement from the IDVP, on cross-examination, that an independent analysis that was done by Brookhaven National Laboratory (BNL) did reveal discrepancies in the seismic model of PGandE. Wray, Tr. D-1930; Cloud, Tr. D-1937-39. The IDVP, however, had l
already uncovered those discrepancies by their own work independently of Brookhaven's analysis. Biggs, Tr. D-1933, D-1937, D-1941.
H. Dr. Cloud testified that in the first phase of the verification program the IDVP started out doing completely independent verification by means of completely independent calculations and that when questionable areas or discrepancies turned up, the IDVP did additional independent analyses. Cloud, Tr. D-1939.
80_.
When the DCP made its decision that it would make a 100% reexamination of the seismic design at DCNPP, the IDVP changed its verification to a design 1.
review approach and regularly did alternate calculations. Cloud Tr. D-1940.
- 81. Because the seismic analyses by the DCP of the containment building, the lurbinebuilding,theauxiliarybuilding,theintakestructure,andthepiping and components were state-of-the-art, the IDVP concluded that independent analysis was unnecessary, but that it was important for the IDVP to verify that what was being done was done correctly. Cloud, Tr. D-1940-42. Had the IDVP done a completely independent analysis of the annulus, for example, there would have been a duplication of effort because the IDVP agreed that the DCP was using appropriate procedures. Biggs, Tr. D-1941-42.
- 82. The IDVP did independent calculations for the diesel fuel oil tank, and the annulus structure was completely redone by the DCP with state-of-the-art analysis. Cloud, Tr. D-1941.
- 83. When asked whether the use of a second independent model would give him more confidence that there remain no design errors in any structure at DCNPP, Professor Biggs, Professor Emeritus in the Department of Civil Engineering of the Massachusetts Institute of Technology, testified, "very little more."
Biggs. Tr. D-1942.
- 84. The Governor's seismic expert, Dr. Roesset, admitted on cross examination f
that independent analyses of the type done by BNL, while valuable, were not necessary. Roesset. Tr. D-2247.
l
- 85. There was no requirement of the Conmiission Order or Staff Letter that the IDVP independently verify any seismic analyses by the use of alternative seismic models of structures. App. Ex. 86, 87.
i
- 86. The Board finds that employment of independent analyses in addition to the efforts of the IDVP is not necessary for the verification of the design of DCNPP, and that the procedures used do provide reasonable assurance that Units j
l l 1
I and 2 of DCNPP are designed in accordance with ifcensing criteria.
Contention 1(e) 87_.
Contention 1(e) maintains that the IDVP review of the seismic and non-seismic design of safety-related structures, systems, and components was too narrow in that the IDVP failed to verify the design of Unit 2.
- 88. The IDVP did not verify the design of Unit 2.
Its review was performed inaccordancewiththeCommissionorderandStaffLetter(App.Ex.86,87) which contemplated only an independent verification of Unit 1 and in accordance with approved Commission Program Plans which included only Unit 1.
Cooper, ff. Tr. D-1459, at 1/2-35-36; Knight et al., ff. Tr. D-2649, at 5.
- 89. The Board finds that the limitation of the scope of the IDVP review to Unit I was in accordance with the Commission Order and Staff Letter and that further expansion is not necessary because of the similarity of design between Units 1 and 2 and the integrated verification and corrective action program of the ITP (see findings under Contention 2(d) infra.), which provide reasonable assurance that Unit 2 is designed in accordance with licensing criteria.
Contention 2(d)
- 90. Contention 2(d) maintains that the ITP review of the seismic and non-seismic design of safety-related systems, structures, and components was too narrow in that the ITP failed systematically to verify the adequacy of the
[
design of Unit 2.
- 91. At DCNPP, the majority of the structures, systems and components are either comon to both units or essentially identical for Units 1 and 2.
The i
units were designed with essentially identical arrdngements, systems and components, and operational characteristics. Anderson et al., ff. D-224, at
- 28. The systems were essentially designed on Unit 1 with very little design 1 l l
work being done on Unit 2.
Anderson, Tr. D-1321.
- 92. The engineering group responsible for the design of DCNPP did the design for both units. A single PGandE organization originally developed and used the same design criteria for both units. Anderson et al., ff. D-224, at 29.
Piping schematics and instrumentation schematics were the same for both 93.
units. Generally, equipment was purchased from the same vendors using the same specifications. Seismic and environmental equipment qualification criteria and guidelines were the same for both units. Anderson et al., ff.
D-224, at 29.
9_4. While there are opportunities for errors in systems design work in Unit 2 which would not necessarily appear in Unit 1, such opportunities would be minimal. Anderson, Tr. D-1321-22; Wermial Tr. D-2776.
- 95. To check the applicability of the Unit 1 IDVP design verification to Unit 2, the DCP established an internal procedure. Anderson et al., ff. Tr. D-224, at 29; Moore, Tr. D-385.
- 96. Under that procedure, any Unit 1 finding that was found not applicable to Unit 2 was documented with the basis for this decision.
If the finding applied to both Units 1 and 2, a determination was made as to whether the Unit
(
1 resolution also applied.
In cases where the Unit 1 resolution applied to l
the Unit 2 design, procedures ensured that the resolution was implemented for Unit 2.
If a finding deemed applicable to Unit 2 involved physical modifications to the plant, the appropriate design change document was issued to PGandE's General Construction Department for implementation on Unit 2.
Anderson et al., ff. Tr. D-224, at 29-30; Cranston, Tr. D-384-85.
- 97. If the substance of the ITP or IDVP review item was not identical for Ethunits,theDCPevaluatedanddocumentedthedifferencesandthe
( l I
l t
1
applicability to Unit 2.
A determination was made whether the item required resolution for Unit 2, and the effect of the differing resolution of the review item on Unit 2 was evaluated and documented. Before implementing the Unit 2 resolution, the Unit 2 Project Engineering group reviewed the resolution to establish or confirm that it was consistent with licensing l
criteria and that appropriate action was taken to ensure that the Unit 2 requirements were satisfied. Anderson et al., ff. Tr. D-224, at 30; Cranston, Tr. D-384-85.
- 98. Unit 2 does have an ongoing systematic evaluation program to review all iarge-boreandsmall-borepiping,includingsupports. Shipley, Tr. D-387-88.
A complete reanalysis of all small-bore piping has been done for Unit 2.
Shipley, Tr. D-393-94. In addition, all common lines that apply to Unit 2 as well as Unit I and structural areas have been reviewed for Unit 2.
Moore Tr.
D-388.
- 99. While the ITP review of Units 1 and 2 was not documented as a review hocess,itwassystematicinthattheITPreviewwasaccomplishedin accordance with existing engineering procedures and was documented in the files as the design process. Anderson, Tr. D-1426.
l 100. The Board finds that the verification effort of the ITP was systematic and does provide reasonable assurance that Unit 2 is designed in accordance i
with licensing criteria.
l Contention 3 101. PGandE presented the testimony of Mr. Richard C. Anderson, Dr. H. Bolton Seed, Mr. Larry E. Shipley, and Dr. William H. White. Dr. Jose M. Roesset testified on behalf of the Governor and Dr. Robert L. Cloud, Professors J. M.
Biggs and M. J. Holley, Jr., and Mr. Ronald Wray testified on behalf of the.
't t-
IDVP. The Staff presented testimony of James P. Knight, Pao-Tsin Kuo, Harold E. Polk, Dr. Charles A. Miller Dr. A. J. Philippacopoulos, Dr. Carl J.
Costantino, and Dr. P. C. Wang.
Contention 3(f)(1) 102. Contention 3(f)(1) maintains that the ITP used improper engineering standards in its modeling of soit properties for the containment building for the DE and DDE in that boundary motion inputs were improperly used.
103. For the containment seismic analysis for the DE and the DDE, the foundation rock mass and the containment system, together with the interior concrete, were modeled as one structure to consider the effect of the rock-structure interaction. The rock main boundary dimensions were found to be large enough, when compared to the structure dimension, that the boundary conditions of the rock mass did not have a significant effect on the rock-structure interaction. White et al., ff. Tr. D-651, at 69.
104. For the seismic analysis of the containment / rock model, the boundary motions, not the free field ground motions, were required as input for the model. The adequacy of the boundary motion to produce the specified rock surface motions was checked by using them to compute a response spectrum at the containment location in the absence of any structure. The resulting spectrum compared favorably with the FSAR prescribed input spectrum, thus justifying the appropriateness of the boundary motions. These same boundary motions were then used for analysis of the structure / rock model for generating the dynamic responses of the containment structure. White et al., ff. Tr.
D-651, at 69-70. Neither the Governor nor JI offered any testimony on this contention.
105. The NRC Staff considered the method used to apply boundary motion to be.
t k
~
acceptable. Kuo, ff. Tr. D-2463, at 14-15.
106. The IDVP testified that the Hosgri earthquake, and not the DE or DDE, in general tends to govern the design of the containment building and equipment area and that the results of a soil / structure interaction analysis for the DE and DDE would be irrelevant. Cloud, Tr. D-1856.
)
i 107. The Board finds that appropriate ground motions were properly applied for the containment seismic analysis and that licensing criteria have been met.
Contention 3(f)(ii) 108. Contention 3(f)(ii) maintains that the ITP used improper engineering standards in its modeling in that the soil structure interaction analysis for containment for the DE and DDE uses a 75 damping value for rock.
109. Applicant's FSAR states that for the rock foundation 7% of critical damping is to be used to compute response spectra for the DE and DDE analysis. Only for the mode with the largest participation of rock was 7%
used in the ITP's analysis. Lesser values of 2% and 5% were used in the other modes which resulted in greater conservatism in its analysis than was required. White et al., ff. Tr. D-651, at 71. Neither the Governor nor JI offered any testimony on this contention.
110. For the soil structure interaction analysis for containment for the DE
(
and DDE, the damping values associated with the structure and soil are in l
conformance with those prescribed in the FSAR and thus are in conformance with I
licensing criteria. White et al., ff. Tr. D-651, at 70, 71. The Staff testified that the damping values used were not excessive. Kuo, ff. Tr.
D-2463, at 15.
111. The Board finds that the ITP's use of damping values was proper and in accordance with licensing criteria..
1 Contention 3(f)(iii) 112. Contention 3(f)(iii) maintains that the ITP used improper engineering f
standards in that its dynamic analyses of the containment for all earthquakes omit any analysis of uplifting of the foundation mat.
113. Uplifting is a phenomenon which represents separation between the supporting media and the bottom face of the foundation slab of a structure.
White, Tr. D-668.
114. Dr. Roesset, on behalf of the Governor, testified that while uplift has bensificial effects at DCNPP it could also have detrimental effects in that stresses in the lower part of the containment shell may increase and that vertical accelerations may increase due to the shifting of the axis of rotation.
Roesset, ff. Tr. D-2206, at 6-7.
115. The IDVP witnesses testified that, in the event of a Hosgri earthquake, there was no certainty that separation or uplift would occur but that if it did, some increase in vertical stresses might occur. This increase would have to be combined with the other three components of earthquake motion (two horizontal and one vertical) by the square root of the sum of the squares method because they do not peak at the same time. Further, the horizontal components of acceleration are reduced by uplift as is the vertical acceleration from the vertical ground motion. The net result is that the effect of uplift is negligible and may be less severe than the environment without uplift. Holley, Tr. D-1874-81; Biggs. Tr. D-1881-86; Cloud, Tr.
1886-94.
116. As a part of its extensive verification effort, the ITP examined the effect of potential uplift of the containment foundation mat. The results of the analysis showed that, for the DE, DDE, and Hosgri earthquake, the maximum.
t 1
stresses in the reinforcing steel were within the allowable limits :pecified in the FSAR and Hosgri Report. White et al., ff. Tr. D-651, at 71, 72.
117. Dr. Roesset, on cross examination, conceded that most of the effects from uplift are beneficial and that the possible effects that may not be beneficial are not very large. Roesset, Tr. D-2271.
118. Dr. Seed testified that in all his experience no one has ever described any damage due to the effects of uplift where earthquake damage was surveyed.
Seed Tr. D-683-84.
119. There is no regulatory guideline that requires censideration of uplift.
White, Tr. D-681. All dynamic analyses performed on the containment structure are based on the assumption that the soil / structure interface does not separate. Not only has this been a standard industry practice for the seismic analysis for nuclear power plant structures, but it i:; generally believed that the effect of such separation is small and negligible in the overall response of the structure. Furthermore, rigorous procedures are not available for including the effect of separation (i.e., uplift) at the soil / structure interface in the seismic analysis. Current research programs are studying the form of the interface model which would be required to consider the uplift effects but are not developed sufficiently for general use. Kuo, ff. Tr.
D-2463, at 15-16.
120. The Board finds that the ITP used proper engineering standards in its dynamic analysis of the containment building, that the effects of uplift are not required to be considered by licensing criteria, and that even if its effects were considered, the maximum stresses within the reinforcing steel in 1
the containment are withia the FSAR and Hosgri criteria.
//.
Y
Contention 3(f)(iv) 121. Contention 3(f)(iv) maintains that the ITP used improper engineering standards in that soil properties were not specified in the modeling of soil springs for the auxiliary butiding.4 122. The soil properties used for determining the soil springs in the dynamic modeling for the auxiliary building were specified based upon extensive soil and rock investigations and laboratory testing. The parameters obtained from these data provided the values of the soil springs included in the seismic model of the auxiliary building. White et al., ff. Tr. D-651, at 72-73. The results of this parametric study concluded that this variation in soil springs had minimal effect on the response of the auxiliary building. The IDVP has concurred with the DCP finding. Also, the NRC Staff finds the soil parametric study acceptable. White et al., ff. Tr. D-651, at 72-73; Kuo, ff. Tr. D-2463, at 16; White Tr. D-699-719; Cloud et al., ff. Tr. D-1843, at 3-8; App. Ex.
145.
123. The Board finds that the ITP used proper engineering standards and that it did properly specify soil spring properties in its modeling of the auxiliary building consistent with licensing criteria.
