ML20081B855
| ML20081B855 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/05/1984 |
| From: | Burns E, Doss K, Etzler R, Gilsclon J, Hoch J, Randy Patterson, Rhodes R, Shiffer J, Thornberry R PACIFIC GAS & ELECTRIC CO., WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20081B795 | List: |
| References | |
| NUDOCS 8403120029 | |
| Download: ML20081B855 (234) | |
Text
UNITED STATES OF AERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD
)
In the Matter of ) Docket Hos. 50-275
) 50-323 PACIFIC GAS AND ELECTRIC )
COMPANY ) (Design Quality Assurance)
)
(Diablo Canyon Nuclear Power )
Plant, Units 1 and 2) )
)
AFFIDAVIT OF J. D. SHIFFER, R. PATTERSON, J. M. GISCLON, K. C. DOSS, J. B. HOCH, R. C. THORNBERRY, R. D. ETZLER, R. K. RHODES, AND E. M. BURNS
(
STATE OF CALIFORNIA )
) ss.
, CITY AND COUNTY OF SAN )
FRANCISCO )
j The above being duly sworn, depose and say:
I, Janes D. Shiffer, am Manager of Nuclear Plant Operations in the Nuclear Power Generation Departnent of Pacific Gas and Electric Cmpany.
I, Robert C. Thornberry, an the Plant Manager for Diablo Canyon Power Plant.
I, Robert Patterson, an Assistant Plant Manager for Diablo Canyon Power Plant, g
I,' John M Giscion, an Assistant Plant Manager for Diablo Canyon Power Plant.
M book o!h% PM
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( I, Ken C. Doss, am a menber of the Cnsite Safety Review Group for the Diablo Canyon Power Plant and from September,1977 to April,1982 was a Senior Instrument and Controls Supervisor of the Instrument and Controls (I&C)
Departnent for Diablo Canyon Power Plant.
I, John B. Hoch, am Project Manager for the Diablo Canyon Project.
I, Richard D. Etzler, am Project Superintendent for the Diablo Canyon Project.
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I, R. Keith Rhodes, am Technical Services Supervisor for the Diablo Canyon Project.
I, Edward M. Burns, an a Lead Licensing Engineer for Westinghouse Electric Corporation.
- 1. The purpose of this affidavit is to respond to the affidavit of John Cooper dated January 23, 1984, and filed with the Joint Intervenors' k Motion dated February 14, 1984. Mr. Cooper's allegations can be broadly grouped into three categories: (1) that the design of the Residual
- Heat Removal (RHR) System is deficient; (2) that there was an inadequate management response when Mr. Cooper expressed his concerns regarding alleged safety considerations; and (3) that Pacific Gas and Electric Company (PGandE) took unwarranted retaliatory actions against Mr. Cooper as a result of his raising safety concerns.
- 2. This affidavit will discuss Mr. Cooper's allegations based upon our personal knowledge of the issues and events as they transpired.
Statements of our background and experience are attached.
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( I. Residual Heat Removal System Design
- 3. Most of Mr. Cooper's affidavit deals with his efforts, while an employee of PGandE, to bring to the attention of managenent and the NRC his concerns over the design adequacy of certain parts of the RHR systen.
Mr. Cooper was employed by PGandE as a construction inspector and an
> operations department instrunent and controls
- naintenance technician between March 1976 and November 1979, and again as a construction field engineer between April 1981 and March 1982. The design issues raised by Mr. Cooper have all been considered and responded to by PGandE nanagement, as well as by the NRC in their subsequent evaluation and detemination of design adequacy.
- 4. The NRC has thoroughly reviewed and evaluated the design of the RHR
(. system at Diablo Canyon. As documented in the Safety Evaluation Report (SER) for Diablo Canyon, NUREG-0675, October 1974, the NRC has found that the design of the RHR system met all of their safety requirements.
Further, in later supplements to the SER, the NRC specifically reviewed the single RHR suction line from the RCS hot leg design in Supplement 7, dated May 1978 (Exhibit 1) and the RHR interlock design for RHR overpressure protection in Supplement 8, dated November 1978 (Exhibit 2) and found these designs acceptable. Further, allegations made by Mr.
Cooper subsequent to his employment at PGandE have been extensively considered and resolved by the NRC in Supplement 21 dated December 1983. The NRC's response to Mr. Cooper's allegations is documented at Pages 2-85 through 2-113 of SER Supplement 21 (Exhibit 3).
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- 5. The primary function of the RHR systen is to renove residual and sensible heat fron the core and reduce the temperature of the Reactor Coolant Systen (RCS) during the second phase of plant cooldown. During the first phase of plant cooldown, the tenperature of the RCS is reduced (
by transferring heat from the RCS to the stean systen using the stean generators and Auxiliary Feedwater Systen (AFWS). To acconplish the second phase of plant cooldown, the RHR systen is aligned to take suction fron one RCS loop hot leg and punp the reactor coolant through the RHR heat exchangers back to the cold legs of the RCS loops. A schenatic drawing of the RHR systen is attached (Exhibit 4).
- 6. The RHR systen is also utilized as part of the Safety Injection Systen (SIS) and Containnent Spray (CS) systen. The SIS provides energency core cooling in the unlikely event of a break in either the RCS or stean If required to operate as part of the SIS, the RHR punps, along
( systen.
with the centrifugal charging punps and safety injection punps, function initially to inject borated water from the Refueling Water Storage Tank ,
(RWST) into the RCS. When this injection is complete, the RHR syster. is aligned to deliver water fron the containment sunp through the RHR heat The exchangers and back to the RCS for long-tem decay heat renoval.
RHR systen also functions as the water source for the containment spray systen during post-LOCA recirculation. Water from the containment sunp is delive' red to the containment spray rings by the RHR punps, as well as to the suction of the SIS punps.
- 7. Mr. Cooper's affidavit focuses on the availability of the RHR systen to renove residual heat from the core. For heat renoval during the second phase of a nomal cooldown, the redundant pump / heat exchanger trains of the RHR systen both take suction from the RCS Loop 4 hot leg via a single suction line. Two notor-operated valves (8701 and 8702) are located in series in this line to isolate the RHR systen fron the higher pressure of the RCS when the RHR systen is not operating in the ,' cay heat renoval node.
- 8. Mr. Cooper alleges that, since there is one suction line fron the RCS to the RHR systen, the single failure criterion is not net if either of the two valves should fail in the closed position (Cooper Affidavit at 5).
His allegation is incorrect. While later Westinghouse designs have employed two separate suction lines, Diablo Canyon and nost of the other
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operating Westinghouse plants enploy the single suction line design.
These other operating plants include North Anna 1 and 2, Beaver Valley 1, Zion 1 ar.d 2, D. C. Cook 1 and 2, Salen 1 and 2, Surry 1 and 2, Sequoyah 1 and 2, and Trojan. The acceptability of the single RHR suction line design with two isolation valves has been documented in l
l the NRC Staff's SER, Supplenent 7, dated May 1976, page 3-3 (Exhibit 1) and Supplenent 21, dated December 1983, page 2-95 (Exhibit 3).
- 9. The Diablo Canyon design fully neets the Diablo Canyon licensing criteria. General Design Criterion (GDC) 34,10 CFR Part 50, Appendix A, post-dates the Diablo Canyon design and is not part of the Diablo Canyon licensing basis. However, as stated at page 3.1 A-11 of the FSAR, the Diablo Canyon design does, in fact, comply with the intent
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( of GDC-34. The decay heat renoval safety function is provided by the diverse, safety-related AFWS (together with the stean generators and the atnospheric dunp valves).
- 10. The AFWS is specifically designed to renove decay heat under nomal circunstances as well as under circunstances such as loss of offsite power where the reactor coolant punps are not operating. The capability of the AFWS to perfom this function has been denonstrated at other plants and will be denonstrated in the special los power test progran scheduled for Diablo Canyon.
- 11. Mr. Cooper also alleges (Cooper Affidavit at 5) that the RHR systen is not redundant for accidents other than a large break LOCA, such as a small break LOCA where only snall anounts of water are released inside containment and where the core is danaged. Mr. Cooper is wrong in
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several respects.
- 12. The RHR rysten functions similarly for a snall break LOCA as it does for a large break LOCA. However, the type of accident deccribe:1 by Mr.
Cooper, a snall break LOCA with insufficient loss of reactor coolant to go into the recirculation node, would not involve any core danage and f would not require that the RHR systen perfom its SIS function. If the 1
l operators should choose to use the RHR systen in its nomal cooldown l
l node (whirch they would not have to do), and it is unavailable because of a closed valve on the single suction line, the containment should still i remain accessible and the operators could manually open the valve.
- 13. For small break LOCAs where there is a large anount of water released to containment, the operators could initiate the recirculation mode, just
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as in a large break accident, by injecting water fron the RWST into the RCS and then switching over to recirculation when there was sufficient water in the containment sump. In the recirculation node, the RHR i
system is aligned into redundant flow paths fron the containnent sunp. l l
Thus, the decay heat renoval function of the Diablo Canyon design satisfies all single failure requirenents.
- 14. Mr. Cooper's other main technical concern, which is repeated throughout his affidavit, seems to be that the design of the RHR systen and related circuitry is such that the valves are subject to being closed by unanticipated electrical signals. This has happened at Diablo Canyon on two occasions in the past during naintenance and testing activities.
Since then, the problen has been solved by revising plant procedures to
' require power to be renoved fron valves 8701 and 8702 at all times other than when the valves are being operated. This practice, which has been approved by the NRC and included in the plant's Technical Specifications
-(Exhibit S), precludes the possibility of valve closure by unanticipated signals.