//
//
l i
- 4. Governor witness Roesset testified on rebuttal that his concerns regarding Contentions 3(f)(iv) and (v) were satisfied if the numbers in Table 5 of ITR-55 (App. Ex. 145) are correct (Tr. D-2218-21. D-2249). On rebuttal, IDVP witness Cloud testified that the numbers in Table 5 had been rechecked and the numbers are correct. Tr. D-3111. However, apparently the contention has not been dropped. Tr. D-2496.,
i
Contention 3(f)(v) 124. Contention 3(f)(v) maintains that the ITP's modeling of soil properties for the auxiliary building was improper in that the motion inputs to the lower ends of the springs do not account for all soil structure interaction phenomena that could be expected.
125. The effect of soil structure interaction in the mathematical models for the seismic evaluation of the auxiliary building is represented by soil springs in accordance with the licensing criteria set forth in the FSAR and the Hosgri Report. The IDVP found the model and method of performing seismic dynamic analysis of the auxiliary building and its soil springs to be appropriate, as did the Staff. White et al., ff. Tr. D-651, at 73; Kuo, ff.
Tr. D-2463, at 17; Cloud et al., ff. Tr. D-1843, at 3-9.
126. The Board finds that the ITP used proper engineering standards and that the motion input is in conformance with the applicable seismic criteria for the auxiliary building.
Contention 3(o) 127. Contention 3(o) maintains that the ITP modeling of the fuel handling building is improper in that neither the use of translational and torsional response from the auxiliary building as an input used nor the validity of the dynamic degrees of freedom used has been adequately justified.5 l
l 128. In its modeling of the fuel handling building, the ITP used translational i
j and torsional response from the auxiliary building as an input. Cloud et al.,
ff. Tr. D-1843, at 3-18.
l l
- 5. Neither the Governor nor Joint Intervenors offered any evidence either directly or by way of cross examination on this contention. !
i
129. The fuel handling building rests on the auxiliary building and during an earthquake the excitation experienced by the fuel handling building, including translational and torsional response, would be the motion of the auxiliary building. Cloud et al., ff. Tr. D-1843, at 3-18.
130. Consistent with good engineering practice, the ITP first included the fuel handling building in the model of the auxiliary building and then utilized a more detailed model of the fuel handling building to obtain local Cloud et al., ff. Tr. D-1843, at 3-18; White et al., ff. Tr.
responses.
D-651, at 81-82.
131. The Staff testified that the procedure utilized in the fuel handling 1
building model is acceptable. Kuo, ff. Tr. D-2463, at 21.
132. The IDVP testified the ITP in its modeling of the fuel handling building selected the appropriate degrees of freedom, that the modeling was properly performed, and that the analyses done by the ITP meet all licensing criteria.
Cloud et al., ff. Tr. D-1843, at 3-18-19; Cloud Tr. D-1922. The Staff, too, found the modeling to be acceptable. Kuo, ff. Tr. D-2463 at 21.
133. The ITP developed a large comprehensive model for its static analysis of the fuel handling building. Following widely used engineering procedures, the large static model was divided into smaller parts and then condensed into two dynamic models with fewer degrees of freedom to make it manageable for dynamic analysis. A sufficient number of dynamic degrees of freedom were included to adequately determine peak accelerations. These peak accelerations were then used to obtain static loads which were applied to the large static model for purposes of member evaluation. The IDVP verified this process and the data transfer from the dynamic models to the static models. Cloud et al., ff. Tr.
D-1843, at 3-18-19; White et al., ff. Tr. D-651, at 81-82; App. Ex. 147..
i
'134. The Board finds that the ITP's mode 11ng of the fuel handling building is justified and proper and that there is no deviation from licensing criteria.
Contention 3(p) 135. Contention 3(p) maintains the ITP used improper engineering standards in that it has not been demonstrated that the modeling of the slabs in the auxiliary building is proper in relation to the use of vertical and rotational springs for columns and the motions used as input at the ends of springs not connected to the slabs. Contention 3(p) further maintains that inplane flexibility of the slabs has not been adequately accounted for and that stresses have not been demonstrated to be within allowable limits at all elevations.6 136. The IDVP verified that the ITP's modeling of the slabs and supporting columns and the choice of locations of input motions were proper for the purpose of evaluating the vertical response of the slab. Cloud et al., ff.
i Tr. D-1843, at 3-21; App. Ex. 145.
l
(
137. The Board finds that the ITP used proper engineering standards in its l
modeling of the slabs of the auxiliary building and that there has been no
(
[
deviation from licensing criteria.
Contention 3(q) 138. Contention 3(q) maintains that in the soils analysis for the buried diesel fuel oil tanks the values of the exponent in curves shown in Figure 14 l
- 6. Although 11 e Me nor's witness Roesset testified that he no longer had concerns re p.dtn3 W s contention ( Roesset Tr. D-2207-08, D-2250),
apparently it has not been dropped. Tr. D-2496.
// l
O of ITR 68 and the variation of shear velocity with depth have not been demonstrated to be justified or proper.
139. Dr. Roesset testified that this contention was raised because the values of the exponent in the curves showing the variation of shear modulus with mean effective stress in Figures 13 and 14 of ITR 68'(App. Ex. 155) did not seem reasonable, and the variation of shear wave velocity of the rock with depth implied by the foruula on page 38 and in Figure 15 of ITR 68 was not justified. Roesset, ff. Tr. D-2206, at 18.
140. The basis for the contention was a concern of Dr. Roesset over the apparent absence of documentation regarding actual sofi values. Roesset, Tr.
D-2265.
141. Dr. Harry Seed, a world renowned soils expert, testified on behalf of the Applicant as to the methodology and development of the curves shown on Figures 13 and 14. Seed, Tr. D-775-80, D-3115-119; Gov. Ex. 41. Dr. Seed pointed out that a substantial investigation had been performed for necessary data, which was referenced in Governor's Exhibit 41, a figure within the Report entitled, Geotechnical Studies: Diesel Fuel Oil Storage Tanks, Diablo Canyon Nuclear Power Plant. Seed, Tr. D-821, D-3115-119. That report was served on all parties and the Board on August 19, 1983, a fact of which we take judicial notice.
142. Dr. Seed explained in detail how the data were obtained and conservatively interpreted consistent with good engineering practice. Seed, Tr. D-3114-119. He also testified as to the reasonableness of the values as determined by standard industry practice and his extensive background and experience as a soils expert. Seed Tr. D-770-73. The IDVP also concluded that the ITP's soils modeling and analyses for buried diesel fuel off tanks.
h
were appropriate. Cloud et al., ff. Tr. D-1843, at 3-23, 3-24.
143. Dr. Roesset, on examination testified that the results of doing calculations over a range of soil properties would be small. Roesset Tr.
D-2265.
144. As to Figure 15 of App. Ex.155, Dr. Seed explained that both the shear wave velocity and compressional wave velocity tend to increase with depth and, at different locations different depths of overburden were excavated at the Diablo Canyon site so that there is no single relationship of shear wave or compression wave velocity which is applicable to the site. Seed, Tr. D-799.
145. Dr. Seed testified that knowing the accuracy with which compressional wave velocities are often measured, along with the variation of the properties of rock as they exist at the Diablo site, he was not surprised over the apparent scatter of the data. Seed Tr. D-801.
146. Dr. Seed also testified that he would have thought that the shear wave velocities at the Diablo Canyon site would have been higher, and that the use of lower velocities was conservative. Seed, Tr. D-832-33.
147. The Board finds that proper engineering practice was used in obtaining and analyzing the data that provided the basis for Figure 13,14, and 15 of ITR 68.
148. The Board finds that the data have been verified as being reasonable and that there is reasonable assurance that the design based on such data is consistent with licensing criteria.
Contention 3(r) 149. Contention 3(r) maintains that because the selection of the modulus strain curve has not been justified, the soils analysis for the auxiliary saltwater (ASW) piping and circulating water intake (CWI) conduits has not.
I t
been demonstrated to be proper.
150. Dr. Roesset testified that this contention was raised because the soil t
properties of the backfill seemed questionable. Roesset, Tr. D-2206, at 20.
151. Dr. Seed and Dr. White explained that the selection of a particular curve or material for the backfill was of no consequence because the conduits and
[
piping are encased in or constrained by rock. The response of the structures will not be influenced by the nature of any soil used as backfill but rather movement of the structure would be the same as the constraining rcck. Seed, Tr. D-836-40, D-3128, D-3142-43; White et al., ff. Tr. D-651, at 86; App. Ex.
55, Fig. 22, 23, 24.
152. Fill from the ASW piping and CWI conduit locations was analyzed and the results corresponded to the standard Seed and Idriss curve for sand, which was used in the ITP analysis. White et al., ff. Tr. D-651, at 86.
153. Dr. Seed explained that because the data obtained from laboratory tests fit the Seed-Idriss curve for sands conservatively, no correction factor was l
applied.
If a correction factor had been applied, the results would have led to lower computed movements to the structures being analyzed, i.e., the results would have been less conservative. Seed, Tr. D-3118.
154. The IDVP verified that the form of modulus versus strain curve used by the ITP was proper. Cloud et al., ff. Tr. D-1843, at 3-25. The Staff also concurred that the approach used by the ITP was proper. Constantino et al.,
D-2643, at 24.
155. The Board finds that the modulus versus strain curve was justified and that the soils analysis for the ASW piping and CWI conduits was proper.
Contention 3(s) l 156. Contention 3(s) maintains that it has not been demonstrated that the,
1
seismic analysis of the turbine building is proper in that bolt bearing capacities were taken from an inappropriate source.7 157. The acceptance criteria in Section 4.1.4 of the Hosgri Report specify the use of AISC, 7th Edition, Part 2, but specifically provide that the lateral force resisting elements are allowed inelastic deformation, in which case the allowable stress limitation of AISC, 7th Edition, Part 2 need not apply. In accordance with this criterion, the 3-bolt connections of the bottom chord roof members of the turbine building were~ analyzed by using an ultimate capacity of 229.5 kips as obtained from the provisions of the AISC, 8th Edition Part 2.
White et al., ff. Tr. D-651, at 86-7.
158. The Board finds that the bolt bearing capacities utilized were appropriate and in accordance with licensing criteria.
Contention 3(t) 159. Contention 3(t) maintains that the use of four different models for vertical analysis of the turbine building is not proper.8 160. The source of this contentien was a concern raised by BNL. However, on cross examination, a BNL witness indicated that his concerns were satisfied and that the use of four models was proper. Wang, Tr. D-2568-69.
161. The Staff considers that the use of four vertical models to be a reasonable approach to determine the vertical seismic response of the turbine building. Constantino, ff. Tr. D-2643, at 24-25.
- 7. Neither the Governor nor the Joint Intervenors presented any direct evidence or conducted cross-examination on this issue.
- 8. Neither the Governor nor the Joint Intervenors presented any direct evidence or conducted cross-examination on this issue.
l.
..., ~ _
162. The Board finds that there is no deviation from licensing criteria in the modeling/ analysis of the turbine building.
Contention 4 163. PGandE presented the testimony of Gary Moore; Richard C. Anderson; and Ed Connell, Jr. The IDVP presented the testimony of John E. Krechting and Dr.
William E. Cooper. The Staff presented the testimony of James P. Knight, on behalf of Dr. Mark Hartzman; Jared S. Wermiel; and John L. Knox. The Governor and Joint Intervenors presented no testimony on any of the contentions under number 4 and conducted cross examination on only 4(h), 4(1), 4(1), 4(q), 4(r),
and 4(t).
Contention 4(a)(i) 164. Contention 4(a)(1) maintains that the IDVP accepted the installation of the auxiliary feedwater system (AFWS) when a steem trap shown on design drawings had not been installed.
165. In reviewing a piping schematic, the IDVP discovered the absence of a steam trap to the turbine-driven AFW pump, which had been shown on a different piping schematic. Krechting et al., ff. Tr. D-2040, at 4-3; Moore et al., ff.
Tr. D-487, at 7.
166. After investigation, the IDVP determined that the design change had never been approved. The change was initially requested in order to alleviate a suspected problem. The problem, however, was resolved by other means and the system was successfully tested without the trap. Krechting et al., ff. Tr.
D-2040, at 4-3; Moore et al., ff. Tr. D-487, at 7.
167. Actual plant construction cor. formed to approved plant design, and the piping schematic was revised accordingly. Krechting et al., ff. Tr. D-2040, at 4-3; Moore et al., ff. Tr. D-487, at 7; App. Ex. 114. No deviation from
- t
licensing criteria occurred. Moore et al., ff. Tr. D-487, at 7; Wermiel ff.
Tr. D-2864, at 7.
168. The Board finds that installation of the trap was not necessary for the proper operation of the plant and that there has been no deviation from licensing criteria.
Contention 4(a)(ii) 169. Contention 4(a)(ii) maintains that the IDVP accepted the installation of the AFWS when there were discrepancies in the physical arrangement of the long-term cooling water supply line from design documents.
170. During a field investigation, the IDVP noted the installation of a check valve which was not shown on a piping schematic drawing of a branch ifne in the make up water system. Moore et al., ff. Tr D-487, at 8.
171. The removal of the valve from the reviewed piping schematic was a drafting error caused by misinterpretation of a drawing attached to a change notice which did not call for removal of the valve. Moore et al., ff. Tr.
D-487, at 8; Krechting et al., ff. Tr. D-2040, at 4-4.
172. The IDVP verified that the valve as installed met system functional requirements and was in accordance with the approved final design.
Consequently, no deviation from licensing criteria occurred. Krechting et a!.. ff. Tr. D-2040, at 4-4; Moore et al., ff. Tr. D-487, at 8-9; Wermiel, ff.
Tr. D-2864, at 7; App. Ex. 114.
173. The Board finds that installation of the check valve was in accordance with final approved design and that there has been no deviation from licensing criteria.
Contention 4(b) 174. Contention 4(b) maintains that contrary to the FSAR, the electrical.
k
t o
design does not fully comply with the commitments regarding separation and color coding.
175. The IDVP questioned whether two low energy pressure indicators met separation criteria of the FSAR. Moore et al., ff. Tr. D-487, at 9.
I 176. Consistent with industry practice, the subject criteria were not intended to apply to low energy instrumentation signals but rather to exposed conductors of mutually redundant power devices. Moore et al., ff. Tr. D-487, at 9; Krechting et al., ff. Tr. D-2040, at 4-5.
PGandE committed to revise the FSAR for clarification. Krechting et al., ff. Tr. D-2040, at 4-5.
177. The DCP analyzed instrument separation to show compliance with the current IEEE standard 384-1981. The instruments were also shown to be qualified to the IEEE flammability standards, with external wiring being wrapped in fireproof material. Moore et al., ff. Tr. D-487, at 10.
178. The IDVP in its review also noted that only some non-safety-related cable was color coded when the FSAR provides that safety-related cables should be f
coded. Moore et al., ff. Tr. D-487, at 10; Krechting et al., ff. Tr. D-2040, at 4-5-6.
179. PGandE committed to revise the FSAR section to clarify color coding requirements. Krechting et al., ff. Tr. D-2040, at 4-6; Moore et al., ff. Tr.
D-487, at 10. There were no deviations from licensing criteria regarding separation and color coding. Krechting et al., ff. Tr. D-2040, at 4-6; Moore et al., ff. Tr. D-487, at 10.
180. The Board finds that color coding for circuits not required for safe shutdown is permissible and that there has been no violation of licensing criteria.
Contention 4(h).
t
181. Contention 4(h) maintains that the IDVP accepted a deviation from a licensing comitment in that the effects of a moderate energy line break (MELB) on control room habitability had not been evaluated during safe shutdown.
182. A MELB analysis had not been conducted for a fire line in accordance with an existing licensing commitment because the line had been installed after the MELB review had been made. Anderson, Tr. D-492. PGandE had made a commitment to have the capability to safely shut down the plant after a postulated MELB.
Moore et al., ff. Tr. D-487, at 20.
183. The IDVP determined that, if a break occurred in the fire line, water could splash onto equipment of the control room ventilation and pressurization system (CRVPS) and render it inoperable with the result that the control room l
- could become uninhabitable. Anderson, Tr. D-490; Krechting et al., ff. Tr.
I D-2040, at 4-14, 4-15.
184. PGandE provided an analysis of the effects of postulated breaks on the CRYPS which showed that only one of the two redundant CRYPS trains would be affected by the resulting line break. The analysis showed that, assuming a concurrent single failure in the redundant CRYPS train, safe shutdown of the plant could still be achieved from the remote shutdown panel outside of the control room. Wermiel et al., ff. Tr. D-2864, at 4.
185. Since the alternative shutdown solution meets accepted NRC Staff criteria, there is no deviation from licensing criteria. Wermiel et al., ff.
Tr. D-2864, at 5.
186. The Board finds that the alternative for safe shutdown meets accepted requirements and that there is no deviation from licensing criteria.
Contention 4(1)(1) !
i
.--,,_---._.-n_-.-___.___,_
i 187. Contention 4(1)(1) maintains that the IDVP accepted a deviation from the licensing commitments in that the existence of a large grated ventilation opening in the AFW pump room is inconsistent with a licensing constitment for fire zone separation.
188. The existence and size of a grated ventilation opening in the ceiling of the AFW pump room raised a question as to whether the plant still met a licensing criteria that a fire in one zone not cause unacceptable damage to the equipment necessary for safe shutdown in another zone. Moore et al., ff.
Tr. D-487, at 22; Kubicki et al., ff. Tr. D-2864, at 3; Krechting et al., ff.
Tr. D-2040, at 4-16.
189. A detailed fire hazards analysis demonstrated that the plant was originally designed to meet licensing criteria with the opening. The analysis considered the actual combustible loading, the curbing provided around the opening in the room above the AFW pump room, and the air flow through the opening. Moore et al., ff. Tr. D-487, at 22. Review of postulated credible fires indicated that a fire in one zone would not propagate through the opening to the other zone. Krechting et al., ff. Tr. D-2040, at 4-16.
190. The Staff independently did an evaluation of the entire fire protection program at DCNPP in conjunction with compliance with Appendix R.
Kubicki, Tr.
D-2874. Considering the nature of the actual fire hazard, and that the fire would be of such limited magnitude and extent, the Staff determined that independent of the size of opening, any propagation would not represent a significant threat to adjoining areas. Kubicki, Tr. D-2875.
191. Based on the foregoing the licensing conmiitment that a fire will not propagate from one fire zone to another is satisfied and the connitment that safe shutdown not be hindered is met. Krechting et al., ff. Tr. D-2040, at.
e I
t
4-16; Kubicki, ff. Tr. D-2864, at 3; App. Ex. 110.
192. The Board finds that the fire hazard analyses of the DCP and the Staff establish that there is no deviation from licensing criteria.
Contention 4(1)(2) 193. Contention 4(1)(2) maintains that the IDVP accepted a deviation from 3censingcommitmentsinthatinstallationofafiredamperwithgapswasnot consistent with a licensing commitment for fire zone separation.
194. The IDVP determined that the fire damper was qualified by Underwriter's s
Laboratories and that the gaps were part of the damper design to facilitate metal expansion under actual fire conditions. Kubicki, ff. Tr. D-2864, at 4; Krechting et al., ff. Tr. D-2040, at 4-17; App. Ex.110.
195. The Board finds that there is reasonable assurance that no deviation from licensing criteria exists as a result of the fire damper.
Contention 4(j) 196. Contention 4(j) maintains that the IDVP accepted a deviation from a licensing comitment in that fire protection for the AFW pump room for assuring cable separation was violated since (1) the pimips for the motor-driven AFW pumps and the control circuitry for a flow control valve i
necessary for operation of the turbine-driven AFW pump were located in a single fire zone and (2) the caoles for some AFW circuits were not routed as described in the Supplemental Information for Fire Protection Review (SIFPR) nor were four AFW circuits identified and reviewed as comitted to in the i
SIFPR.
l 197. The IDVP, upon further review, determined that the control circuitry for FCV-95 was not located in the same fire zone as the circuits for the motor-driven AFW pimps; hence, there was no violation of applicable fire
- l 3
i
o I
protection requirements. Moore et al., ff. Tr. D-487, at 23-24; Krechting et al., ff. Tr. D-2040, at 4-18; Kubicki, ff. Tr. D-2864, at 4-5.
198. While the IDVP found that some AFW circuits were not routed as described in the SIFPR or were not assessed in the SIFPR, it determined that the cables had been rerouted subsequent to the issuance of the SIFPR. Furthermore, the IDVP determined that the rerouting, with the exception of FCV-95, met licensing separation requirements. With regard to the FCV-95 circuit, the IDVP determined that the ongoing rerouting of FCV-95 circuitry was completed in accord with the licensing requirements. In these instances, the IDVP field verified that proper routing was maintained in accord with fire protection Krechting et al., ff. Tr. D-2040, at 4-18-19; Moore et al., ff.
requirements.
Tr. D-487, at 24.
199. Thus, there were no deviations from separation requirements and a single Krechting et al.,
fire could not prevent proper operation of the AFW system.
ff. Tr. D-2040, at 4-19; Kubicki, ff. Tr. D-2864, at 5-6.
200. The Board finds that a single fire could not prevent proper operation of the AFW system and that there are no deviations from licensing criteria.
Contention 4(k) 201. Contention 4(k) maintains that the IDVP accepted a deviation from licensing commitments since, contrary to the requirements of PGandE's SIFPR, eacii of the three 4160V cable spreading rooms has a ventilation opening leading up to the 4160V switchgear rooms.
202. The IDVP noted that a ventilation opening existed between each of those Krechting et al.,
4160V cable spreading rooms to the 4160V switchgear rooms.
ff. Tr. D-2040, at 4-20; Kubicki, ff. Tr. D-2864, at 6.
203. However, the IDVP reviewed and accepted PGandE's detailed fire hazards.
I
analysis which demonstrated that the fire zone criterion had been met and that a fire could not credibly propagate from one zone to another zone which would preclude the safe shutdown capability. Krechting et al., ff. Tr. D-2040, at 4-20; Moore et al., ff. Tr. D-487, at 24-25; Kubicki, ff. Tr. D-2864, at 6.
204. Both the IDVP and the Staff concluded that no deviation from licensing criteria exists. Krechting et al., ff. Tr D-2040, at 4-20; Kubicki, ff. Tr.
D-2864, at 6-7.
205. The Board finds that there is reasonable assurance that a fire could not propagate from one zone to another which would preclude safe shutdown capability and that there has been no deviation from licensing criteria.
Contention 4(1) 206. Contention 4(1) maintains that the IDVP has accepted a deviation from a licensing commitment in that jet impingement loads have not been considered in the design and qualification of safety-related piping and equipment inside containment.
207. Applicant's FSAR connits to analyses of jet impingement resulting from high energy pipe breaks inside containment. Moore et al., ff. Tr. D-487, at 25; Knight, ff. Tr. D-2864, at 3.
No formal analysis was required when DCNPP was designed. However, the DCP, in response to the IDVP, conducted an extensive jet impingement analysis considering all hypothetical jets and credible safety-related targets. Moore et al., ff. Tr. D-487, at 26.
208. While the FSAR did not include criteria for a combination of pressure and temperature to establish which lines inside containment should be analyzed, the FSAR did list the specific lines for which protection was to be provided.
l All listed lines were analyzed for the effects of jet impingement. Connell, l
Tr. D-589 D-616..
I
209. Impingement loads were considered as provided in licensing criteria and the IDVP did not accept deviations from those criteria regarding possible jet impingement loads on safety-related piping and equipment inside containment.
Knight et al., ff. Tr. D-2864, at 4; Krechting et al., ff. Tr. D-2040, at 4-22.
210. The Board finds that impingement loads were considered as provided in the licensing criteria and that there have been no deviation from such criteria in the analysis of jet impingement.
Contention 4(q) 211. Contention 4(q) maintains that the IDVP accepted deviations from licensing criteria in that modifications to protect two auxiliary feedwater valves from the effects of MELB were not implemented.
212. In 1979, PGandE indicated by letter to the NRC Staff an intention to add splash shields to two motor-operated auxiliary feedwater valves to assure that water dripping from postulated cracks in moderate energy pipes would not prevent the valves from performing their safety function. Moore et al., ff.
Tr. D-487, at 31.
213. Subsequent analysis showed that the proposed protection was not n&;ssary and none was installed. Moore et al., ff. Tr. D-487, at 31; Moore, Tr. D-515.
214. The DCP provided the IDVP with an analysis which indicated that the two valves which are on the suction supply line to the AFWS pumps from the long term alternate water supply are not needed to mitigate the postulated MELB.
I Wermiel, ff. Tr. D-2864, at 7.
The valves can be operated manually and provide the same level of assurance as automatic alignment. Anderson, Tr.
i D-515. Either valve would be sufficient to provide required cooling water.
Connell, Tr. D-518.
215. Of eight alternative ways to supply cooling water to the auxiliary,
- - ~ - - - -
feedwater system, the two valves are involved with the third preferred method. Connell, Tr. D-516. D-517. It would be 18 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> under normal operation before the third alternative would be necessary, during which time repair of the MELB could be reasonably expected. Connell, Tr. D-519.
216. There is reasonable assurance that the two valves can perform this intended safety function and that there is no deviation from licensing criteria. Krechting et al., ff. Tr. D-2040, at 4-26; Wermiel, ff. Tr. D-2864, at 7; App. Ex. 113.
217. The Board finds that there is reasonable assurance that the two valves can perform their intended safety function and that there is no deviation from licensing criteria.
Contention 4(r) 218. Contention 4(r) maintains that the IDVP accepted deviations from licensing criteria in that four components were accepted where postulated high energy line breaks could cause temperatures in excess of specification for the components.
219. The IDVP independent analysis of high energy line cracks (HELC) identified two separate potential high energy line crack locations of concern, one of which could cause a loss of the motor-driven AFWS pvmps and two AFWS pressure switches, the other which could affect two AFWS level control valves. Wermiel, ff. Tr. D-2864, at 8.
220. PGandE re-evaluated the high energy line break analysis for the line in question. This line was the steam supply line to the turbine driven AFW pump downstream of the flow control valve. Plant procedures specify that the motor driven AFW pumps be used in lieu of the steam turbine driven pumps during startup and shutdown. Because the line is not pressurized during normal plant.
,_,----..._,_-.-..e_
.,.-__..,.7-
_,,._,..,_,..-...-,--,,..,...,,,,,m,
l operating conditions including startup and shutdown, a break in that line need not be postulated. A crack in the line would not cause either the turbine generator or reactor to trip. Consequently, there are no safety consequences if the AFW equipment were exposed to HELC impingement temperatures. Wermiel, ff. Tr. D-2864, at 8; Krechting et al., ff. Tr. D-2040, at 4-28; Moore et al.,
ff. Tr. D-487 at 32; Krechting, Tr. D-2050 - 2051; App. Ex.113 at 2-5 and 4-2.
221. With respect to the postulated pipe break which affects the AFWS level control valves, PGar.dE performed a reanalysis of the blowdown jet temperature using the latest approved ANS 58.2 methodology in lieu of the original FSAR method. The recalculated blowdown jet temperature was below the qualification temperature for the valves. Wermiel, ff. Tr. D-2864, at 8-9; Moore et al.,
ff. Tr. D-487, at 33; Krechting et al., ff. Tr. D-2040, at 4-27-28; App. Ex.
113, at 4-2.
222. The IDVP's acceptance of the DCP's justification for the pipe break concerns was based on adequate engineering justification and is in accordance with licensing criteria. Wermiel, ff. Tr. D-2864, at 9.
223. The Board finds that with regard to AFW equipment there is no deviation from licensing criteria caused by a HELC.
Contention 4(s) 224. Contention 4(s) maintains that the IDVP accepted a deviation from a licensing commitment in that a conduit was identified whose failure due to a HELC could eliminate redundant auxiliary feedwater system flow in violation of the minimum system redundancy commitment of Section 3.6A of the FSA't.
225. The IDVP determined that even if a jet were to impinge on the conduit, the cable would not be damaged by the jet and redundancy of the AFWS would not be affected. Moore et al., ff. Tr. D-487, at 34; Krechting et al., ff. Tr..
O t
D-2040, at 4-29.
226. The DCP demonstrated by analysis that the cables are qualified to temperatures higher than those expected from a postulated jet impingement and that the conduits can withstand pressures greater than the corresponding jet pressures without damage. Moore et al., ff. Tr. D-487, at 34; Wermiel, ff.
Tr. D-2864, at 10; Krechting et al., ff. Tr. D-2040, at 4-29.