- 15. Mr. Coo;,er's proposal is to avoid unanticipated signals by redesigning the power sources for the valve relays so that they bypass the Solid State Protection System (SSPS). The PGandE resolution is based upon evaluations by Westinghouse, the systen designer, which show there are reasons ccnsistent with Westinghouse design standards for redundant protection channels to have all such signals, including the RHR valve relay signals, channeled through the SSPS. By modifying plant procedures to preclude the potential for the unanticipated signals, both
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the Westinghouse design standards and Mr. Cooper's concerns are directly ,
snd effectively accorredated. In arriving at this resolution, all of Mr. Cooper's concerns on this issue, as described in his affidavit and as raised by him in his May 19, 1981 menorandun to Mr. Gisclon (Cooper Affidavit at 9), his October 10, 1981 Design Change Request (Cooper Affidavit at 11,12), and his discussions and correspondence with the NRC (Cooper Affidavit at 9,11,12) were specifically and appropriately
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addressed. Mr. Cooper's affidavit is dedicated to pronoting his own solution, but gives no reasons whatsoever why the NRC-approved resolution is inadequate. See SSER 21, Decenber 1983, at pages 2-85 through 2-113, and NRC Inspection Reports 50-275/82-26, 50-323/82-13 and 50-275/82-42 (Cooper Affidavit, Exhibits 10b and 10c).
- 16. Mr. Cooper (Cooper Affidavit at 10 and 13) expressed concern over the lack of control reon annunciation to alert the operator of loss of RHR
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fl ow. In fact, PGandE has connitted to install an RHR low flow alam.
In addition, there are several other means for the operator to determine that loss of RHR flow has occurred due to the closure of an isolation valve. Flow through the RHR systen is indicated in the control roon on l flow instrunents in the RHR/ SIS return lines. The RHR suction valvc positions are indicated in the main control roon by red / green status i lights near the valve control switches. The suction valves are included
. on a monitor light box such that the nonitor light is on when the valves L
l are open.
- 17. In sunnary, the design of the Diablo Canyon RHR systen is a standard Westinghouse design enployed in most of its operating plants and is fully in accord with the Diablo Canyon licensing basis requirenents.
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s Plant procedures have been inplenented to preclude the possibility of unanticipated valve operation in the RHR systen in accordance with the requirenents of the operating license Technical Specifications.
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( II. Management Responsiveness
- 18. Mr. Cooper suggests that PGandE has not been responsive to concerns raised by employees, particularly his concerns about the RHR suction valve interlocks (Cooper Affidavit at 1). In fact, the opposite is true.
- 19. PGandE has had a longstanding policy of encouraging employees to identify items or areas of concern and bring them to the attention of management. Managenent, in turn, is responsible for assuring that these concerns are appropriately resolved in a timely manner consistent with their significance, regulatory requirenents, and other ongoing work activi ties..
2 0.- This policy has been embodied in various management directives which l
have been issued over the years (See Exhibits 6-11). To assure that
( this policy was effectively inplemented on a day-to-day basis, fomal l
systems were developed to document and track problems. These systens l
are described in written procedures developed by each department. For example, Nuclear Plant Operations (NPO) has issued Nuclear Plant Administrative Procedure C-12: " Identification and Resolution of Problems and Nonconfomances." Similarly, General Construction (GC) has issued G.C.Q. A. Program Procedure GCP-12.1: " Documenting Discrepancies and Assessing Reportability" (Exhibits 12 and 13).
- 21. Under the NPO system, each identified problem involving equipnent, design, materials, proccJures, etc., is first documented on a Nuclear Plant Problem Report (NPPR). Any individual who identifies a problem may initiate a NPPR, regardless of tne organzation with which he is affiliated. Employees are encouraged, and for many types of situations
{ are required, to initiate NPPRs, even for what may be considered to be
( trivial matters. p.'cer the NPPR is generated, it is routed to plant supervision for approval and then to plant management for review. Plant management review serves several purposes: (1) to assess potential ,
significance and reportability, (2) to assure that the NPPR is routed to the proper group for resolution, (3) to propose and/or concur in a resolution, and (4) to establish a priority for resolution.
l C 22. Any item which is judged to be potentially reportable or which meets significance criteria established by the PGandE Quality Assurance Department is elevated to the level of a potential nonconfomance, and a Nonconfomance Report (NCR) is initiated. The NCR is used by all departments, and all potential nonconfomances are fomally dispositioned by an ad hoc cormittee which includes representatives fron Quality Assurance as well as other affected departments. Each NPPR and
( NCR is tracked by computer until ultimately resolved.
- 23. Under these policies, an enomous number of such documents are initiated, all of which must be carefully controlled to assure appropriate disposition. In the years 1978 through 1983, approximately 20600 NPPRs were generated at Diablo Canyon. Of these NPPRs, approximately 3700 were initiated by the Instrunent and Controls Maintenance Department, in direct contradiction to Mr. Cooper's assertion that there was a policy within this group to discourage the l
writing of NPPRs.
1 With the large numbers of NPPRs which have been generated, a 24.
correspondingly heavy burden is placed on plant management for timely and responsible disposition. Accordingly, some prioritization of these tasks is required. Inevitably, some items will remain open longer than
{ others, depending on the prioritization process. However, the plant
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staff at Diablo Conyon continues to work aggressively to close out these items, and they will be closed out on a schedule consistent with need, significance, and regulatory requirements. For example, as of January 31, 1984, approximately 18500 of the aforementioned NPPRs have been dispositioned. All NPPRs required to be closed out prior to fuel loading were closed out. Similarly, all NPPRs required to be closed ou; prior to comencetant of heatup were dispositioned prior to comencing heatup. This process will continue for initial criticality and ultimately for full power operation.
- 25. A similar situation exists regarding the disposition of Design Change Requests (DCRs). Since the design verification progran comenced in late 1981, approximately 3500 DCRs have been initiated. The sheer number of DCRs has required prioritization . As a result, certain itens remain open longer than others. At this time, approxinately 3360 have been dispositioned by Engineering. Again, these itens are closed out on a schedule consistent with need, significance, and regulatory requirements.
26, As discussed below, in each of the instances cited by Mr. Cooper in his affidavit, particularly those related to his concerns over the RHR system, the issues he raised by memorandun, NPPR, DCR or other means were responded to by management and properly dispositioned prior to the time when such dispositions were needed.
- 27. Mr. Cooper (Cooper Affidavit at 9) states that he wrote a memorandun to Mr. Giscion on May 19,1981, " explaining that valves 8701 and 8702 would fail closed when SSPS [ solid state protection system] output fuses were removed and that emergency procedure OP-8 ' Control Room Inaccessibility'
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( was in error." Contrary to the implication in the affidavit, Mr.
Cooper's concern was addressed on July 30, 1981, through the issuance of WPR DCO-81-TI-P0237.
- 28. During the interval between May 19 and July 30, 1981, Mr. Gisclon had been engaged in discussions with the NRC staf f regarding the Technical Specifications, including the manner in which valves 8701 and 8702 would be operated in their role for low-temperature RCS overpressure protection. The staff agreed in July that power should be removed fron valves 8701 and 8702 during operating nodes 4, 5, and 6 when the reciprocating charging pump was in operation. The NPPR issued on July 30,1981 thus encompassed and responded to the topics raised by Mr.
Cooper's nemorandun and established that the energency operating procedure would be revised to require that the circuit breakers on
( valves 8701 and 8702 be opened after the valves were opened.
- 29. Mr. Cooper suggests (Cooper Affidavit at 9) that he was reprinanded for sending the memorandun to Mr. Giscion, and warned against going to the NRC. In fact, Mr. Cooper had ignored the established procedures by sending a nenorandum directly to a departnent head in an entirely different organization. He should have initiated a NPPR or a Minor Variation Report, or initiated a Design Change Request (DCR), any of which would have officially entered his concern into fornalized tracking 1
systems. In this case, there was no reason for Mr. Cooper's failure to follow established procedures, and he was requested by supervision to follow these procedures in the future. This was not only reasoaable, but necessary. The established reporting systens would break down and concerns would not be systematically and adequately addressed if
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employees bypassed the established procedures and sent then by menorandum directly to managenent, particularly in another organization.
- 30. Mr. Cooper was not warned against going to the NRC or threatened with the loss of his job as he alleges (Cooper Affidavit at 9). He was requested by supervision, Mr. P. Gilbreath and Mr. R. D. Etzler, to attenpt to resolve problens within his organization and within PGandE prior to reporting directly to the NRC. At no time was he told, either directly or by inplication, that he would lose his job or be subject to disciplinary action. In fact, Mr. Cooper connunicated with the NRC on several occasions after the alleged threat (Cooper Affidavit at 9,11, 12-13) with neither loss of job nor reprinand or criticisn fron nanagement.
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- 31. Mr. Cooper initiated a DCR on October 10,1981 (Cooper Affidavit at 11, 16-17). In that DCR he requested that the valve interlocks be renoved, since the renoval of power fron the valves would nake the interlocks no longer usable. While he conplains that his DCR "has been sitting in some engineer's in-basket since 1981, unreviewed and unresolved," (Cooper Affidavit at 16), he had previously noted (Cooper Affidavit at 11) that the Gnsite Safety Review Group (0SRG) had reviewed and recomended rejection of the requested design change on the grounds that it would increase the probability of overpressurization of the RHR systen. .The DCR, by now a decidedly low priority iten, was officially dispositioned by Engineering, which rejected the requested design change in 1983. Procedures were correctly folloucd, and disposition was appropriately made prior to the tine when the RHR systen would be called into service to remove decay heat.