227. Furthermore, no system necessary to mitigate the effects of the postulated break or to safely shut down the plant in the event of such a break are affected by these particular AFW cables. Thus, no licensing criteria deviation exists. Moore et al., ff. Tr. D-487, at 34; Krechting et al., ff.
Tr. D-2040, at 4-29.
228. The Board finds that there are no deviations from licensing criteria as respects the HELC analysis of the auxiliary feedwater system.
Contention 4(t) 229. Contention 4(t) maintains that the IDVP accepted a deviation from licensing criteria in that the nameplate rating for circuit breakers on certain 4160V buses was below operational requirements for adequate short circuit interrupting capacity.
230. The DCP furnished test data and a 1983 letter from the manufacturer which indicates that the circuit breakers in question would perform above rating and l
within the required maximum short circuit current. Moore et al., ff. Tr.
l D-487, at 35; Krechting et al., ff. Tr. D-2040, at 4-30; Knox, ff. Tr. D-2864, at 8; Krechting, Tr. D-2054 - 2055; App. Ex. 116.
231. There is reasonable assurance that the circuit breakers can perform their l
)
intended safety function and no deviation from licensing criteria occurred.
Krechting et al., ff. Tr. D-2040, at 4-30. l 1
i l
232. The Board finds that there is no deviation from licensing criteria as the breakers in question can perform their intended safety function..
Contention 4(u) 233. Contention 4(u) maintains that the IDVP accepted a deviation from a licensing commitment in that circuit separation and single failure deficiencies were identified in reviews of the AFW and CRYP systems as well as other safety-related systems.
234. The IDVP identified certain electrical circuits (Class IE) in the AFW and CRYP systems enclosed in panels and termination boxes that were not separated in accordance with the methods listed in FSAR Section 8.3.3.
Additionally, an electrical component in the CRVP was identified that did not meet the single failure criterion. Moore et al., ff. Tr. D-487, at 35-36; Krechting et al.,
ff. Tr. D-2040, at 4-31-32.
235. In response to these IDVP items, PGandE undertook a complete review of all PGandE-designed safety-related systems to confirm that all safety-related electrical circuits met single failure and separation criteria. Where necessary, modifications were made. Moore et al., ff. Tr. D-487, at 36.
236. After completion of the PGandE review, the IDVP selected samples to verify the DCP review. This review included a review of system drawings as well as field inspections to determine whether the installation met separation criteria. In addition, the IDVP reviewed the DCP's single failure analysis.
This included a review of drawings and analyses to verify that all mutuc11y redundant circuits connected to the same device had been identified. The IDVP did not find any case where a single failure could adversely affect the operation of a sample system and no modifications were required. From this review, the IDVP concluded that the circuit separation and single failure l 1
t E
requirements had been met. Hence, no deviation from licensing criteria exists. Krechting et al., ff. Tr. D-2040, at 31-33; Moore et al., ff. Tr.
D-487, at 36; Knox, ff. Tr. D-2864, at 3-4.
237. The Board finds that there is no deviation from separation and single failure licensing criteria in the AFW and CRYP systems.
Contention 5 238. Contention 5 maintains that the verification program has not verified that DCNPP Units 1 and 2 "as-built" conform to the design drawings and analyses.
239. Testimony concerning this contention was presented on behalf of PGandE by its Panel No. 1 (Anderson et al., ff. Tr. D-224, at 31, 32), by the NRC Staff (Morrill, ff. Tr. D-2906, at 1-4) and by the IDVP (Cooper et al., ff. Tr.
D-1459, at 5-1-4).
240. In perfonning its design verification the ITP performed field walkdowns and verifications to ensure that the design documents of record reflected the actual physical installation. Any deviations identified from these field walkdowns were incorporated into the design documents of record. These documents were then used as input for the design review. Anderson et al., ff.
Tr. D-224, at 31; Moore, Tr. D-348-64; D-461-65.
241. With regard to Unit 1 IDVP did independent field verification as part of its seismic and non-seismic verification programs including field verification of all modifications performed by the DCP to resolve EDIs (errors or open items). The specific IDVP field verifications are described in detail in various ITRs. As part of the audit of the implementation of the CAP, RFR audited the procedures for engineering review of field design changes and related procedures for incorporating the changes into final design drawings. ',
e
e Anderson et al., ff. Tr. D-224, at 31-32; Cooper et al., ff. Tr. D-1459, at 5-1-4; App. Ex.106,108,111-120,133,140-154; Gov. Ex. 48, 49.
R. F. Reedy examined all the E0I's showing as-built discrepancies and found none that had any QA implications. Reedy, Tr. D-1640-643.
242. All modification work performed by the DCP has strictly conformed to PGandE Engineering Department Procedure 3.60N for Unit 1 and 3.7 for Unit 2, which provide for engineering review of construction results and revision of design documents to reflect as-built conditions. The Engineering Department also issues guidelines and tolerances to PGandE's General Constuction and Nuclear Power Generation Departments defining documents which must be revised to reflect as-built conditions. Anderson et al., ff. Tr. D-224, at 31, 32; App. Ex. 161; Staff Ex. 49-53.
243. The Staff conducted inspections throughout the verification program to verify the adequacy of the procedures of the IDVP and to verify their implementation. Morrill, ff. Tr. D-2904, at 3-4; Staff Ex. 49-53, 244. Hr. Hubbard, on behalf of the Governor, testified that this contention was really directed toward the difference between the "as-built" condition and the support design documents or "as-analyzed" condition. Hubbard, Tr.
D-2155-56. On cross examination, however, Mr. Hubbard agreed that, where conservative assumptions were used in analysis, as was the case with examples i
he had used, the as-built drawings could not possibly reflect the as-analyzed condition. Hubbard, Tr. D-2157.
245. Mr. Hubbard also conceded on cross examination that, in general, the as-built drawings do reflect the construction at DCNPP. Hubbard, Tr. D-2157.
He did not testify as to any situations where the as-built drawings did not reflect the actual construction.
I '
e t
=
246. The Board finds that the verification program did verify that DCNPP Units 1 and 2 "as-built" conform to the design drawings and analyses consistent with licensing criteria.
I Contention 6 247. Contention 6 maintains that the verification program failed to verify that the design of safety-related equipment supplied by Westinghouse to PGandE met licensing criteria.
248. PGandE presented the testimony of Edward J. Kreh, Jr., a consultant to Westinghouse Corporation and a prior Westinghouse Product Assurance Manager; Robert A. Wiesemann, Manager of Regulatory and Legislative Affairs in the Nuclear Safety Department of the Nuclear Technology Division of Westinghouse; and John B. Hoch, PGandE Project Manager at DCNPP. The IDVP presented testimony of Dr. Wil11am E. Cooper, Dr. Robert L. Cloud, John E. Krechting, and Roger F. Reedy. The Governor presented testimony of Richard B. Hubbard, a former QA manager for an instrument and controls section of General Electric l
Company.
249. The NRC Order and Staff Letter of November 19, 1981 require review of t
" service contractors utilized in the design process for safety-related structures, systems, and components." Neither required verification of the design of safety-related equipment supplied by vendors. App. Ex. 86, 87; Kreh i
et al., ff. Tr. D-1088, at 1-3.
250. Westinghouse was not considered a service-related design contractor under the IDVP program which was submitted to and approved by the Comission. App.
Ex. 101, 156, 157, 158, 159. Kreh et al., ff. Tr. D-1088, at 1-3; Cooper, et al., ff. Tr. D-1459, at 6-2.
251. The equipment furnished by Westinghouse was environmentally qualified to t
standard requirements. The standards were developed as a result of detailed NRC review and constitute an equipment qualification program which Westinghouse submitted to the NRC. Wiesemann, Tr. D-1146.
252. Mr. Wiesemann, who is secretary of the Westinghouse Safety Review Cossnittee which would review any failure of equipment to meet environmental requirements, was not aware of any equipment supplied in the NSSS for DCNPP which would not meet current environmental specifications. Wiesemann, Tr.
D-1147.
253. The verification program did not reanalyze all safety-related equipment furnished PGandE by Westinghouse. Rather, whenever findings of the verification program altered the input to specific pieces of safety-related equipment, that equipment was requalified by Westinghouse. Hubbard, ff. Tr.
D-2084, at 20. All the equipment that was affected was evaluated against the new requirements to ensure that the equipment would meet safety-related functional requirements. Wiesemann, Tr. D-1111.
254. Most of the design work. on the equipment supplied by Westinghouse for Units 1 and 2 was completed prior to 1974. Kreh, Tr. D-1089. The design of l
the NSSS system is a generic, standardized design. Kreh, Tr. D-1091-92; Wiesemann, Tr. D-Il03, D-Il05. The Westinghouse design process, procedures, l
and controls for the NSSS system have been audited extensively over the years by the NRC, utilities, architect / engineers, and ASME. Kreh, et al., ff. Tr.
D-1088, at 3; Kreh Tr. D-1091, D-lll6; Wiesemann, Tr. D-1093, D-lll5. The NRC audits that took place during the reanalysis by Westinghouse did include work which involved DCNPP. Wiesemann, Tr. D-1115.
255. Prior to receiving fomal approval of the Westinghouse QA program after adoption of Appendix B of 10 CFR Part 50 in 1973-1974 (Kreh, Tr. D-1089, \\
I-
_... _.., _ ~,.. _, _
D-Il49), the design work of Westinghouse was carried out according to the requirements of MIL Q 9858, which was the quality assurance specification used by the nuclear Navy program and the precursor to Section 3 of Appendix B of 10 CFR Part 50. Kreh, Tr. D-1151.
256. All design work of equipment furnished by Westinghouse to PGandE was accomplished under an approved QA program that was carried out in accordance with either Appendix B to 10 CFR Part 50 or its predecessor, MIL Q 9858.
257. On cross examination, the Governor's QA witness, Mr. Hubbard, conceded that if Westinghouse had an acceptable QA program or had original compliance with 10 CFR Part 50, he would have less concern about the Westinghouse furnished systems than he would have if their design were verified by a verification program such as the IDVP. Hubbard, Tr. D-2145-46.
258. The IDVP did verify that the interf ace between PGandE and Westinghouse for the NSSS system included appropriate controls for the transfer of design information and that Westinghouse, with one exception, properly used the applicable information. Cooper et al., ff. Tr. D-1459, at 6-2; Kreh. Tr.
D-ll41; App. Ex. 90 (Sec. 4.1.3), 103, 114, 134.
In that instance, Westinghouse requalified the equipment that was affected. Hubbard, ff. Tr.
D-2084, at 20; Wiesemann, Tr. D-llli.
259. Because of the existence of an approved and accepted QA program during all phases of design, and the use of a standardized design which has previously been accepted for such equipment, there is reasonable assurance that the safety-related equipment furnished by Westinghouse meets licensing criteria.
Contention 7 260. Contention 7 maintains that the verification program failed to identify
- the root causes for the failures in the PGandE design quality assurance program and failed to determine if such failures raise generic concerns.
261. PGandE, the IDVP, and the NRC Staff offered testimony on Contention 7.
Testifying for PGandE were Steven M. Skidmore, Manager of Quality Assurance for PGandE; Thomas G. De Uriarte, Director of Program Management for PGandE Quality Assurance; Charles W. Dick, Project QA Manager for the Project Completion Team on the DCP; and Michael J. Jacobson, Quality Assurance l
Engineer for the Project Completion Team on the DCP. On rebuttal, PGandE presented Gary H. Moore, DCP Engineer for Unit 1; Leigh A. Gouveia, a consulting engineer with Project Assistance Corporation (PAC), who was involved in the review of the QA Manual conducted by PAC, and William A.
Stokes of Shalako Energy Services, who was the Project Engineer for EDS Nuclear during the EDS review of certain quality control manuals of the various departments. Testifying for the IDVP were Dr. William Cooper and Roger F. Reedy. Testifying for the NRC Staff were James P. Knight, Hartmut E.
Schierling, Walter P. Haass, Assistant to the Director, Office of Inspection and Enforcement, and Philip J. Morrill, Reactor Inspector, Division of Resident, Reactor Projects, and Engineering Programs, Region V.
Testifying for the Governor was Richard B. Hubbard. The JI did not present any testimony on Contention 7.
262. Pursuant to the Commission Order and Staff Letter, the verification program, including PGandE's internal efforts and the IDVP's independent efforts, identified the basic causes for failures in the Diablo Canyon design quality assurance program and for the design errors identified during the verification program. De Uriarte et al., ff. Tr. D-847, at 1; Reedy et al.,
ff. Tr. D-1459, at 7-1; Staff Ex. 36, at C.5-5 to 5-7. l L
t
263. In November, 1981 PGandE's QA Department initiated a series of "Took back" reviews of its own quality assurance program and the quality assurance programs of its service-related contractors. At the same time the IDVP, through R. F. Reedy, initiated an independent review of the QA programs of PGandE and its service-related contractors. Both sought to determine the basic causes of the deficiencies in the QA program. Current interpretations and applications of 10 CFR Part 50, Appendix B, were used in these reviews.
De Uriarte et al., ff. Tr. D-847, at 1, 2.
264. The "look-back" reviews made the initial identification of the basic causes of the failures in the design quality assurance program. De Uriarte et al., ff. Tr. D-847, at 2-3.
265. For the service-related contractors, the basic causes identified were i
failures to require quality assurance controls prior to mid-1978 and failures of PGandE in the control of transmitted information, record disposition, and interface control. For PGandE's design engineering effort, the basic causes that resulted in deficiencies of design quality assurance were inadequate control of changes in FSAR descriptions, inadequate control of documents, and inadequate documentation of design inputs. De Uriarte et al., ff. Tr. D-847, at 3.
The IDVP's determinations of basic causes are reported in the IDVP Final Report, Section 6.3.
Reedy et al., ff. Tr. D-1459, at 7-1; App. Ex. 90.
l 266. The two basic causes identified by the IDVP were control of design interfaces and documentation and interpretation of design. These were similar to the basic causes identified by PGandE. De Uriarte et al., ff. Tr. D-847, at 6; Reedy et al., ff. Tr. D-1459, at 7-1.
The consequences of these basic l
l causes have been identified and corrected. Reedy et al., ff. Tr. D-1459, at 7-2..
t
.. _ _ _, _ -, - - -..,. - _. - -. _., ~.. _ _ _. -.. ~. _.. - _ _ _.. _. _.. - _
267. The IDVP and the ITP identified other basic causes related to the design engineering process in Section 1.8 of PGandE's Phase I Final Report, Section 3.0 of PGandE's Phase II Final Report, and Section 6.3 of the IDVP's Final Report. These basic causes include the evolution of seismic technology and criteria and the extended length of time associated with the interpretation of design requirements. These basic causes were not strictly related to design quality assurance, but were factors related to the design engineering process which influenced the design quality assurance program. De Uriarte et al., ff.
Tr. D-847, at 5; Reedy et al., ff. Tr. D-1459, at 7-1, 7-2; App. Ex. 90, 91, 92.
268. The Staff concurs with PGandE and the IDVP that the basic causes identified in PGandE's Phase I and Phase II Final Reports and the IDVP Final Report were the major contributing factors to the design errors found during the course of the verification effort. Knight et al., ff. Tr. D-2906, at 4; Staff Ex. 34, at C.5-6.
269. Mr. Hubbard alleged that neither the IDVP nor the ITP correlated basic causes to identified errors. Hubbard, ff. Tr. D-2084, at 26-33. On cross-examination Mr. Hubbard was shown two sample non-conformance reports (which were issued for each and every error) (Gov. Ex. 43, 44) and admitted that the non-conformance reports identified the cause of the non-conformance (error). When asked if this showed that the ITP looked at the cause of errors he replied he would have to look at the non-conformances, which he had not done, to decide whether the cause was looked ht in-depth. Hubbard, Tr.
D-2163-164.
270. Counsel for the Joint Intervenors attempted to establish by cross examination of Mr. De Uriarte that the single underlying csuse of the basi:.
t
o causes identified for the DQA failures at Diablo Canyon was PGandE's management's lack of commitment to QA. Mr. De Uriarte disagreed with the conclusion sought to be reached by counsel. Mr. De Uriarte explained that the problems found existed in the interface between seismic design consultants and PGandE engineer design groups. There was not a similar problem found with internal design interfaces, nor with non-seismic design groups.
If lack of L
management control or management attention was a basic cause, the discrepancies or deficiencies would have been found in all areas. De Uriarte, Tr. D-1013-14.
271. To correct the identified basic causes, PGandE's Quality Assurance Department reviewed PGandE's QA program to assure that each of the basic causes were properly addressed through existing or strengthened controls in the program. The ITP implemented additional corrective measures for all of the identified basic causes. De Uriarte et al., ff. Tr. D-847, at 3.
272. Control of design interfaces is addressed in the current PGandE and DCP QA Programs. Interfaces with design consultants are specifically addressed in Engineering Manual Procedures 3.8 and 4.6., respectively. Among the controls included in these procedures are:
(1) identification of interfaces between PGandE and the consultant by name, title, and responsibility for decision-making and for providing and reviewing information; (2) documentation of the transmittal of design information between PGandE and the consultant; (3) verification that information submitted to the consultant for use has been reviewed and approved; (4) review and approval of consultants' design documents to provide assurance that the documents meet Diablo Canyon requirements. De Uriarte et al., ff. Tr. D-847, at 4.
273. The DCP has initiated a number of corrective actions to address the
- specific issue of inadequate control of changes in the FSAR descriptions.
These actions include initiating a nonconformance report to resolve the issue, updating the FSAR, and developing a series of Design Criteria Memoranda to define criteria and inputs to be used for DCP activities. De Uriarte et al.,
ff. Tr. D-897, at 4.
274. Under current practice an engineer is not allowed to use the FSAR as the sole source of desi:n information without verification of the validity of the information contained in the FSAR. The FSAR is verified by comparing it to other documents in the file, the drawings, the as-bufft condition of the plant, and talking to other disciplines. De Uriarte, Tr. D-868-74; Skidmore, Tr. D-874-77.
275. Inadequate control of documents was addressed through various actions, including reviews of calculations and calculation indices, updating when necessary, reviewing file copies of specifications, updating wherever necessary, and accounting for all annotations to the specifications.
De Uriarte et al., ff. Tr. D-847, at 5.
276. Because of the known concerns about control of design interfaces, the IDVP paid particular attention to review of the flow of information among PGandE and its contractors and within PGandE. This concern was addressed by both QA and design process verification efforts.
In the case of the CAP, the
" design office verification" procedure was developed specifically by the IDVP to assure that the QA procedures and their implementation were adequate to the f
specific design aspects. Reedy et al., ff. Tr. D-1459, at 7-2.
277. With respect to the docrentation and interpretation of design, the IDVP was sensitive to the possibility that any identified deficiencies in these areas could have-generic impacts. Throughout the verification efforts, l H
+
criteria and methodology were carefully defined and documented so as to assure that generic concerns associated with problems in documentation and interpretation of design were identified and resolved. Reedy et al., ff. Tr.
D-1459, at 7-3.
278. The IDVP identified generic concerns associated with design errors identified during its program. Reedy et al., ff. Tr. D-1459, at 7-3.
For each E0I file, a review was performed to determine if it was a significant condition adverse to quality or if it represented a generic concern.
Jacobson, Tr. D-987.
279. On cross examination, Mr. Hubbard agreed that both the IDVP and ITP had looked for generic concerns. Hubbard, Tr. D-2160.
280. The generic concerns related to the root causes have been resolved.
Knight et al., ff. Tr. D-2906, at 5.
281. The IDVP reviewed each E0I resulting from the verification effort for generic concerns and resolved all such generic concerns as part of the verification effort. Reedy et al., ff. Tr. D-1459, at 7-3.
The IDVP also reviewed each E01 for QA discrepancies and found no violations of any significance in the program. Reedy, Tr. D-1642-43.
282. The identification of generic concerns was an important part of the l
IDVP. A generic concern was a concern which could impact design acceptability beyond the immediate system, structure, or component (SSC) for which the concern was initially identified. The IDVP conclusion that a generic concern i
existed was identified in an ITR (e.g., ITRs 1, 34). When generic concerns were identified, the steps that were taken included, as appropriate, the l
evaluation of the effect of the generic concern on other safety-related systems, structures, or components. Cooper et al., ff. Tr. D-1459, at 1/2-26.
- \\
t
283. The verification effort for the initial and additional samples was performed under the assumption that design QA was deficient, and the basic causes identified turned out to be design QA related. Since the IDVP assumed inadequate design QA in developing the IDVP programs, the identification of root causes which were actually associated with QA deficiencies was no surprise to the IDVP and did not result in a requirement for additional expansion. Reedy et al., ff. Tr. D-1459, at 7-3.
284. In resolving every E01, the IDVP not only disposed of the specific concern raised by the E0I, but determined whether there existed a generic In some cases, it was necessary to examine the cause of the E0I as concern.
part of the specific concern which had to be remedied; while in other instances the cause of the E0I ted to its being designated as a generic Although the IDVP documentation did not necessarily include a concern.
specific label for "cause", in the case of each E01, the IDVP determined whether the factors relating to the cause of the E01 required that any additional action be taken. Reedy et al., ff. Tr. D-1459, at 7-4.
285. In the opinion of the IDVP, the separate identification of the basic cause for each and every E01 was unnecessary. In the IDVP's view, assessment of basic cause is more meaningful when it can encompass a review of all the deficiencies identified in an entire program, rather than by focusing on isolated items. This is what the IDVP did as reported in Section 6.3 of the Final Report. Reedy et al., ff. Tr. D-1459, at 7-4.
286. The QA program is a management tool to provide assurance that the design process is being effectively carried out. The program relies on procedures, training of personnel, and audits to determine that procedures are effect.ve and are being properly implemented. The QA program provides assurance that.
\\
the necessary reviews are being performed per procedures, but it cannot be expected to detect each and every technical error on the part of the engineers f
I who performed these reviews. De Uriarte et al., ff. Tr. D-847, at 7, 8.
On cross examination, Mr. Hubbard admitted he would not expect a QA program to l
result in zero defects. Hubbard, Tr. D-2130.
287. The causes of the QA deficiencies have been identified for each design error uncovered. They are both QA-related and engineering-related and include I
the originating error and the secondary errors. Since the causes were found t
to raise generic concerns, the causes of secondary errors were addressed in the overall resolution of the originating error. The extensiveness of the design verification program was partly because the nature of the design quality assurance errors raised such generic concerns. De Uriarte et al., ff.
Tr. D-847, at 8.
288. The initial activities of the IDVP and the subsequent reverification efforts by the ITP cover all significant design functions where information transfer deficiencies were identified. In other instances, the IDVP found the engineering design process to be working well despite shortcomings in either the contractors' or PGandE design QA programs and closed out their findings on the basis of the quality of the normal engineering design controls in effect.
The vast undertaking represented by the total reverification effort at DCNPP and, in particular, the creation of the DCP represents, in the Staff's view, ample evidence that the highest levels of management at PGandE have recognized and implemented the rigorous design quality control programs necessary for nuclear plant design. Knight et al., ff. Tr. D-2906, at 4-6.
289. Based on the foregoing, the Board finds that the verification program s
fully complied with the Conmiission Order by identifying the root causes for.
b
- ~. -.
the failures in the PGandE Design Quality Assurance Program and determining if such failures raised generic concerns.
Contention 8 290. Contention 8 maintains that the ITP failed to develop and implement in a timely manner a design quality assurance program in accordance with 10 CFR Part 50 Appendix B to assure the quality of the recent design modifications to the DCNPP facility and further questions whether the IDVP assured that the work which was done met licensing criteria.
291. PGandE, the IDVP, and the NRC Staff offered testimony on Contention 8.
Testifying for PGandE were Steven M. Skidmore. Thomas G. Da Uriarte, Charles W. Dick, and Michael J. Jacobson. Testifying for the IDVP were Dr. William Cooper and Roger F. Reedy. Testifying for the NRC Staff were James P. Knight, Hartmut E. Schierling, Walter P. Haass, and Philip J. Morrill. Testifying for the Governor was Richard B. Hubbard. The JI did not present any testimony on Contention 8.
292. Between November 1981 and August 1982 the developing DCP was conducted under the PGandE QA Program. The acceptability of that program was confirmed in NRC Inspection Reports 50-275/81-29,50-323/81-18 in November 1981.
De Uriarte et al., ff. Tr. D-847, at 9, 10. From August 1982 to the present, all work was performed under the DCP QA Program. De Uriarte et al., ff. Tr.
D-847, at 10. During the entire review, Project Engineering worked under and f
followed procedures as contained in the PGandE Engineering Manual. Moore, Tr.
D-3161.
293. Between 20% to 25% of the overall verification effort was performed between November 1981 and August 1982. Moore, Tr. D-3156-157. This consisted of an issue-by-issue review to a point where it was decided to redo all of the.
O
i design in a particular area. The reviews conducted after those decisions consisted of a complete redoing of the review without incorporation of the previous review work. The reviews started off with a " clean piece of paper."
Moore. Tr. D-3157.
294. The decisions to redo the reviews were made as follows: Piping, July, 1982; Auxiliary Building, June, 1982; Fuel Handling Building, May, 1982; Turbine Building, October 1982; Intake Structure, June, 1982; the final Annulus Review, January, 1983; Raceway and HVAC Supports, July, 1982. Moore, Tr. D-3158. All of the Phase II reviews, including the HVAC technical reviews, the Electrical reviews and Mechanical reviews August, 1982. Moore, Tr. D-3159.
295. Between 90% and 100% of the design work which resulted in plant modifications was performed under the DCP QA Program after August, 1982.
Moore, Tr. D-3160; De Uriarte et al., ff. Tr. D-847, at 11.
296. On cross examination, Governor's witness Hubbard admitted that a QA program for safety-related items should be designed to assure that all critical errors will be detected. When asked whether the term " critical", as used in his testimony, was used in a safety sense, he replied in the negative. Mr. Hubbard explained that cri,tical meant an error Class A, B, A/B, or C in a QA sense. Mr. Hubbard also admitted that a good QA program is designed to detect major critical errors and that the PGandE QA program in effect from November 1981 to August 1982 was designed to catch major critical errors. Hubbard, Tr. D-2140, D-2141, D-2142, D-2143.
297. With the formation of the Diablo Canyon Project, a new organizational structure was prepared. The QA Program consisted of the NRC approved Bechtel Topical Report, BQ-TOP-1, with modifications as needed to fit Diablo Canyon. t
De Uriarte et al., ff. Tr. D-847, at 10.
298. JI attempted to establish by cross examination that the Bechtel QA program was adopted by the DCP rather than the PGandE QA program because of deficiencies existing in the PGandE QA program used from November of 1981.
Mr. Dick disagreed and explained that it was the judgment of management that the best course to pursue would be to adopt the Bechtel program. Both programs had been accepted by the NRC; both were based on proven concepts.
However, it was felt that the Bechtel program would have better acceptance because of the environment at the time with respect to the suspended license and the fact that Diablo Canyon's QA program was " suspect." At no time was the conclusion ever reached nor was it implicit in the selection, that the PGandE QA program was inadequate. Dick Tr. D-1016-17.
299. The Staff concluded that the design activities during the time the IDVP was conducted were performed in accordance with the quality assurance program that was approved by the Steff in 1982. Staff Ex. 34, at C.5-4.
300. The DCP/PGandE interface is that of Architect Engineer / Licensee. PGandE reviewed the DCP QA Program for compliance with PGandE's licensing comaitments and QA Program requirements. PGandE's QA Department has performed and continues to perform audits of the DCP quality-related activities. De Uriarte l
et al., ff. Tr. D-847, at 11.
i l
301. As part of the supplier qualification program, all of the other ITP contractors responsible for safety-related design activities and the IDVP and its contractors have been approved by the PGandE QA Department. De Uriarte et i
l al., ff. Tr. D-847, at 12.
302. Project QA personnel have participated in a comprehensive training l
program offering an orientation to the fundamental concepts and principles of.
t
quality (including application of 10 CFR 50 Appendix B, ANSI M45.2, ANSI daughter standards, codes, etc.), functional training in quality policy and implementing documents, communications training, and auditor training.
i De Uriarte et al., ff. Tr. D-847, at 17, 18.
303. The DCP has taken appropriate actions to assure continuing implementation of the quality assurance program. These actions include audits performed by PGandE Quality Assurance, Project Quality Assurance, and Bechtel San Francisco i
Power Division Quality Assurance. Audit results have been reported to management and appropriate remedial, investigative, and corrective actions have been formulated and pursued where findings have been identified. PGandE, Project, and Bechtel SFPD management groups have kept informed about the status of the DCP QA Program and provided direction as appropriate.