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- 32. Mr. Cooper states (Cooper Affidavit at 13) that his notification of an
(- error in the Plant Manual "had not been corrected eight months after the original notification." This involved a correction to Volume 16 of the Plant Manual ider.tified by Mr. Cooper in April 1981. The correction involved a Hi-Lo Level alarn on the reactor coolant pump lube oil system which was incorrectly identified as a Lo Level alam. A long tem revision of Volume 16 was in progress at that time. Since the plant was
' not then in operation, plant management decided to incorporate the correction into the long tem revision. The milestone requirement In established for completing this revision was prior to plant startup.
late 1982, in order to address Mr. Cooper's concerns, and at the suggestion of the NRC, a plant engineer, Mr. R. L. Fisher, issued a revision (on-the-spot change) to the existiag version of Volume 16.
This revision was ultimately superseded by the complete revision of
{ Volume 16 issued in September 1983, prior to Unit 1 fuel load in November 1983.
- 33. Mr. Fisher also discussed the suggested correction with Mr. Cooper in June 1981 and advised him in writing via a memorandun dated June 9, 1981, that appropriate changes had been nade in the draft copy of Volume
- 16. In a response to Mr. Cooper's memorandum to Mr. Thornberry, dated December 19, 1981, Mr. Fisher again docunented the status of this matter in a memorandum dated January 11, 1982. The January 11 memorandun contained a marked-up copy of the Volune 16 pages showing Mr. Cooper that his concern was b:.ing addressed.
- 34. Mr. Cooper states (Cooper Affidavit at 13) that NPPRs he had w itten three years previously had not been resolved. The two NPPRs referred to
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( by Mr. Cooper involved a change to a PGandE drawing to correct an air supply pressure for a control component, and a change to two vendor These NPPRs drawings to show the correct instrunent channel nunbers.
were closed out in Septe:fber 1983 and September 1982, respectively. As discussed previously, problems are corrected and NPPRs closed consistent with the time at which equipment is required for plant operation. Both problem reports would have been addressed more rapidly if required sooner by plant operational or maintenance needs.
- 35. Mr. Cooper alleges without explanation "that the FSAR description of 1
this systen [ presumably the RHR systen] was incorrect, and PGandE refused to change it" (Cooper Affidavit at 13, 14-15). To the contrary, PGandE has been and is engaged in a conprehensive revision of the FSAR t
as required by 10 CFR 50.7)(c). In accordance with an extension granted
( by the NRC, the revision is due on September 22, 1984. The revision will include the updated description of the RHR systen.
- 36. Mr. Cooper (Cooper Affidavit at 23) alleges an "unuritten policy" to attenpt to cover up plant deficiencies. There is no policy, written or f
unwritten, to cover up any deficiencies identified in plant design, construction, or operation. In fact, as discussed above, PGandE's policy is just the opposite. Indeed, the most significant example which refutes this claim is PGandE's response to the discovery of the diagran design error in September 1981. At that time, PGandE reported it to the NRC and voluntarily stopped preparation for fuel loading at Diablo Canyon Unit 1.
- 37. NPO documents and reports deficiencies, concerns, and niscellaneous work itens using documents such as NPPRs, NCRs, and Licensee Events Reports
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(LERs). During 1983, 6409 NPPRs 113 NCRs, and 36 LERs were issued at Diablo Canyon. This level of documentation shows that PGandE is actively resolving problems, not covering then up. ,
- 38. It is the policy of the management at Diablo Canyon that all NPPRs, and all such other connunications involving safety concerns, be addressed and resolved. The resolution of such problen reports is perfomed on a schedule consistent with when the problen itself nuit be resolved. In other words, if there is a problen report which affects the safe operation of a particular systen, it will be resolved prior to declaring the systen " operable? If a stated problem is not worthy of action, it is so stated in the answer / resolution section of the problen report before the report is te tiinated or conpleted. PGandE has identified i
five nenoranda which Mr. Cooper generated in the spring of 1979. All of the concerns and all of the points raised in these nenoranda have been
' addressed. All of the problen reports which Mr. Cooper specifically referred to in his affidavit have similarly been addressed and closed.
- 39. Mr. Cooper alleges (Cooper Affidavit at 24) that "problen reports" are destroyed by nanagenent. Managenent at Diablo Canyon has never deliberately destroyed or voided a NPPR in order to cover up or disniss a plant safety issue. Any employee found deliberately destroying a problen report for such a reason would be subject to disciplinary action.
- 40. Any person who believes he has identified a plant problen can write a NPPR to docunent the problem. When a NPPR is generated, it initially is reviewed by first level supervision to verify thtt a legitinate problen exists. If supervision determines that the problen is not valid, the NPPR can be voided. Real problens identified with safety-related
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l If procedures or safety-related equipment are not voided by management.
a NPPR is voided, an explanation is provided to the person who initiated the NPPR. If the employee is dissatisfied with the explanation or resolution and continues to believe a valid problem exists, he can pursue the matter through successive levels of management via the "Open Door Policy." Once entered into the tracking system, if a NPPR is voided by supervision, it remains a pennanent plant record.
- 41. In December 1983, completely independent of Mr. Cooper's allegation, the Diablo Canyon Quality Control Department conducted a management review of the NPPR logging and tracking system. The status of the approximately 20600 NPPRs generated during the period from 1978 through 1983 was examined in the management review. Prior to the advent of the current conputerized logging system, NPPRs were tracked with a e
( handwritten log maintained by the department. When the current computerized system was established, the records for previous years were entered into the system. A searth of the computerized records showed that 100 NPPR nLmbers in the ISC Department were indicated to be missing. Based on a review of the handwritten NPPR log and other l
l records, it was detennined that nine were missing due to data entry errors,11 were shown to have been dispositioned,19 are still open and will be dispositioned as scheduled, and 61 were voided by the supervisor
! in charge. The voided WPPRs are being researched to ensure that the perceived problem has been adequately addressed. Fe11owing this process, they will be officially closed out. At the present time, approximately 16 of the voided NPPRs have been closed and we anticipate completion by March 31, 1984. Thus, there are no identified instances of destruction of problem reports or improper voiding of NPPRs.
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- 42. Since the 1978-79 time frame referred to by Mr. Cooper, the system for tracking NPPRs has been significantly upgraded and improved. Under current procedures, after approval by the supervisor, a number is assigned to the NPPR and a brief description is entered into the Records Management System (RMS). The status of the NPPR is tracked using this computer-based system until the NPPR is completed and closed. The current procedures are intended to ensure that a NPPR cannot be lost or remain unprocessed, although occasionally, due to clerical errors in processing the large number of NPPRs, some may become temporarily lost.
In all such cases an investigation is conducted to locate the NPPR and resolve the issue.
- 43. Mr. Cooper alleges (Cooper Affidavit at 24) that he was " reprimanded" for correcting an error. The accusation takes on a sonewhat different hue when examined in its proper context. Administrative controls are in
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place to assure proper implementation, documentation, and resolution of problems. Notwithstanding Mr. Cooper's statement that all that was l
involved was "a simple wiring change," the error " corrected" by Mr.
Cooper constituted an unauthorized design change, contrary to the The requirements for approval and control of all design changes.
procedures for resolving the error were underway, through comunications l with Westinghouse, when Mr. Cooper made the unauthorized change.
Unauthorized changes of the type made by Mr. Cooper constitute serious l
violations of procedures which are deserving of reprimand.
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- 44. Mr. Cooper alleges (Cooper Affidavit at 25) that maintenance technicians were " routinely denied access to necessary infomation," and that "the problem of inadequate reference materials was not corrected." The
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a statements are not true. Contrary to Mr. Cooper's allegations, required documentation was always available to the technicians, and significant improvements were made in the degree of availablility and accessibility of the documents.
- 45. The documentation in question consisted of design infomation desired by the instrumentation and control technicians to perfom their maintenance ,
activities. At no time were the technicians " denied access" to documents. The required documentation was always available, although the limited number of controlled copies sometines made access inconvenient. When management was requested, in a letter dated April 3, 1979 from the technicians, to improve docunent accessibility, timely action was taken to distribute and make available more copies of such i
high-use documents. It should be noted that a " satellite" file, or extension of the plant master file, had been established in the
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instrument shop since about 1975. Also in 1983 an RMS data teminal was I located in the I&C Maintenance shop area to provide more convenient access to this system.
- 46. Several technicians also had concerns about the adequacy of existing plant procedures. Mr. Giscion responded properly and reasonably by requesting the technicians to document any inadequacies that were believed to exist. This documentation cane in the fom of four memoranda and associated NPPRs from Mr. Cooper. All of the items raised in the memoranda were addressed and responded to.
- 47. NPO has also established a fom41 program whereby relevant experience from other power plants, as obtained from such sources as the NRC, the INPO SEE-IN (Significant Event Evaluation - Infomation Network),
.s
. , , , . .. . _. ...- -. .... - .- - --.--. _- - - - _ . - - _ - - _ - - - ~ - - - - .
Nuclear Notepad, and NOMIS (Nuclear Operations and Maintenance ;
~
Infomation Service) is regularly disseminated to personnel at DCPP.
l This program was established in 1981 as part of our response to requirements contained in NUREG-0737, " Clarification of TMI Action Plan i
Requirements."