De Uriarte et al., ff. Tr. D-847, at 18.
304. The audit system which evaluated DCP activities included management level audits, external audits, and Project audits. Management level audits include audits by PGandE QA and Bechtel SFPD QA. Both organizations are independent of the DCP and provide a management overview of the adequacy of the DCP QA Program. External audits include audits by the IDVP and the NRC. Project audits include audits by the Project QA Group and also PGandE QA. The audit system of DCP activities has been planned, scheduled, and executed. The audit system meets the requirements of ANSI N45.2.12-1977. De Uriarte et al., ff.
Tr. D-847, at 18, 19. These auditors found that an overall effective QA program was being implemented by the DCP. De Uriarte et al., ff. Tr. D-847, at 22-23.
305. All audits are performed by quality groups that are independent of those performing the productica work. Audit team leaders and auditors are certified k
t
in accordance with ANSI N45.2.23-1978. Audit results a:e documented and reported to management. De Uriarte et al., ff. Tr. D-847, at 19.
306. The overall results of audits performed reflect that the DCP QA Program has been effectively implemented. No audit findings have been identified that resulted in the issuance of PGandE Nonconformance Reports or Bechtel Management Corrective Action Reports (MCAR). These documents are used for significant conditions adverse to quality or for deficiencies determined to be potentially reportable to the NRC. Similarly, no NRC Notices of Violation have been issued with respect to the DCP design QA Program. As documented in ITR-41, no IDVP Error or Open Item Reports were issued as a result of the IDVP audit of the Corrective Action Program. De Uriarte et al., ff. Tr. D-847, at 20.
307. After numerous audits by several organizations, no MCARs were issued and there was less than 100 findings for a project which has as many as 1200 technical people performing design work over a period of some 18 months.
De Uriarte et al., ff. Tr. D-847, at 21.
308. The DCP QA Program reviewed all E01s issued by the IDVP to determine if the conditions reported represented a p,roblem with respect to implementation of the QA Program and to determine if any adverse quality trends were apparent. The vast majority of E0Is issued concerned past, i.e., prior to the initiation of the Corrective Action Program, design work. For EDIs relating to the CAP, Project QA reviewed the E0I, the Project response to the E0I, and the IDVP close-out. If necessary, Project QA performed reviews of documentation supporting the Project response to ascertain the nature of the problem. De Uriarte et al., ff. Tr. D-847, at 23, 24.
309. All elements of the QA Program have been applied to the Unit 2 design.
I t
efforts. A Project audit of the Unit 2 Internal Review of Unit 1 verification program results found that the program was being effectively implemented. No deficiencies were noted in the audit. De Uriarte et al., ff. Tr. D-847, at 24.
310. The IDVP, through R. F. Reedy, performed an in-depth audit of the QA 4
Program applied to the CAP. The results are contained in ITR-41. De Uriarte et al., ff. Tr. D-847, at 25. The IDVP has reviewe'd the ITP's implementation of corrective and preventative action programs and has concluded that the ITP's actions in this area, when measured against applicable licensing requirements,have been acceptable. De Uriarte et al., ff. Tr. D-847, at 25.
311. Both the Phase I and Phase II IDVP program plans required the IDVP to verify any PGandE corrective action resulting from the design verification performed by the IDVP. The CAP was considered to involve such corrective action. Reedy et al., ff. Tr. D-1459, at 8-1.
312. R. F. Reedy conducted an initial audit of that portion of the ITP work performed under the CAP from November ll, 1982 through December 7, 1982 and a followup audit was completed on March 17, 1983. Reedy et al., ff. Tr. D-1459, at 8-1.
313. The initial audit showed that a number of design and QA activities were incomplete at the time of the audit or not yet fully documented. As a result, they concluded that an insufficient amount of completed documentation was available for them to determine the adequacy of the DCP QA Program implementation. Twenty-four (24) conditions or areas that were fcund to be incomplete were identified in the first audit for subsequent followup by the IDVP. Reedy et al., ff. Tr. D-1459, at 8-3.
314. A followup audit was cc:apleted on March 17, 1983 since design activities performed by the CAP had progressed to a point where a sufficient volume of 1 vn,
,,w,,
-e-,
a-,-w,,-.-.,--
,.--...,,--..ge-
-se--,
-w,,,,.
,,,---,-.e---
documentation had been completed which could then be reviewed to assess the adequacy of the overall implementation of the QA Program. Reedy et al., ff.
Tr. D-1459, at 8-4.
315. Even though the IDVP found some aspects of the QA Program were not yet fully implemented at the time of the initial audit, the IDVP concluded that the QA Program had been implemented in a timely manner. Reedy et al., ff. Tr.
D-1459, at 8-3; Jacobson Tr. D-1021.
316. Counsel for the JI questioned Mr. Reedy on whether the DCP QA Program was fully implemented in December, 1982 when Reedy completed the first part of their audit. Mr. Reedy stated that, as viewed by a reasonable person with sufficient QA expertise, the program was fully implemented at that time.
Reedy, Tr. D-1699.
317. The followup audit included a specific review of the 24 conditions noted during the first audit. For each of these conditions the previous documentation was reexamined to determine the adequacy of correction or completion action, and documents not available initially were requested and examined to determine compliance with QA Program commitments. Finally, responsible DCP personnel were questioned to determine whether they understood the requirements of the QA Program. Reedy et al., ff. Tr. D-1459, at 8-4.
318. The results of the followup audit confirmed that open or unresolved items from the previous audit had been satisfactorily resolved and no items of non-compliance were identified. Reedy et al., ff. Tr. D-1459, at 8-4, 8-5.
319. The IDVP in ITR-8 required that interface control and project indoctrination be verified for each subject where design process verification was required. RFR performed these aspects of a design office verification (DOV) between December 20, 1982 and March 11, 1983. Audit teams verified.
h
'I t
.. ~ _ _ _ -. _ _,. -..,.. - -. _
technical interface controls and project indoctrination in order to assure that the Hosgri and non-Hosgri seismic design inputs were correctly translated into applicable design documents and across design interfaces using the most recent inputs. The audit was performed by tracking seismic inputs from the ground acceleration values to each seismic Category I structure and to the building floor spectra applicable to the piping design documents sampled. The 4
D0V also verified that computer programs used in the seismic design analyses had been verified by the DCP. As described in ITR-41, a D0V was performed in the areas of mechanical equipment, the auxiliary building, the intake structure, large bore piping and supports, instrument tubing and supports, the fuel handling building, the turbine building, the HVAC system, electrical equipment and instrumentation, small bore piping and supports, electrical raceway supports, and the containment structure. Each DOV was conducted by selected professionals experienced in design control and qualified to ANSI N.45.2.23. The audit team used a checklist derived from ITR-8 and the DCP QA Program procedures applicable to the control of the design interfaces, training, and the verification of computer programs. Reedy et al., ff. Tr.
0-1459, at 8-5, 8-6.
i 320. The results of the DOVs showed that control of internal and external interfaces was adequate to assure the use of correct seismic inputs and the correct translation of seismic inputs into corresponding design documents.
The auditors also determined that design personnel who were utilizing seismic information were aware of the applicable QA Program controls. Further, they determined that computer programs used by the DCP were appropriately l
verified. Reedy et al., ff. Tr. D-1459, at 8-6.
321. As a result of the CAP audits and the DOVs, the IDVP concluded that the I.
.t l.
._,,.n
DCP QA Program was effectively implemented. The IDVP's conclusions are reported in ITR-41 and in the IDVP Final Report, Section 4.2.
Reedy et cl.,
ff. Tr. D-1459, at 8-7.
322. Additionally, the IDVP conducted a detailed engineering peer review on a sampling basis of the engineering work of the ITP as applied to the CAP apart from its verification of the QA Program. Reedy et al., ff. Tr. D-1459, at 8-7.
323. As a result of the peer review, the IDVP has determined that the ITP has been effective in resolving earlier E0Is and in reviewing the seismic design of the DONPP-1. Reedy et al., ff. Tr. D-1459, at 8-7.
324. The IDVP verified the Licensee's corrective and preventative action programs implemented by the ITP are sufficient to assure that the deficiencies identified by the IDVP now meet licensing criteria. Morrill, ff. Tr. D-2906, i
at 6.
325. The ITP developed and implemented in a timely manner a QA program in accordance with 10 CFR Part 50, Appendix B.
De Uriarte et al., ff. Tr. D-847, at 9; Reedy et al., ff. Tr. D-1459, at 8-9; Morrill, ff. Tr. D-2906, at 4; Haass, ff. Tr. D-2906, at 3, 4.
326. Based on the foregoing, this Board finds that the ITP implemented its design QA program in accordance with 10 CFR Part 50, Appendix B in a timely fashion to assure the recent design modifications to the DCNPP facility. This Board further finds that the IDVP has ensured that the Corrective and l
Preventative Action Programs implemented by the ITP are sufficient to assure i
that the DCNPP facilities meet licensing criteria.
l l
Contention 9 327. Contention 9 maintains that PGandE has failed to provide adequate assurance of component cooling water system (CCWS) heat removal capacity and l i
that a technical specification limitation does not provide an equivalent level of safety to compliance with GDC-44.
328. Testimony on this contention was presented by PGandE Panel 3 (Testimony ff. Tr. D-487, at 37) and the NRC Staff. Wermiel et al., ff. Tr. D-2864, at 1-4.
Neither the Governor nor JI provided any testimony but did conduct cross examination of the PGandE and Staff witnesses.
329. PGandE has proposed to incorporate a specification in the plant technical specifications which requires periodic monitoring of ocean water temperature.
When the temperature approaches the maximum allowable limit of 640F, the normally isolated second CCWS heat exchanger will be put on line to provide the additional heat removal capability needed to maintain an acceptable CCWS temperature in the event of the design basis LOCA and most limiting single failure. While the heat exchanger is a passive component and is expected to be unavailable on rare occasions and then not for any significant length of time (Connell Tr. D-551; Wermiel, Tr. D-2900), if the second heat exchanger is not available, the plant must be shut down. Thus, the requirements of GDC-44 have been met. Connell et al., ff. Tr. D-487, at 37; Wermfel, ff. Tr.
D-2864, at 1-4; Staff Ex. 55; Connell, Tr. D-543-46.
330. The Board finds that the technical specification limitation on water temperature provides level of safety as called for by GDC-44.
. 4
Conclusions of Law The Board has considered the record in this reopened proceeding and concludes as follows:
1.
The scope of the verification program, including both the IDVP and ITP, gives reasonable assurance that any significant deficiencies in the final design of Diablo Canyon due to failure to properly implement a quality assurance program meeting the requirements of 10 CFR 50 Appendix B have been discovered.
2.
The IDVP has properly carried out its assigned duties in accordance with the Comission Order and Staff Letter of Novenber 19, 1981, and thus gives reasonable assurance that the overall program has found and appropriately resolved all questions on the adequacy of the design at Diablo Canyon.
3.
The IDVP and ITP combined have implemented the IDVP finding:; of design deficiencies identified during the course of the IDVP in a manner which gives reasonable assurance that the necessary fixes have been accomplished in a proper manner.
4.
All activities of the IDVP and ITP have been conducted in accordance with the quality assurance requirements of 10 CFR 50 Appendix B.
5.
There is reasonable assurance that the existing design of Diablo Canyon meets applicable licensing criteria and commitments.
6.
The activities authorized by this license can be conducted without endangering the health and safety of the public insofar as the issues discussed herein are concerned.
7.
The issuance of this license will not be inimical to the comon defense and security or to the health and safety of the public.
// '
t t
It Is Ordered, in accordance with the Atomic Energy Act of 1954, as amended, and the Comission's regulations, and based on the findings and conclusions set forth herein, that the Director of Nuclear Reactor Regulation is authorized to issue a full-power operating license for the Diablo Canyon Nuclear Power Plant, Units 1 and 2, consistent with the Board's decisions in this case, subject to the Comission's determination and order.
Respectfully submitted, ROBERT OHLBACH PHILIP A. CRANE, JR.
RICHARD F. LOCKE DAN G. LUBBOCK Pacific Gas and Electric Company P. O. Box 7442 San Francisco CA 94120 (415) 781-4211 ARTHUR C. GEHR Snell & Wilmer 3100 Valley Center Phoenix AZ 85073 (602) 257-7288 BRUCE NORTON THOMAS A. SCARDUZIO, JR.
Norton, Burke, Berry & French, P.C.
P. O. Box 10569 Phoenix AZ 85064 (602) 955-2446 Attorneys for Pacific Gas and Electric Company Dated: December 9, 1983 By Bruce Norton l
l t
)
APPENDIX A 1.
The scope of the IDVP review of both the seismic and non-seismic aspects of the designs of safety-related systems, structures and components (SS&Cs) was too narrow in the following respects:
(a) The IDVP did not verify samples from each design activity (seismic and non-seismic).
(b) In the <tesign activities the IDVP did review, it did not verify samples from each of the design groups in the design chain performing the design activity.
(c) The IDVP did not have statistically valid samples from which to draw conclusions.
(d) The IDVP failed to verify independently the analyses but merely checked data of inputs to models used by PGandE.
(e) The IDVP failed to verify the design of Unit 2.
2.
The scope of the ITP review of both the seismic and non-seismic aspects of the designs of the safety-related systems, structures and components (SS&Cs) was too narrow in the following respects:
(a) The ITP did not verify samples from each design activity (seismic and non-seismic).
(b) In the design activities the ITP did review, it did not verify samples from each of the design groups in the design chain performing the design activity.
(c) The ITP did not have statistically valid samples from which to draw conclusions.
(d) The ITP has failed systematically to verify the adequacy of the design of Unit 2.
1
- A
- n i
t 3.
In various situations listed below the ITP used improper engineering standards to determine whether design activities met license criteria. In j
some of these situations the IDVP either used or approved the use of such improper standards or did not verify them at all.