- 48. And finally, Mr. Cooper's allegation (Cooper Affidavit at 26) that l management " destroyed" individual files is actually a criticism of PGandE for proper records control. In 1982, a continuing program was in 4
1 progress to assure that current infomation was available to all personnel working on safety-related equipment. In many areas, management found it necessary to remove unofficial documentation from working files to assure that the documentation in use reflected the actual configuration of the plant, i.e., if < design change had been made to a circuit, it might not have been translated to a technician's
(
notes and unofficial drawings. Using this incorrect, unofficial documentation could have resulted in a maintenance error.
- 49. Mr. Cooper (Cooper Affidavit at 10, 11-12) alleges that a violation of internal control procedures occurred in the disposition of the NPPR l
initiated to document the spurious closure of valve 8701 on Septenber 29, 1981. The concern was that the NPPR had been finally dispositioned without initial plant management review. The allegation is correct, insofar as a portion of the NPPR fom had not been signed as required.
However, i.he reviews required by the initiation of the NPPR were properly, adequately, and timely perfomed.
5 50. The NPPR fom has three principal sections. The top section is where the problem is reported. The middle section is for an initial management review, where the responsible department head looks at the
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circumstances and decides whether the problem is a potential
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nonconfonience or is potentially reportable, establishes a priority for l 1
- resolution, and identifies any further instructions he feels to be ,
appropriate. The bottom section indicates the final resolution and '
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sign-off.
- 51. Review of this NPPR indicates that the botton section was signed, indicating final resolution, but that the middle section was not signed as required by Administrative Procedure C-12 to indicate the initial
. review. Mr. Sexton, the responsible nanager, cannot recall the exact I
circumstances surrounding the NPPR. He notes, however, that because the resolution (issuance of a procedure change) and an initial punp test and inspection were conpleted imediately after the problem occurred, and prior to his seeing the NPPR, the initial review was superfluous.
Ordinarily, resolution of a problem is nct accompli:;hed so quickly, and the two-step process indicated on the NPPR fom is gemane.
- 52. Plant Managenent responded to the September 29, 1981 valve closure in an appropriate nanner. The Shift Foreman's log for that date clearly indicates that the Operations Department, in conjunction with the Maintenance Department, inspected the pump after the valve closure and the_ Operations Department successfully completed Surveillance Test Procedure P-38 to detemine operability of the pump. Further, a procedure change was-initiated by the Operating Department on September
, 29, 1981 to Operating' Procedure B-2 'to help prevent a recurrence of the event. Additional followup actionL ook t place on this event. The OSRG
- - initiated a second NPPR (NPPR DC-1-81-NO-P0010) on November 13, 1981, j
' ~
i calling.for a' perfomarce test (P-3A) to be run on the pump and the k '
perfomance t6st was successfully completed.
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III. ALLEGED NANAGEENT RETALIATION
- 53. Mr. Cooper suggests throughout his affidavit (Cooper Affidavit at 1, 9, 12, 19, 24 27) that he has suffered retaliation from nanagement because of his efforts to identify safety concerns and bring them to management's attention. There is no policy, written or unwritten, Pemitting or encouraging management to take retaliatory action by any means against employees who raise safety concerns.
In fact, the policy is just the opposite; it encourages employees to identify safety concerns and specifically reassures them that such action will not be hamful to them.
- 54. This pclicy is strongly stated in Mr. Thornberry's menorandum on the "Open Door Policy" (Exhibit 9):
Voicing concerns regarding plant safety or operation will
(- not be documented in the employee's personnel record and iKil never be used for any type of disciplinary action.
I give my fim guarantee of that.
- 55. We have carefully reviewed the cin:umstances described in Mr. Cooper's affidavit. The incidents cited by Mr. Cooper involved no elements of retaliation or punishment for attempting to bring real or perceived concerns to the attention of management or the NRC, and we find no evidence that he was disciplined or retaliated against in any way for expressing his concerns. This will become clear in the following discussions of his various allegations.
- 56. Mr. Cooper alleges (Cooper Affidavit at 9) that he was threatened with See paragraph loss of his job if he spoke to the NRC. This is false.
30 above.
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- 57. Mr. Cooper's allegation (Cooper Affidavit at 24) that he was reprimanded for correcting an error is true. As discussed in paragraph 43 above, Mr.~ Cooper's actions constituted an unapproved design change in
-violation of plant procedures, and was deserving of reprimand.
- 58. Mr. Cooper (Cooper Affidavit at 25) alleges that retaliation was taken against whistleblowers by assigning them to the "nore distasteful jobs"
'.or giving them poor perfomance reviews. The only examples he cites, however, are his own. He states that when he found errors in safety-related procedures, he was " isolated" in the coffee room and giveis the " distasteful"' job of rewriting the procedures. He also alleges that in 1979 he was given a poor perfomance review.
- 59. With respect to rewriting procedures, he was the obvious and nost logical candidate for the job, since he himself had identified the needed revisions. It comes as a surprise to learn that he considered it
{ a distasteful job. As Mr. Cooper well knows, there are literally thousandsof prA.edures rJquired for the operation of a nuclear power plant. Prior to operation, the initial preparation of these procedures and subsequent revisions after initial use is a monumental effort which requires the combined efforts of virtually every engineer, foreman, and technician in the plant. To the extent that procedure preparation is
" distasteful," it is a distaste to be shared by all.
- 60. Since the shop area lacked office space for such work, temporary tables were often set up. in the " coffee room" of the administration building.
This roon-had ready access to the central file system located in the admir.istration building. ,No r'.Wi;tions were placed upon Mr. Cooper's 24 .
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freedom to go where he pleased, and he was no more isolated from the
( crew than any other person, including the plant's IAC Supervisor, whose work location was in the administration butiding.
- 61. Mr. Cooper's allegation regarding his perfomance review is equally erroneous. PGandE supervisors are trained and instructed to give perfomance reviews which accurately reflect the employee's true perfomance, including strengths and weaknesses and areas needing improvement. Perfomance reviews are reviewed by at least one higher level of management to insure that the evaluation is fair to the employee. There is no policy at Diablo Canyon to use the perfomance review for retaliation or any other unjustified reason.
- 62. Moreover, the perfomance review quoted by Mr. Cooper in his affidavit was given in 1979, well prior to his becoming, in his words, a "whi stleblower."
{
- 63. PGandE did not consider the perfomance review to be a particularly poor review, This is substantiated by the decision to re-employ Mr. Cooper in April 1931 after he had resigned in November 1979. No reputable conpany intending to retaliate against an employee by using a perfomance review would rehire the employee 1-1/2 years later. One can only speculate why Mr. Cooper wanted to return to work for PGandE in 1981 with such ill feelings toward management which he apparently I
1 developed in 1978-1979.
- 64. Mr. Cooper alleges (Cooper Affidavit at 10) that he was being " punished" However, when his security clearance was interrupted on August 6,1981.
as Mr. Cooper goes on to note, it was reinstated the very next day. Not only can this action not in any way be construed as harrassing, or even
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( significantly inconvenient, no suggestion was ever made to Mr. Cooper that the one-day interruption of his security clearance was in any way retributive.
- 65. Security records that old are routinely removed from the files, so we have no records to check on the incident. Mr. Cooper's supervisors in General Construction have no knowledge of a harrassing er disciplinary Mr. Thornberry certainly
" hold" being put on his security clearance.
did not put a " hold" on Mr. Cooper's security clearance. If Mr.
Cooper's security clearance was, in fact, temporarily interrupted, there For could be several purely administrative reasons for this to occur.
example, if an employee's status level expires in the security conputer, It the employee cannot enter Unit 1 until the status level is updated.
1 is not an uncommon occurrence for employees (or even entire departments)
( to have their status level expire because of various paperwork It problems. Also, security status riust be reinstated every 30 days.
Such is not uncornon to miss a particular individual's 30-day update.
occurrences nomally take less than a day to correct.
- 66. Mr. Cooper alleges (Cooper Affidavit at 27) that PGandE's " Behavioral
' Reliability Program" will be employed to " weed out whistleblowers" and inflict punishment by " involuntary psychiatric examination and loss of The job." This charge is astonishing, and totally without foundation.
behavioral obserntion program in effect at Diablo Canyon is a necessary component of PGandE's required and approved Physical Security Plan and has been developed in accoraance with NUREG 2076, " Behavioral Reliability Program for the Nuclear Industry"; Appendix B of NUREG 0768,
" People Related Problems Affecting Security in the Licensed Nuclear
- 26 -
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( Industry"; and ANSI /ANS Standard 3.3-1982, " Security for Nuclear Power Plants." As part of the NRC-required security program, Diablo Canyon employeet are administered the Minnesota Multiphase Personality Inventory (MPI).
- 67. Mr. Cooper has a history of vociferous opposition to the administration of this program, and notwithstanding his acceptance of employment with PGandE, he has declared to both the press and his employer that he would not work for a company that " spies on its own employees" (Exhibit 14).
That, of course, is not quite how the program works. A detemination that a person exhibits unreliable or untrustworthy behavior which could jeopardize the safety of the general public, the plant staff or the plant itself can be made only after a recornendation by a licensed clinical psychologist or psychiatrist not in PGandE's employment. Thus, there is no way to use the behavioral observation program to " weed out whistleblowers", inflict punishment, or spy on employees.
i 68. Mr. Cooper states (Cooper Affidavit at 12) that he was transferred to a
" remote, snowbound site" as punishment for " speaking out on safety concerns at Diablo Canyon." The facts indicate othenvise.
- 69. Mr. Cooper was notified by supervision, Mr. R. K. Rhodes, on February 24, 1982, that he should report to the Helms Project in the Sierra Nevada mountains, on March 1,1982. His assigned task was to make instrument take-offs in order for General Office Engineering to provide set-point data back to the Helms Project. Mr. Rhodes explained that this was a temporary assignment to which Mr. Cooper was assigned due to his experience and familiarity with instrument systems. His work was completed on approximately Mart:h 9,1982. Mr. Cooper was requested to
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remain at Helms through the end of the week (March 12, 1982) to answer any questions concerning the take-offs in the absence ci the forenan in charge of the instrument group. Mr. Cooper was also told that he would be sent back to Diablo Canyun early in the week of March 15, 1982.
- 70. Instead of complying with his assignment instructions, Mr. Cooper left the-jobsite on March 9,1982, without prior notice or explanation to anyone. When contacted by supervision (Mr. Rhodes) on Monday, March 15, 1982, and asked to explain his actions, Mr. Cooper stated that he had resigned as of March 9,1982. He told Mr. Rhodes he was going to quit anyway because of his dispute with PGandE over the requirenent that he take the MMPI.
- 71. It is standard PGandE policy and practice within General Construction to make work assignnents to employees on relatively short notice for various durations due to the nature of construction and startup work.
Mr. Cooper was definitely not assigned to the Helns Project as punishnent.
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Dated March 5,1984. ,
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/2 J AMES Dr SHIFF [/ ( JOHN B.]C'H
& W- JOHN M. GISCLON hm
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EDWAKD M. ILURN5
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k, e vM KEN C. DOSS Subscribed and sworn to before ne this 5th day of March 1984.
M SEAL Rancy J. Lef. aster, '
Notary Public in and for the City and County of San Francisco,
( State af California.
My comission expires April 14, 1986 rxm:cce.m>:xx:.:s::cca>:::xxonzeas R
I NANCY J. LEMASTER {
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- '.dk,NOIGY f'UC'.lC-CAtlFORNIA Clif AhO CO 'f TY OF xy r:
. , .j:,Y SAN FitA?.CiSCO L:
l My Cemi.issbn Dp:rcs Apr;f 14.1986 Q l
f{AM - . :%*:M ':M%MSTO:cc:MMTcxMxxxXM l
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JAMES D. SHIFFER JOHN M. GISCLON LEN C. DOSS JOHN B. HOCH EDWARD M. BURNS Subscribed and sworn to before me this 5th day of March 1984.
Nancy J. Lemaster, Notary Public in and for the City and County of San Francisco, State of California.
My comission expires April 14, 1986 1 t a m-
~
I ' ROBERT PATTER 5ON RCBERT C. THORNBERRY RICHARDD.ETZLEp
$ W<< WO h wR. KEITH RHOBES Subscribed and sworn to before me this 5th day of March 1984.
..._ m e-ML '
WENDY SPROUL NOMW PU8UC - CAUFORMA WendySproul,$ l Notary Public in and for the City and County of San Luis Obispo, g"j sgim mw g ,,?;
=--;*
State of California.
My comission ' expires June 30, 1986 I
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List of Exhibits
( Exhibit 1. NRC Safety Evaluation Report, Supplement 7, pages 3-3 to 3-4, May
'978 Exhibit 2 NRC Safety Evaluation Report, Supplenent 8, page 7-1, Novenber l
1978 l
Exhibit.3 NRC Safety Evaluation Report, Supplement 21, pages 2-85 to 2-113, Decenber 1983 Exhibit 4 Schematic Diagran of Residual Heat Renoval Systen Exhibit 5 Technical Specifications, Section 3.4.9.3 Exhibit 6 Letter from J. D. Worthington,
Subject:
PGandE Policy on Identifying and Reporting Safety Concernt. April 29, 1976.
Exhibit 7 Menorandum fron R .D. Etzler,
Subject:
Personnel Particpation via the Employee Suggestion Plan, July 25, 1980.
Exhibit 8 Letter fom J. 0, Schuyler,
Subject:
Reaffimation of PGandE Policy on identifying and Reporting Safety Concerns, August 7, 1980.
Exhibit 9 Letter from R. C. Thornberry,
Subject:
Open Door Policy and PGandE Quality Hotline, February 5,1982.
Exhibit 10 Letter from F.W.Mielke, Jr. and B. W. Shackelford,
Subject:
Reaffimation of PGandE Policy on Identifying and Reporting Safety Concerns, March 22, 1982.
L Exhibit 11 PGandE Hotline Notice I
i Exhibit 12 Nuclear Plant Administrative Procedure C-12 " Identification and resolution of Problems and Nonconformances", Revision 6, June 20, 1983.
i Exhibit 13 General Construction Quality Assurance Progran Procedure GCP 12.1 l " Documenting Discrepancies and Assessing Reportability,"
l
{ Revision 3, June 30,1983.
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Article from SAN LUIS OBISP0 TELEGRAM-TRIBUNE dated January 23, Exhibit 14
(, 1982.
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EXH1 BIT 1 - SSER 7, page 3-3, May 1973
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Normal instrimentation is available to maintain surveillance of incertant primary and secondary system parameters such as pressure, tescerature, and water levels.
The applicant has presented an analysis of the primary system water volume shrink- l I- age due to cooldown and has determined that the shrinkage is sufficient to accco-3 -
modate chemical and volume control systes input required for boration and cooling of the pressurizer. Therefore, the normal latoawn system is not required to 4
achieve long-ters cooling with residual heat removal system. Stailarly, the appiteant has demonstrated that degasification of the reactor nolant is not necessary because the potential hydrogen concentration in the solution is less than the saturation value at cold shutdown conditions.
All of the operator actions needed to perform plant cooldown (except for periodic surveillance of the Doron corcentration) can be accomplished from the control mee assuming no single failure. The asp 11 cant has demonstrated that recundant paths or systems are availaale to cerfore the essential functions using qualified equip-ment in the event of a single failure. In some instances this would require operator action outsica of the control room to activate the redundant path.
With regard to the residual heat removal systas, the suction line is a single line
^ with two itolation valves in series. A failure that prevents opening these valves 4 / would prevent activation of the residual heat removal systee for long-tem cold d- shutdown heat removal. An electrical failure could readily be corrected by sanu-ally actuating the valve (s). We were concerned about a possible mechanical failure of one of these valves. The appifcant addressed this question in a fattar dated January 26, 1978, asintaining that the protaatlity of a valve disc seoarating free the valve stem is low enough that it need not be considered in the single failure study.
Based on our review of this matter we havs concluded that this design feature is ceceptacle for the rvasens statec belcw. If mechanical valve failures of the type that preclude opening the valves are consider,1 to be rancos events, then the probability of such failures occurring at the . , a time as a severe earthquake does appear to be quite low. On the other hand, ; h failures say reasonaoly be i
considered to be related to the ear +.hquake. We ao ;it believe that the prob-l l ability of an earthquake causing the failures has been quantified. However, an earthquake should not affect the valves since they are designed to withstand the Hosgri event. Accordingly, although the coatined probability level has not been quantified, it is unlikely that a severe earthquake will occur in concination with l eschanical failures that preclude opening one of the two residual heat removal l
syste, suction valves. Furthermore, if a severe ear *.hquake should occur in com-bination with sechanical failures that preclude opening one of the assidual heat removal suction valves, long-tem heat removal can be accomplisned indefinitely with the steam generators and the auxt?tary feeewater systes rather than the
'$.,' residual heat removal system. As a result, we consicer the likelihot: of this b/ failure in comeination with its potential consequences to De acceptacle.
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g Since, without offsite power, the reactor coolant pumps could not be run to s
provide reactor coolant systes mixing, the applicant has committed to performing a g .
- natural circulation test to demonstrate acecuate boron mixing and accectasie ~
cooldown conditions using natural circulation. These tests will be performed during startup when core heat is available.
The systems and equipment needed to perform these functions will be qualified for the Hosgri event and have been included in the seismic reevaluation program which is discussed in other sections of this supplement.
The following items are still outstanding in this pertion of our review: .
(1) A review of the system function indication available to the coerator fn the control race in connection with performing the shutdown. (Section 7.5 of this supplement)
(2) A review of the seismic qualification of the raw water storage rese*voirs, including:
(a) A review of the potential for seiches causing a significant loss of water from the raw water storage reservoirs. (Section 2.4 of this supolement)
O (b) A review of the potential for slopes sliding into the raw water storage '
reservoirs. (Section 2.5.3 of this supplement) ,)
(c) A review of the properties of the rock underlying the raw water storage reservoirs. (Section 2.5.3 of this supplement)
(d)
A review of the provisions to ensure that the rese-voir would not drain through connected piping that is not cualified. (Section 10.5 of this supplement) j We have reviewed the capability to cool the plant to cold shutdcwn conditions and provide long-term cooling. The applicans has demonstrated that suff feient systems are availacle for residual heet removal with or without offsite power and assusing a single failure in accordance with Critarion 34 of the General Design Critaria.
Stailarly, these systems will be qualified for operation in the event of the Hosgri event in accordance with Criterion 2 of the General Cesign Criteria.
Accordingly, subject to satisfactory resosution of the outstanding questions described above, we find these provisions acceptable.
We will provide our evaluation of the outstanding itees discussed above in a future supplement to the Safaty Evaluation Reoort.
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EXHIBIT 2 - SSER 8, page 7-1, November 1978
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- 7. 0 INSTRUMENTATION AND CONTROL 6
i 7.2 Reactor Trip System In Supplement Number'7 to the Safety Evaluation Report we found the basic seismic scram system proposed by the applicant acceptable.
However, we required further information from the applicant regarding how the system would satisfy our requirements for separation, isolation quality, testability, and qualification for Class lE circuits.