(f) The ITP's modeling of the soil properties for the containment and auxiliary buildings was improper in that:
(1) in the soil structure interaction analysis of containment for the DE and the DDE, use of boundary motion inputs to the model were improperly used; (ii) the soil structure interaction analysis for containment for the DE and the DDE uses a 7 percent damping value for rock, which is unconservative, especially for the DE; (iii) the dynamic analyses of the containment for all earthquakes omit any analysis of uplifting of the foundation mat; (iv) the modeling of the soil springs for the auxiliary building does not specify soil properties; (v) in the modeling of the soil springs for the auxiliary building, the motion inputs to the lower ends of the springs does not account for all soil structure interaction phenomena that could be expected.
(o) The ITP has not demonstrated, and the IDVP has not verified, that the DCP modeling of the seismic response of the fuel handling building is proper, in that the DCP has not adequately justified the use of the translational and torsional response of the auxiliary buf1 ding as input to the fuel handling building nor has it demonstrated the validity of the dynamic degrees of freedom selected.
(ITR 57)
- A,
(p) The ITP has not demonstrated, and the IDVP has not verified, that the DCP seismic model of the slabs in the auxiliary building is proper, in relation to the use of vertical and rotational springs to model the columns, and the motions used as input at the ends of the springs not connected to the slabs.
In addition, in the study of the diaphragms, the ITP has not adequately accounted for the inplane flexibility of these slabs, and has not adequately demonstrated that stresses are within allowable limits at all elevations.
(ITR 55)
(q) The ITP has not demonstrated and the IDVP has not verified, that the soils analysis for the buried diesel fuel oil tanks is proper in I
that the values of the exponent shown in Figure 14 of ITR 68 have not been demonstrated to be appropriate and the variation of shear velocity with depth is not properly justif.ied.
(ITR 68)
(r) The ITP has not demonstrated and the IDVP has not verified that the soils analysis for the auxiliary saltwater piping and circulating water intake conduits is proper in that the selection of the modulus versus strain curve utilized is not justified.
(ITR 68)
(s) The ITP has not demonstrated and the IDVP has not verified that l
the seismic analysis of the turbine building is proper in that bolt l
bearing capacities were taken from an inappropriate source.
(ITR 56)
(t) The ITP has not demonstrated and the IDVP has not verified that l
the seismic analysis of the turbine building is proper in that the use of four different models for the vertical analysis has not been justified.
(ITR 56) 4.
The IDVP accepted deviations from the licensing criteria without providing adequate engineering justification in the following respects:
- A.
h L
i
(a) Contrary to the requirements of FSAR Section 17.1 regarding compliance of the as-built in'tallation with the design documents, the IDVP review of the AFWS disclosed that the as-built installation failed to meet the design drawings in that (1) a steam trap on the turbine-driven AFW pump steam supply line is not provided and (ii) there are discrepancies in the arrangement of the long-term cooling water supply line.
(b) Contrary to FSAR Section 8.3.3., the electrical design does not fully comply with the comitments regarding separation and color coding.
(h) Contrary to PGandE's September 14 and December 28, 1978 licensing commitments, CRYPS equipment identified in the FSAR as necessary to maintain control room habitability during safe shutdown has not been evaluated regarding the effects of a moderate energy pipe break.
(1) The fire protection for the motor-drive AFW pump room is not consistent with the PGandE licensing commitment for fire zone separation as stated in its November 13, 1978 Supplemental Information for Fire Protection Review (SIFPR) in that:
(1) there is a large grated ventilation opening in the ceiling of the room; (2) a fire damper has gaps when it is closed; (j) The fire protection for the AFW pump room is not consistent with the PGandE licensing commitment for cable separation as stated in its SIFPR of Noved er 13, 1978 in that:
(1) the pumps for the motor-driven AFW pumps and the control circuitry for a flow control valve necessary foe operation of the turbine driven AFW pump are located in a single fire zone;
- A.
I k
i i
(2) cables for some AFW circuits are not routed in accord with descriptions in the SIFPR and four AFW circuits PGandE committed to identify and review in the SIFPR were not included in 4
that document.
(k) Contrary to the licensing commitment set forth in its SIFPR of November 13, 1978, each of the three 4160 volt cable spreading rooms has a ventilation opening leading up to the 4160 voit switchgear rooms.
(1) Contrary to FSAR Section 3.6, possible jet impingement loads have not been considered in the design and qualification of safety-related piping and equipment inside containment.
(q) Contrary to PGandE's December 28, 1979 licensing comitment letter to the NRC, modifications to protect two Auxiliary Feedwater valves from the effects of moderate energy line breaks were not implemented.
(r) Contrary to the licensing comitment to maintain minimum system redundancy as stated in FSAR Section 3.6A (NSC evaluation of pipe break outside containment), four components were identified for which high energy line cracks could cause temperatures in excess of the specification temperatures of the components.
(s) Contrary to the licensing commitment to maintain minimum system redundancy as stated in FSAR, section 3.6A (NSC evaluation of pipe break outside containment), a conduit was identified whose failure due to a high energy line crack could eliminate redundant Auxiliary Feedwater system flow.
(t) Contrary to the FSAR Section 8.3 comitment to provide switchgear buses with adequate short circuit interrupting capability, the
- A.
i
e i
calculated duties for circuit breakers on 4160 V buses F, G, and H were above the nameplate ratings for those buses.
(u) Contrary to single failure criteria stated in FSAR Section 3.1.1, reviews of the Auxiliary Feedwater and Control Room Ventilation and Pressurizaton systems identified circuit separation and single failure deficiencies. Similar deficiencies were identified in additional verification reviews, which included other safety-related systems.
5.
The verification program has not verified that Diablo Canyon Units 1 and 2 "as built" conform to the design drawings and analyses.
6.
The verifiction program failed to verify that the design of safety related equipment supplied to PGandE by Westinghouse met licensing criteria.
7.
The verification program failed to identify the root causes for the failures in the PG&E design quality assurance program and failed to determine if such failures raise generic concerns.
8.
The ITP failed to develop and implement in a timely manner a design quality assurance program in accordance with 10 CFR Part 50, Appendix B to assure the quality of the recent design modifications to the Diablo Canyon facility and the IDVP failed to ensure that the corrective and preventative l
action programs implemented by the ITP are sufficient to assure that the Diablo Canyon facilities will meet licensing criteria.
9.
Contrary to General Design Criteria 44 (GDC-44) of Appendix A to 10 CFR l
Part 50, PGandE has f ailed to provide adequate assurance of component cooling water system (CCWS) heat removal safety function capacity in that the maximum ocean water temperature of 64 F is not conservative because it has already 0
l been exceeded in 1983. Furthermore, a technical specification limitation which permits plant operation at reduced power levels in lieu of enlarging the i
- A '
i
o capacity of the CCWS does not provide an equivalent level of safety as compliance with the requirements of GDC-44 (SSER 16 (Aug. 1983) and September 1983 ocean water temperature readings).
//
//
//
//
//
//
//
//
//
//
//
//
//
//
//
//
//
//
//
//
//
//
//
- A,
b
. _. _ =. -. _
..... ~
APPENDIX B Pacific Gas and Electric Company Exhibit No. Description Identified Admitted 86 Commission Order (CLI-81-30) dated D-250 0-251 November 19, 1981 I
87 NRC Staff Letter dated November 19, 1981 D-250 D-251 88 Diablo Canyon Nuclear Power Plant Design D-250 D-251 Verification Program Management Plan, Phase I (March 29,1982) 89 Diablo Canyon Nuclear Power Plant Design D-250 D-251 Verification Program Management Plan, Phase II (June 18, 1982) 90 Independent Design Verification Program D-250 D-251 Final Report, Diablo Canyon Nuclear Power Plant - Unit 1 (as revised through October 10,1983) l 91 Pacific Gas and Electric Company Phase I D-250 D-251
(
Final Report - Design Verification Program (as revised October 14,1983) l 92 Pacific Gas and Electric Company Phase D-250 D-251 l
l II Final Report - Design Verification I
Program (as revised October 11, 1983)
Interim Technical Reports (ITRs) 1 Through 68 of IDVP ITR f
93 1
Additional Verification and D-250 D-251 Additional Sampling (Phase I) l l
- B l l
j t
'.t-
- )
9' Exhibit No.
ITR Description Identified Admitted 94 2
Comments on the R. F. Reedy, Inc.,
D-250 D-251 Quality Assurance Audit Report on Safety-Related Activities Performed by PGandE Prior to June 1978 95 3
Tanks D-250 D-251 96 4
Shake Table Testing D-250 D-251 97 5
Design Chain D-250 D-251 l
98 6
Auxiliary Building D-250 D-251 99 7
Electrical Raceway Supports D-250 D-251 100 8
Independent Design Verification D-250 D-251 l
Program for Verification of PGandE Corrective Action 101 9
Development of the Service-Rei ded D-250 D-251 Contractor List for Non-Seismic Design Work Performed for DCNPP-1 Prior to June 1, 1978 102 10 Verification of Design Analysis D-250 D-251 Hosgri Spectra 103 11 PGandE-Westinghouse Seismic D-250 D-251 Interface Review j
104 12 Piping D-250 D-251 105 13 Soils - Intake Structure D-250 D-251
//
//
- B.
Exhibit No.
ITR Description Identified Admitted 106 14 Verification of the Pressure.
D-250 D-251 Temperature, Humidity, and Submergence Environments Used for Safety-Related Equipment Specifications Outside Containment for Auxiliary Feedwater System and CRVP System 107 15 HVAC Duct and Supports Report D-250 D-251 108 16 Soils - Outdoor Water Storage Tanks D-250 D-251 109 17 Piping - Additional Samples D-250 D-251 110 18 Verification of the Fire Protection D-250 D-251 Provided for Auxiliary Feedwater System Control Room Ventilation and Pressurization System Safety-Related Portion of the 4160V Electric System 111 19 Verification of the Post-LOCA Portion D-250 D-251 of the Radiation Environments Used
(
for Safety-Related Equipment Specifica-i tion Outside Containment for Auxiliary t
Feedwater System and Control Room Ventilation and Pressurization System i
112 20 Verification of the Mechanical /
D-250 D-251 Nuclear Design of the Control Room l
Ventilation and Pressurization System
//
- B,
I
}
l i
l Exhibit No.
ITR Description Identified Admitted 113 21 Verification of the Effects of High D-250 D-251 Energy Line Cracks and Moderate Energy Line Breaks for Auxiliary Feedwater System and Control Room Ventilation and Pressurization System 114 22 Verification of the Mechanical /
D-250 D-251 Nuclear Portion of the Auxiliary Feedwater System 115 23 Verification of High Energy Line D-250 D-251 Break and Internally Generated Missile Review Outside Containment for Auxiliary Feedwatar System and Control Room Ventilation and Pressurization System 116 24 Verification of the 4160V Safety-D-250 D-251 Related Electrical Distribution System 117 25 Verification of the Auxiliary D-250 D-251 Feedwater System Electrical Design l
118 26 Verification of the Control Room D-250 D-251 Ventilation and Pressurization System Electrical Design l
//
p
//
- B '
i l- --
-~
t
r e
Exhibit No.
ITR Description Identified Admitted 119 27 Verification of the Instrument and D-250 D-251 Control Design of the Auxiliary Feedwater System 120 28 Verification of the Instrument and D-250 D-251 Control Design of the Control Room Ventilation and Pressurization System 121 29 Design Chain - Initial Samples D-250 D-251 122 30 Small Bore Piping Report D-250 D-251 123 31 HVAC Components D-250 D-251 124 32 Pumps D-250 D-251 125 33 Electrical Equipment Analysis D-250 D-251 126 34 Independent Design 'ierification of D-250 D-251 DCP Efforts by SWEC 127 35 Independent Design Verification D-250 D-251 Program Verification Plan for DCP i
Activities 128 36 Final Report on Construction Quaifty D-250 D-251 Assurance Evaluation of G. F. Atkinson 129 37 Valves D-250 D-251 130 38 Final Report on Construction Quality D-250 D-251 Assurance Evaluation of Wismer and Becker
//
- B \\
i l
1
r Exhibit l
No.
ITR Description Identified Admitted 131 39 Soils - Intake Structure Bearing D-250 D-251 Capacity and Lateral Earth Pressure 132 40 Soils Report - Intake Sliding D-250 D-251 Resistance 133 41 Corrective Action Program and Design D-250 D-251 Office Verification 134 42 R. F. Reedy, Inc., Independent D-250 D-251 Design Verification Program Phase II Review and Audit of PGandE and Design Consultants for DCNPP-1 135 43 Heat Exchangers D-250 D-251 136 44 Shake Table Test Mounting Class lE D-250 D-251 Electrical Equipment 137 45 Additional Verification of Redundancy D-250 D-251 of Equipment and Power Supplies in Shared Safety-Related Systems 138 46 Additional Verification of Selection D-250 D-251 of System Design Pressure and Temperature and Differential Pressure Across Power-Operated Valves 139 47 Additional Verification of D-250 D-251 Environmental Consequences of Postulated Pipe Ruptures Outside of Containment
- B l l
1 Exhibit No.
ITR Description Identified Admitted 140 48 Additional Verification of Jet D-250 D-251 Impingement Effects of Postulated Pipe Ruptures Inside Containment 141 49 Additional Verification of Circuit D-250 D-251 Separation and Single Failure Review j
of Safety-Related Electrical Equipment 142 50 Containment Annulus Structure D-250 D-251 Vertical Seismic Evaluation 143 51 Containment Annulus Structure-D-250 D-251 Verification of DCP Corrective Action 144 54 Corrective Action Containment D-250 D-251 Building 145 55 IDVP Verification of Corrective D-250 D-251 j
Action Auxiliary Building 146 56 Corrective Action Turbine Building D-250 D-251 147 57 Review of DCP Activities Fuel D-250 D-251 Handling Building 148 58 Verification of DCP Activities D-250 D-251 Intake Structure 149 59 Corrective Action targe Bore Piping D-250 D-251 l
150 60 Corrective Action Large and Small D-250 D-251 Bore Pipe Supports 151 61 Corrective Action Small Bore Piping D-250 D-251
//
- B.
k
~
t
r Exhibit No.
ITR Description Identified Admitted 152 63 Corrective Action HVAC Ducts, D-250 D-251 Raceways, Instrument Tubing and Associated Supports 153 65 Corrective Action Rupture Restraints D-250 D-251 154 67 Corrective Action, Equipment D-250 0-251 155 68 Verification of HLA Soils Work D-250 0-251 156 SECY-82-89 D-251 0-254 157 SECY-82-414 D-251 D-254 158 Memorandum of Commission Secretary D-251 D-254 dated March 8, 1982 159 Memorandum of Commission Secretary D-251 D-254 dated December 9, 1982 160 Diablo Canyon Project Project D-251 D-254 f
Engineer's Instruction No. 13 Revision 1 161 PGandE Engineering Department D-251 D-254 Procedures 3.6 ON and 3.7 l
162 JCAE Hearings D-2099 163 IDVP Program Management Plan Procedure D-2235 DCNPP - IDVP-PP-007 Rev. 3 Governor's Exhibits GOV-ll INPO Report dated January 25-30, 1982 D-2085 D-2087 GOV-12 ITR-ll, Rev. O D-2085 D-2087 GOV-13 PGandE Letter dated October 12, 1981 D-2085 D-2087
- B.
0 k
Exhibit No.
Description Identiffed Admitted GOV-14 NRC Draft NUREG 0862 0-2085 D-2087 GOV-15 Potential Program Resolution Report D-2085 D-2087 dated October 11, 1982 (E0I 7002)
GOV-16 OIR, Rev. 5. File 7002 D-2085 D-2087 G0Y-17 OIR, Rev. O. File 8010 D-2085 D-2087 GOV-18 Error Report, Rev. 8 File 8010 D-2085 D-2087 GOV-19 Prog. Res. Report, Rev. 11, File 8010 D-2085 D-2087 GOV-20 OIR, Rev. O. File 8017 D-2085 D-2087 GOV-21 OIR, Rev. O, File E01-8022 D-2085 D-2087 GOV-22 Error Report, Rev. 5. File 8022 D-2085 D-2087 GOV-23 OIR, Rev. O. File E01-8023 D-2085 D-2087 GOV-24 Prog. Res. Report, Rev. 5 File 8023 D-2085 D-2087 Gov-25 OIR, Rev. 0, File E01-8060 D-2085 D-2087 l
GOV-26 Error Report, Rev. 5 File 8060 D-2085 D-2087 GOV-32 "Q"-List D-264 D-1447 GOV-33 QA Management Audit Report #317, D-935 D-936 f
December 20-28, 1982 l
GOV-34 Intra-Company Memo Dated November 2 D-856 D-857 1982, "Look Back Review Sumary for PGandE Design Activities and Corporate QA Program"
//
//
//
- B.
t
~
n 4
Exhibit No.
Description Identified Admitted GOV-35 Review of the PGandE QA Manual for D-897 D-996 Nuclear Power Plants - Summary Report May-June, 1982 - Project Assistance Corporation 1
GOV-36 EDS Letter to Frank Dodd from: Ron D-921 D-996 Polivka, Dated June 7, 1982 - Project Summary Report GOV-37 Project Engineer's Instruction 5 D-930 D-934 Rev 0, August 10, 1982 GOV-38 Project Engineer's Instruction 5 D-930 D-934 Rev. 1, October 29, 1982 GOV-39 Project Engineer's Instruction 5 D-933 D-934 Rev. 2, March 3, 1983 GOV-40 Memo to D. A. Brand from C. E. Ralston, D-951 D-954 Dated April 21, 1983, " Audit Report to Management" GOV-41 Harding Lawson Figure IV 2 - Variation D-782 D-782 of Shear Modulus with Shear Strain GOV-42 Audit No. EQ-8303, April 8, 1983 D-954 D-957 Engineering Manual Audit Results Attachment B GOV-43 NCR DC0 79 EN004 - Implementation of D-890 D-895 URS/Blume QA Program
//
- B l It
s r
Exhibit No.
Description Iden,tified Admitted GOV-44 NCR DC0 79 EN006 - Dcsign Interface -
D-893 D-895 Engineering Quality Control GOV-45 "On the Quantitative Definition of D-1335 D-1446 Risk," Risk Analysis. Vol.1. No.1, 1981, by Stanley Kaplan & John Garrick GOV-46
" Safety Goals and Related Questions,"
D-1337 D-1446 Reliability Engineering 3, 1982, S. Kaplan GOV-47 Letter Teledyne to D. G. Eisenhut D-1463 D-1464 (NRC) Dated October 15, 1982 Regarding Region V Comments on Phase II Activities GOV-48 R. F. Reedy, Inc. Audit Checklist D-1627 D-1630 Audit No. 014-003-A03-6 GOV-49 R. F. Reedy, Inc. Audit Checklist D-1627 D-1630 Audit No. 014-003-A03-5 GOV-50 Audit Note Coversheet No. 014-003-A03-7 D-1651 D-1657 GOV-51 Draft ITR-14. Rev. 1. October 27, 1982 D-1563 D-1586 GOV-52 Draft ITR 14, Rev. 1 D-1592 D-1607 GOV ~M Excerpt IDVP Final Report, Rev. O, D-1551 D-1554 Page 3.5.8 GOV-54 Excerpt IDVP Final Report, Rev.1 D-1551 D-1554 Page 3.5.8 GOV-55 Safety and Evaluation Report - Review of D-2576 D-2634 Diablo Canyon Turbine Building
- B.
i
1*
Exhibit No.
Description Identified Admitted GOV-56 Summary and Evaluation Report, IDVP D-2576 Pumps, ITR-32, Rev. O GOV-57 SER; IDVP
's; ITR 37 Rev. O D-2576 GOV-58 SER; Verifici. tion of Corrective Action D-2576 for Equipment, ITR-66, Rev. O GOV-59 Memo to Enge1 Ken from Denton, Dated D-2750 November 9, 1982, Diablo Canyon Design Verification Program GOV-60 Memo to Norman Romney from David D-2576 Rubinstein, Dated September 1, 1983 on Reactions on PGandE Evaluation Spot Weids GOV-61 Transmittai Letter to P. A. Crane from D-2981 NRC and Inspection Report No.
50-275/83-09 l
GOV-62 Transmittal Letter to J. O. Schuyler D-2990 l
l from NRC and Inspection Report No.
50-275/83-22 GOV-63 SER; Verification of Corrective Action D-2576 for Equipment - ITR 67, Rev.1 GOV-64 SER; Corrective Action HVAC Ducts, D-2576 Electrical Raceways, Instrument Tubing and Associated Reports, ITR 63, Rev. 0
//
- B,
e
4 i
Exhibit No.
Description Identified Admitted GOV-65 SER; IDVP Review of Corrective Action, D-2576 Large and Small Bore Pipe Supports, ITR-60, Rev. O GOV-66 SER; Verification of Correction Action, D-2576 Large GOV-67 Transmittal Dated August 20, 1983 from D-2917 Mr. Bishop to J. O. Schuyler and attached Notice of Violation; Inspection Report No. 50-275/83-24; 50-323/83-17 Joint Intervenors' Exhibits JI-128 NRC, " Quality Assurance Case Study D-427 D-3031 Working Paper, Case C," Draft Working Paper (July 1983)
JI-129 QA Review & Audit. Phase I by: Reedy D-999 D-1000 Inc. on:
Safety-related Activities Performed by PG&E prior to June 1978 JI-130 BNL, " Independent Seismic Evaluation D-2611 D-2612 of the Diablo Canyon Unit 1 Containment l
Annulus Structure and Selected Piping Systems" (May 1982)
JI-131 Letter, Denton to Cooper (July 1, 1982)
D-2611 D-2613 JI-132 Letter, Reich to Kuo (May 17,1983)
D-2611 D-2614 i
//
//
- B.
.1
e i
Exhibit No.
Description Identified Admitted JI-133 ITR 34, Draft (December 1982) Phase II D-1792 D-1795 Additional Sample & Additional Verification JI-134 Design Criterion Memorandum M-65 Rev. O D-579 JI-135 FSAR Section 3.6 D-585 JI-136 Letter dated 12/18/72 from Giambusso D-591 D-651 to PGandE JI-137 Summary & Evaluation Report ITR 9 D-1682 Rev. 0 10-15-82 Development JI-138 Letter, Cloud to Hoch dated 11-23-82 D-1759 D-1760 JI-139 Letter, Maneatis to Eisenhut dated D-3040 D-3047 September 21,1983, re: " Quality Assurance Case Study Working Paper Case C" l
JI-140 Changes from Draft to Final Case Study C D-3047 NRC Staff Exhibits 36 SER, NUREG-0675, Supplement No. 18 D-2464 D-2467 37 SER, NUREG-0675, Supplement No. 19 D-2464 D-2467 38 SECY-82-89 D-2464 39 SECY-82-414 D-2464 40 NRC Inspection Report: 50-275/82-31 D-2464 D-2467 41 NRC Inspection Report: 50-275/82-36 D-2464 D-2467 l
l 42 NRC Inspection Report: 50-275/82-41 D-2464 D-2467
)
50-323/82-19 14 -
/
I h
p Exhibit No.
Description Identified Admitted 43 NRC Inspection Report: 50-275/82-42 D-2464 D-2467 44 NRC Inspection Report: 50-275/83-14 D-2464 D-2467 50-323/83-11 45 NRC Inspection Report: 50-275/83-27 D-2464 D-2467 50-323/83-19 46 NRC Inspection Report: 50-275/83-26 D-2464 D-2467 47 NRC Trip Report: Memorandum, Herring D-2464 D-2467 to Miraglia, dated February 3, 1982 48 NRC Trip Report: Memorandum, Herring D-2464 D-2467 to Miraglia, dated March 3, 1982 49 Region V Inspection Report: 50-275/82-02 D-2464 D-2467 50-323/82-02 50 Region V Inspection Report: 50-275/82-17 D-2464 D-2467 51 Region V Inspection Report: 50-275/82-20 D-2464 D-2467 1
50-323/82-10 i
52 Region V Inspection Report: 50-275/82-30 D-2464 D-2467 50-323/82-14 53 Region V Inspection Report: 50-275/83-04 D-2464 D-2467 50-323/83-03
//
//
//
//
// 15 -
i
('
i-
l e
f J
Exhibit l
No.
Description Identified Admitted 54 Board Notification 83-135 A (Case C D-2464 D-2910 Study dated 9-19-83, including the October 14th memo to the Commissioners from Eisenhut; a September 28, 1983 memo to Eisenhut from Taylor; a August 2, 1983 letter to PGandE (Maneatis) from Taylor; and a September 27, 1983 transmittal memo from W. D. Altman, Case Study Team Leader.
55 SER, NUREG-0675, Supplement No. 16 D-2465 D-2467
//
//
//
//
//
//
//
//
//
//
l
//
l
//
//
//
//
i l
- B l i
a o
9 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of
)
)
PACIFIC GAS AND ELECTRIC COMPANY
)
Docket No. 50-275
)
Docket No. 50-323 Dichlo Canyon Nuclear Power Plant, )
Units 1 and 2
)
CERTIFICATE OF SERVICE The foregoing document (s) of Pacific Gas and Electric Company han (have) been served today on the following by deposit in the United States mail
- properly stamped and addressed:
Judge John F. Wolf Mrs. Sandra A. Silver j
Chairman 1760 Alisal Street Atomic Safety and Licensing Board San Luis Obispo CA 93401 US Nuclear Regulatory Commission Washington DC 20555 Mr. Gordon Silver 1760 Alisal Street Judge Glenn O. Bright San Luis Obispo CA 93401 Atomic Safety and Licensing Board US Nuclear Regulatory Commission John Phillips, Esq.
Joel Reynolds, Esq.
Wechington DC 20555 Center for Law in the Public Interest Judge Jerry R. Kline 10951 W. Pico Blvd. - Suite 300 Atomic Safety and Licensing Board Los Angeles CA 90064 US Nuclear Regulatory Commission Washington DC 20555 David F. Fleischaker, Esq.
P. O. Box 1178 Mrs. Elizabeth Apfelberg Oklahoma City OK 73101 c/o Betsy Umhoffer 1493 Southwood Arthur C. Gehr, Esq.
Scn Luis Obispo CA 93401 Snell & Wilmer 3100 Valley Bank Center Janice E. Kerr, Esq.
Phoenix AZ 85073 Public Utilities Commission State of California Bruce Norton, Esq.
5246 State Building Norton, Burke, Berry & French, P.C.
350 McAllister Street P. O. Box 10569 Sen Francisco CA 94102 Phoenix
- 4 85064 Mrs. Raye Fleming Chairman 1920 Mattie Road Atomic Safety and Licensing Shall Beach CA 93449 Board Panel US Nuclear Regulatory Commission Mr. Frederick Eissler Washington' DC 20555 Scenic Shoreline Preservation Conference, Inc.
Via Network Courier for 4623 More Mesa Drive Saturday delivery.
Senta Barbara CA 93105 s
R I
m..-
b o
Chzirman
- Judge Thomas S. Moore Atomic Safety and Licensing Chairman Appeal Panel Atomic Safety and Licensing US Nuclear Regulatory Commission Appeal Board Wcshington DC 20555 US Nuclear Regulatory Commission Washington DC 20555 i
S2cretary US Nuclear Regulatory Commission
- Judge W. Reed Johnson Washington DC 20555 Atomic Safety and Licensing Appeal Board Attn:
Docketing and Service US Nuclear Regulatory Commission Section Washington DC 20555
- Ltwrence J. Chandler, Esq.
- Judge John H. Buck Hsnry J. McGurren Atomic Safety and Licensing US Nuclear Regulatory Commission Appeal Board Office of Executive Legal Director US Nuclear Regulatory Commission Washington DC 20555 Washington DC 20555 Mr. Richard B. Hubbard
- Michael J.
Strunwasser, Esq.
MHB Technical Associates Susan L. Durbin, Esq.
1723 Hamilton Avenue Suite K Peter H. Kaufman, Esq.
San Jose CA 95125 3580 Wilshire Blvd.
Suite 800 Los Angeles CA 90010 Mr. Carl Neiberger Tolegram Tribune
'** Maurice Axelrad, Esq.
P. O. Box 112 Lowenstein, Newman, Reis, and Snn Luis Obispo CA 93402 Axelrad, P.C.
1025 Connecticut Ave NW
- Eric Havian Washington DC 20036 HDller, Ehrman, White & McAuliffe 44 Montgomery Street - 30th Floor San Francisco CA 94104 Via Network Courier for Saturday delivery.
Express Mail for Monday delivery.
),
1 Date:
December 9, 1983
/
Philip
' Crane, Dr.
( i
.