The applicant has provided additional information on this subject in Amendment 67 to the Final Safety Analysis Report. Based on our review of the additional inforsation, we have concluded that the seismic scram system is of a similar design and meets the same criteria as the reactor protection system and is, therefore, acceptable.
We consider this matter resolved.
(Rb Systems Required for Safe Shutdown
%/ 7.4 In Supplement Number 7 we stated that we would require further information about the indication available to the control room operator in connection with performing a shutdown a.fter a Hosgri event. We have now completed our review of this matter and it is resolved as discussed in Section 3.2.1 of this supplement.
7.6 RHR Overpressure Protection Interlocks In the Safety Evaluation Report we described the interlocks for the motor operated valves on the residual heat removal system suction lines (Valves 8701 and 8702).
The interlocks operate on diverse principles to prevent opening the , valves wnen reactor coolant system pressure exceeds about 425 pounds per square inch and to automaticaily close the valves when reactor coolant system pressure exceeds abcut 600 pounds p r ;quare inch. The interlocks are provided to prevent overpressuriza-tion of the residual Peat removal system W eq reactor coolant system pressure is l
high, primarily during operation. We found these interlocks acceptable.
In our fire protection review it was postulated that fire damage to electrical cables could cause both valves te open. To correct this the applicant has proposed m removing power from the valves' motor operators by opening cancel circuit breakers.
I \ Since this wifl prevent the postulated fire damage from opening both valves we Q
7-1
EKHIBIT 3 - SSER 21, pages 2-85 to 2-113, December 1983 Task: Allegation No. 37 '
ATS No.: RV 83A41 SN No.: 83-169(}0/20/83) _
r Characterization The solid state protection system (SSPS) relays that ini'iate closure of RHR letdown isolation valves 8701 and 8702 perform no safety function, reduce the reliability of the RHR system, and cause a potential for RHR pump damage.
Therefore, these relays should be removed.
Implied Significance to Plant Design, Construction, or Operation i
The RHR letdown line contains two isolation valves (8701 and 8702) in series that are normally closed during power operation. These valves are opened when entering Mode 4 (hot shutdown) to allow the RHR pumps to take suction from the reactor coolant system (RCS) to the RHR heat .exchangers for decay heat removal.
Both valves 8701 and 8702 are interlocked so that they will automatically close to isolate the RHR system from the RCS if RCS pressure increases to a pre-de-termined setpoint. This automatic isolation function (performed by the West-inghouse designed SSPS) is provided to protect the low pressure RHR system
, piping,from higher RCS pressures. Isolation is accomplished using a " fail safe" design (i.e., on a loss of SSPS power, valves 8701 and/or 8702 will automatically close). The concern here is that a loss of SSPS power will cause an unwanted (spurious) isolation of the RHR letdown ifne causing event-ual RHR pump damage assuming no operator action, d
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s Assessment of Safety Significance l Isolation of the low pressure RHR system from the high pressure RCS must be prcvided to prevent RHR system overpressurization that could potentially re-sultinalossofcoolantaccident(LOCA)outsidecontainment. Therefore, RHR letdown line isolation is a safety function. Tha SSPS, including relays, which performs this function is safety related and designed to Class 1E re-quirements. Both valves 8701 and 8702 are provided with this automatic clo-sure interlock on increasing RCS pressure so that a single failure will not prevent RHR letdown line isolation. Therefore, the relays used to initiate closure of these valves are essential ano should not be removed.
Diverse indications and alarms are provided in the control room (including a 3-RHR system low flow alarm to be install,ed during the first refueling outage) s' to allow the operator (s) to assess RHR system status and to alert them to potential system degradation. Technical Specification surveillance require-ments at Diablo Canyon include periodic verification of RHR system flowrate
~
when using the RHR letdown line. In addition, diverse means' of decay heat removal (i.e., reactor coolant loops) can be readily made available should the RHR letdown line be inadvertently / spuriously isolated.
e d
2-86
V Based on the above, the staff concludes that the existing SSPS design regard-t ing RHR letdown line isolation is acceptable.
Staff Position This allegation does not involve considerations that question plant readiness for power ascension testing or full power operation.
Action Required None.
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l 31 Task: Allcgatten or Concar'n No. 38 1
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i ATS No. RV83A47 BN No.83-169 (10/20/83) l t I
, Characterizatfor.
PG&E is ignoring evidence that the spurious closure of a n.otor operated valve is not " impossible."
Implied Significance to Plc.it Design, Construction or Operation The allegation suggests the licensee has not sath. Srily analyzed operational data.
Assessment of Safety Significance The alleger has described operating events at the Diablo Canyon facility and other Westinghouse facilities during which motor operated valves in the residual heat removal (RHR) system have, upon spurious initiation of their automatic closure circuitry, moved from the normally open position (for RHR operation) to the closed position, these presenting the potential for damage to RHR pumps.
The staff has examined in (fenth the licensee's actions in resoonse to an event invnivino the sourious initiation of RHR motor operated valve closure as wil es l the concerns exoressed hv the allener renarding the potential for such event, l
, and concluded that timely evaluation and corrective measures were taken to preclude 2-88 l
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repetition cf such conditions. (See Allegation or Concern Nos.: 42 & 44).
Staff Position The staff's position regarding the interlock cricurity which causes automatic closure of the RHR isolation valves is duscussed in Allegation or Concern No. 45. It does appear that the licensee is giving proper attention to the spurious closure of the valves in question.
e-
Task: Allegatien #39
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. ATS No.: RV 83A47 BN No: 83-169 (10/20/83) n Characterization There is no control room annunciation provided to alert the operator (s) when the RHR letdown line has been isolated during Modes 4, 5, and .6 (hot shutdown, cold shutdown, and refueling respectively).
Implied Significance to Plant Design, Construction, or Operation During modes 4, 5, and 6 the residual heat removal (RHR) system is aligned in the shutdown cooling mode by taking suction from reactor coolant system (RCS) loop 4 througn the RHR letdown line to the RHR pumps. The RHR pumps direct flow through the RHR heat exchangers for decay heat removal via the component cooling water (CCW) system, and then back to the RCS cold legs. There are two isolation valves (8701 and 8702) in series located in the RHR letdown line.
If one of these valves should inadvertently close, RHR pump suction would be lost. The, concerns here are loss of decay heat removal capability and poten-i tial damage to the RHR pumps. It has been estimated that pump damage could t ,
occur as soon as 10 to 15 minutes following a spurious isolation of the RHR letdown line.
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Assessment of Safety Significance For those modes of operation where RHR shutdown cooling is used, only one RHR
-train jyr one filled reactor coolant loop is necessary to provide sufficient decay heat removal capability. The Diablo Canyon Technical Specifications require either two RHR trains be operable and/or two filled reactor coolant loops be available in order to allow for single failures. If both RHR trains are being used and the RHR letdown line becomes isolated, the operator (s) would have sufficient time to fill at. least one coolant loop (assuming no loops are filled) for decay heat removal. Control room indications of loss of decay heat removal include RCS temperature, RHR system flow, and RHR pump discharge pressure. With less than the required number of reactor coolant _
/ loops and/or RHR trains operable, the Technical Specifications require im-k mediate corrective actior.s to return the required loop / train to operable sta-tus as soon as possible.
Indication provided in the control room of RHR letdown line isolation in-cludes position indication for valves 8701 and 8702 (red and green position status lights next to the valve control switches on the main control board) as well as RHR ss .J :m flow, pressure, and pump status information. Although these features do provide a capability to assess RHR system status, the staff has recognized the need for installation of a RHR low flow alarm. Accordingly, l
, __ 2-31__ - - _ _ _ _ _ _ _ _ .
the licensee is required to install a RHR low flow alarm during the first re-fueling. This requirement is documented in S;pplement No.13 of NUREG-0675,
" Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2." The staff has concluded that the existing con-trol room indications and procedures are sufficient to assure adequate decay heat renoval in the interim.
Staff Position This allegation does not involve considerations that question plant readiness for power ascension testing or full power operation.
Action Required None.
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Reactor Systems Branch Tack: Allegation #40 ATS No.: RV83A 47 BN No.: 83-169(10/20/83)
Characterization The question raised was with regard to _whether or not the single RHR pump suction line from the RCS hot leg meets safety related standards.
The newer PWRs are designed with redundant RHR pump suction lines from the RCS hot legs.
Implied Significance.to Plant Design, Construction or Operation The RHR suction line from the RCS hot leg in Diablo Canyon contains two isolation valves (8701 and 8702) in series that are normally closed during power operation. When the RHR system is operated as a part of the ECCS, the RHR pump suctions are aligned with either the RWST or the
' containment emergency sumps. The RHR suction line from the RCS hot leg is only used during modes 4 (hot shutdown while RCS temperature is less than323*F),5(coldshutdown)and6(refueling). A postulated failure of either isolation valve (8701 or 8702) in the RHR suction line to open during plant shutdown could prevent the plant from reaching a cold shutdown condition.
i Assessment of Safety Significance -
In the Diablo Canyon SER Supplement No. 7, the staff states that the
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single RHR suction line from the RCS hot leg was acceptable. The staff conclusion was based on the following:
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(1) The Diablo Canyon design has a safety ralated Auxiliary Fesdwater l System (AFWS). The condensate storage tank is the primary source
- of AFW with & bout an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> water supply. In order to ensure the capability to remove heat via the steam generators for extended periods, provisions have been made to connect the raw water reservior to the suction line or the AFW pump. This will provide enough AFW to allow an additional 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of steam generator operation for both units.
(2) The licensee has indicated that the combination of a mechanical failure of the RHR isolation valves and an earthquake results in a risk of about 10% of the core melt risk from all causes calculated in the Reactor Safety Study.
Branch Technical Position RSB 5-1 was not approved at time SSER No. 7 for Diablo Canyon was issued. In accordance with the implementation schedule of BTP RSB 5-1, the Diablo Canyon Units are considered class 2 plants which are not required to fully implement this BTP. Table 1 of BTP RSB 5-1 shows what is necessary to be implemented for class 2 plants. A single RHR suction Ifne from the RCS hot leg is considered acceptable for a class 2 plant as long as a single failure could be corrected by manual actions inside or outside of containment, or the plant could be returned to hot standby until manual actions (or repairs) are accomplished. (page 5.4.7-16 of SRP 5.4.7). Also, BTP RSB 5-1 for class 2 plants requires that the RHR isolation valves have independent, diverse interlocks to protect against one or both valves being open during an RCS pressure increase above the design pressure of the RHR 2-94 l
system. There was no assessment of the degree of compliance of the Diablo Canyon design against BTP RSB 5-1 documented in any staff SSER.
Based on the above facts, the staff evaluation of the subject allegation is as follows:.
The RHR suction line from the RCS hot leg is not required for ECCS functionability. .The RHR pumps take suction from RWST or containment emergency sumps, and serve the ECCS function during a LOCA. The suction
! line from RCS hot leg is used only for modes 4 ( 323'F),5and6. GDC 34 of Appendix A to 10CFR 50 requires that the decay heat removal safety l
i function should be accomplished assuming a single failure. THe Diablo l Canyon design complies with this requirement by having a RHR system plus a safety related AFWs (with steam generators and atmospheric steam dump valves). Based on the above, we conclude that the Diablo Canyon design
! meets GDC 34 and the intent of BTP RSB 5-1. The current RHR design is.
adequate for safe operation at Diablo Canyon.
The staff is currently conducting a reevaluation of the adequacy of the decay heat removal system design of all LWRs. This work is being i
performed as an Unresolved Safety Issue (TAP-A-45), and the Task Action Plan is projected to be complete within one year. Diablo Canyon, will be subject to any new requirements that may result from the work of TAP A-45.
1 r
Staff Position This allegation does not involve considerations that question plant L readiness fo'r power ascension testing or full power operation.
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Action Required None l
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Task: Allegation or Concern No. 41 ATS No.: RV83A47 BN No.: 83-1G9(10/20/83)
Characterization The power source of certain relays is not shown on certain drawings and this caused an operational problem, the failure (closure of RHR isolation valves).
Implied SignO71 cant to Plant Design, Construction or Operation Sufficient infomation may not be readily available to plant operators or maintenance personnel regarding the effects of deenergizing certain portions of plant safety related systems causing unexpected plant behavior which, in turn, can be of safety concern.
Assessment of Safety Significance Preliminary examination by the staff of the drawings and circuit schematics of concern to the alleger revealed that a detailed review of several drawings, circuit diagrams, and logic diagrams is necessary to fully comprehend the effect of the removal of power to the SSPS output relays. This removal of power can cause this RHR hot leg suction valves to close, resulting in potential damage to safety-related RHR pumps, and a condition which may nnt be detectable by operators in the control room.
The alleger's specific concern is that renoval of power to a portion of the l' SSPS on September 29, 1981 did result in unexpected closure of the RHR isolation l
valves with an RHR pump running. (See Allegation or Concern No. 44).
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Examination of facility records and discussions with licensee personnel know-i legable of the circumstances of the event of September 29, 1981 revealed the following. In preparation for " trouble-shouting" the cause of apparent power supply difficulties in a portion of the SSPS, a "... Clearance Request and Job Assignment Sheet" was prossed and approved, as required by plant administrative procedures, to authorize such activity. Subsequent disablement of the power supply (removal of a fuse) caused automatic closure of the RHR isolation valves thus interrupting RHR system flow. Initiation of the closure of the RHR valves j had not been anticipated by either the operation supervisor or maintenance personnel involved in the activities Operations personnel did respond to the unpected closure of the RHR isolation valves in a resonably timely manner such that the RHR pump continued to operate without flow for approximately five l
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minutes. The pump substained no detectable damage in this instance.
It was also revealed in discussions with licensee personnel that a simplified -
sketch of the RHR initiation circurtry has been constructed to clarify inter-actions between various components previously shown only on individual plant drawings and circuit diagrams. The construction of this simplified sketch has resulted in a much improved understanding of the cricuitry by the plant's maintenance as well as operations personnel.
Staff Position Activities involving maintenance or texting of systems associated with the nuclear plant should be planned in advance sufficiently to anticipate the reasonse of such systems when these activities are undertaken. Adequate preplanning measures in this regard appear not to have been taken by the licensee in this instances. However, measures have been taken by the l
licensee to preclude a repitition of the specific occurrence in this instance.
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' No further specific action'is required. The staff will focus attention in this
.-inspectioti program to the preplanning and procedural precautions established by' the licensee in carrying out maintenance and testing activities of a similar nature In t;ie future.
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Task: Allegation or Concern No. 42 ATS No.: RV83A47 BN No.: 83-169 (10/20/83)
Characterization
, Licensee management was unresponsive to recommendations to prevent spuriou closure of the isolation valves on the residual heat removal (RHR) system.
4 Closure of the valves disables operation of the RHR system for decay heat i[ removal.
Implied Significance to Design, Construction or Operation
! A lack of appropriate response by the licensee, could indicate an undesirable
{ 1evel of management sensitivity toward employee concerns and recommendations aimed at improving operation of the reactor facility.
Assessment of Safety Significance Facility records were examined, discussions were held with facility personnel, and observations were made by the staff.
Periodic discussions were also held with the alleger.
Since the alleger's concerns had been examined by Region V inspectors previously, reports of prior inspections were reviewed and discussions were held with Region V inspectors relating thereto. In addition to the specific concern (or allegation) characterized above, other concerns of l
the alleger, as discussed below, were also examined.
2 100
Tha alleg2r had d:cumented concerns regarding spurious closure of the RHR isolation valves because of certain-steps in an emergency operating procedure related to safe shutdown from outside the control room. The licensee's response consisted of the initiation of a nuclear plant problem report, and investigation of the alleger's concern. The licensee's resolution to the concern was to revise the emergency procedure.
A design change request (DCR) authored by the alleger addressed the alleger's more general concern of poter.tial for inadvertant closure of the RHR isolation valves.
A revision to the DCR was subsequently initiated by the alleger providing the Licensee Event Repor'ts (LERs) of other facilities relating to instances of RHR system disablement due to spurious closure of the isolation valves similar to those which were the subject-of the alleger's concern.
The alleger's preliminary evaluation of the DCR determined that the requested change involved an unreviewed safety question requiring prior NRC approval in accordance withs10 CFR 50.59. The DCR is still undericonsideration by the licensee's engineering department, the plant operating department and Westinghouse, r
Preliminary discussions have been' held between the licensee, Westinghouse and the NRC staff relating to an informal proposal by the licensee (supported by Westinghouse) to remove the RHR interlock circuitry from the Diablo Canyon facility. Tne proposals and actions required to resolve-this DCR are still open.
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The staff determined that other procedural changes have been made by the licensee in an effort to preclude closure of the RHR isolation valves from spurious actuation of the interlock circuitry.
The staff also reviewed a concern documented by the alleger in a memo in April 1981 to plant engineering regarding reactor coolant pump bearing oil level annunciators. In postulating a tube failure in the lube oil heat exchanger, the view was expressed that an incorrect alarm response procedure may lead i the operator to take improper action. Written acknowledgement of the alleger's concern was provided by a plant engineer in June 1981, indicating that the procedures m' anual was being revised to resolve the concern. The alleger observed approximately eight months later, that no change to the Plant Procedures Manual had been made. The alleger documented this observation by an additional memo. . The same plant engineer who had previously responded to the alleger responded to this memo. The engineer explained that the Plant Manual had been the subject of an extensive revision effort for the past year and all changes resulting from this effort were to be incorporated into the Manual "... in one major revision" which would be published "... definitely prior to low power physics testing." A major revision, which included the alleger's initial connent, was subsequentiv made to the Manual in September 1983.
During the intervening period between the time of the alleger's second memo and implementation of the major revision to the Manual NRC resident inspectors pursued the alleger's concern with licensee personnel. In response, the licensee implementing a temporary change to the specific procedure of concen. This 2-102
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- l temporary change was accomplished by the issuance of a Procedure ON-THE-SPOT Change in early 1983.
l Etaff Position A period of approximately 2 years appears to be excessive in attempting to resnive the RHR concerns of the alleger. The issue is not yet fully resolved.
However, unusual circumstances did exist in that resolution of the alleger's concerns regarding the RHR system and his specific recommendation to remove the interlock circuitry involve substantial safety analyses by the licensee, as
, well as NRC staff review and approval. In the interim, procedural changes had been implemented by the licensee which had substantially addressed the concern of the alleger. A similar period, approximately 21/2 years, to fonnally address the alleger's concern regarding the accuracy of an annunciator response
/ s procedure also, under nonnal circumstances, appears excessive. In this instance, however, the unusual circumstances of a ma.ior revision to the procedures manual was in progress.
It is the judgment of the staff that there is not a prevailing attitude by licensee management which in itself discourages employees from expressing concerns or making reconnendations for improvement in facility operations.
Action Required
! The Region V staff will give particular attention in its ongoing routine inspection program to evaluate the performance of licensee management in this area.
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Task: Allegation or Concern No. 43 ATS No. RV83A47 BN No.83-169 (10/20/83)
Characterization The loss of the residual heat removal *(RHR) system on 9/29/81 due to unplanned 4
closare of the RHR isolation valves was an event which should have been reported to the NRC in accordance with 10 CFR 50.72. The licensee's failure to make such a report was in violation of NRC regulations.
Implied Significance to Design, Construction, or Operation
. The failure of the licensee to report this occurrence, would indic-te a deficiency in the licensee's management control systems to provide adequate review and reporting of events to the NRC.
Assessment of Safety Significance The circumstances associated with the event were examined by review of facility records and discussions with licensee personnel.
The loss of residual heat removal capacity during a time when significant fission product decay heat is present in the core would have safety signifi-cance. In this particular instance, fuel had not been loaded into the Diablo Canyon Unit 1. Therefore, no fission product decay heat was present and loss of RHR capability had no actual safety significance.
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The intent of then applicable provision 10 CFR 50.72 of the NRC regulations /
I was to insure that holders of operating licenses for power reactors report promptly by telephone to the NRC Operations Center significant events, such as those which involve intitiation of the licensee's emergency plan; the nuclear reactor not be in a controlled or expected condition; fatality or serious injury or radioactive contamination of personnel; or acts which seriously threaten the safety of the reactor or site personnel.
The event in question was reviewed by the staff and it was determined that this event is not required te be reported in accordance with 10 CFR 50.72. Licensee representatives did indicate that an ire:armational report of the event was ta i be made in writing to the NRC.
Staff Position The staff concludes that the event did not meet the reporting requirement of 10
- CFR 50.72.
Action Required None i
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i Task: Allegation or Concern No. 44 l
ATS No. RV83A47 BN No.83-169(10/20/83) i Characterization The licensee failed to properly process a Nuclear Plant Problem Report.
Implied Significance to Design, Construction, or Operation The allegation, could indicate a weakness in the implementation of the licensee's Quality Assurance Program for Operations.
Assessment of Safety Significance The Nuclear Plant Problem Report (NPPR) is the document used at Diablo Canyon to record events such as significant equipment failures and operational problems.
The NPPR fom becomes the record of the identification of a problem, its evalua' -
tion, and the action taken to correct and prevent recurrence.
On September 29, 1981, inadvertent closure of the residual heat removal (RHR) system isolation valves occurred while the RHR pump No.1-1 was running. The alleger's concerns are that the NPPR which was initiated following this event was not processed properly in that it was, " signed off as complete without any plant management review... classified as 'non-reportable' and without any follow-up action such as an RHR pump inspection or investigation into the cause of the event."
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The processing of th9 NPPR was assessed through an examination of facility records; discussions with facility personnel (including all those persons whose identitywasprovidedbythealleger)andtheallager;andobservationsbythe inspectors.
The NPPR record in question was examined. Itwaswrittenon9/21/81andclosed on 10/5/81.
i The resolution of the three issues are as follows: i i
Signed-off without any plant management review The inspector detennined that licensee management, including the. plant super-intendent and operations supervisor, were involved in the review and evaluation of the NPPR.
The alleger's concern included the fact that when he examined the NPPR (after if had been completed) there was no signature to indicate the results of management's evaluation of cause and corrective action (s) taken. The alleger had called this discrepancy to the attention of a QC supervisor, who obtained the proper signature on the NPPR. When the NRC inspector examined the NPPR record (in December 1983) the Operation Supervisor's signature was found on the document, It was undated. In discussions with the NRC inspector, the Operations Supervisor stated he may have signed the NPPR after it had been closed, but he could not accurately recall the circumstances.
NPPR classified as "non-reportable" 2-107
The inspector verified that the NPPR was in fact classified as "non-reportable" by licensee management. The classification is considered appropriate by the staff and is addressed in Allegation or Concern No. 43. !
- No followup action was taken into the cause of the event The NPPR indicated that a revision.to operating procedures was necessary to prevent recurrence of the event, and that such a revision had been implemented.
Facility records indicate that the NPPR relating to the event was the subject of review by the On-Site Safety Review Group (OSRG) on two occassions--
October 19, 1981 and November 24, 1981. On October 29, 1981 the OSRG observed that the operating procedures had been changed, and that a proposed change to remove the RHR isolation valve initiating circuitry had been proposed. The latter, it was determined, was a Design Change Request (DCR) which had been initiated by the alleger (see Task Allegation or Concern No. 42). The OSRG detemined that it would review the event further during a subsequent meeting.
On November 24 '1981 the OSRC directed that an operational test of the RHR pump be conducted, and that the DCR not be approved since it would provide less protection for RHR over pressurization than presently existed.
Staff Position The NPPR was properly processed and subsequently reviewed by the OSRG.
Action Required None.
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Reactor Systems Branch TASK: Allegation #45 ATS NO.: RV 83A47 l
BN NO.: 83-169(10/20/83)
I Characterization:
Section 5.5. of the Diablo Canyon FSAR describes the autoclosure inter-lock for the RHR Suction line isolation valves (8701 and 8702).
j Section 3.4.9.3.a of the Diablo Canyon Technical Specifications requires power to be removed from these isolation valve operators during modes 4 (Hot shutdown when RCS cold lag temperature is less than 323*F), 5 (cold shutdown) and 6 (refueling). This requirement defeats the function of
' autoclosure interlock for the valves.
Implied Significance to Plant Design, Construction or Operation As the result of Technical Specification Section 3.4.9.3.a. the iso-lation valves (8701 and 8702) will be left in an open position with power removed during low pressure / temperature operation of the plant.
The automatic closure interlock to these isolation valves causes them to lose their design function. This will result in a situation in which there is in sufficient isolation capability feature to prevent an intersystem LOCA between the high pressure RCS and the low pressure RHR system, t
i Assessment of Safety Significance
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Section 5.5 of the Diablo Canyon FSAR states that during low pres-sure/ temperature operation, the isolation valves (8701 and 8702) between the RCS and the suction of the RHR pumps are interlocked with a pressure signal to automatically close the valves whenever the RCS pressure increase above approximately 600 psig. S'ction e 3,4.9.3.a of the Diablo Canyon Technical Specification requires the RHR system isolation valves 2-109
O (8701 and 8702) to be open with* power removed from the valve operators while the positive displacement charging pump is in operation. The applicability of the T.S. is during mode 4 when the temperature of any RCS cold leg is less than or equal to 323'F, mode 5, or mode 6 with the reactor vessel head on this Technical Specification requirement defeats the automatic closure interlock function as designed.
Power removal from valves 8701 and 8702 while the RHR system is operat-ing wais required by the staff as the result of a meeting with the licensees on RCS low temperature overpressure protection (LTOP) and RHR pump protection concerns. Since the Diablo Canyon design has only one ;
RHR suction line from the RCS, a spurious automatic closure of the
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isolation valve would result in loss of RHR pump suction flow and would -
result in a RCS pressurization as a result of the loss oi' latdown flow. ,5 However, there was no documentation (SS'ER, letter or meeting minutes) of the staff's basis for requiring power removal from those isolation valves during modes 4, 5 and 6.
, In the Diablo Canyon SER Supplement No.13 section 6.3. (ECCS), dated April 2,1981, the staff concluded that the licensee should be required to provide an alam to alert the operator to a degradation in ECC3 (during long tem recirculation). A low flow alarm was stated to be an acceptable method to satisfy this concern and the staff indicated that -
an alarm should be installed at the first refueling outage. Until then, procedures and dedicated operators were to be implemented during long term recirculation to manage and monitor ECCS performance. There was no documentation to indicate that the licensee comitted to this 2- 110
staff position, nor was this staff position included in the Ofablo Canyon low power license. SRP 5.4.7 (BTP RSB 5-1) requires an autoclosure interlock on the RHR suction line isolation valves. Without po+er to the valve operators, the autoclosure function is defeated.
Based on the above facts, the staff evaluation of the subject allegation is as follows: '
Without power to the isolation valve operators, the plant design does not conform to BTP RSB 5-1, Position B.I.C. for the requirement of autoclosure interlock. By having power available to the isolation valves during shutdowns ensures an event Y (intersystem LOCA) will not occur as a result of the operator failing to close both isolation valves during a return to power.
With power on the isolation valves, a spurious closure of the isolation valves would result in a loss of suction flow to the RHR pumps. Howev-er, the low flow alarm discussed in SSER No.13 would enable rapid operator detection and mitigation. The licensee has informally indicat-ed that a minimum of 10 minutes without adequate suction pressure would be available without pump damage. Also, there are numerous indications available to alert the operator to improper RHR valve alignment ( A list is provided in staff evaluations to allegation No. 37 and 39). -
t Staff Position .
To implement the staff position stated in SSER No. 13, the installation l of a low flow alarm for RHR pump protection is being considered as a
! 2-111
license condition in the Diablo fanyon full power license 5 Additionally, it is the staff position that power be available to the RHR MOVs when in a shutdown condition. However, there is a question as to when these requirements should be implemented. If the low flow
'em were not installed until the first refueling outage, ' reinstating power to the RHR MOVs in the meantime would result in the au:.0 closure interlock being enabled to provide protection against intersystem LOCA I
However, the chances of spurious autoclosure and consequent loss of RHR suction pressure (without the low flow alarm) and of an overpressure event would be increased. If power restoration to the RHR MOVs were not implemented until the low flow alann is installed at the first refueling outage, the chance of loss of RHR suction in the interim is reduced but there is a possibility of an intersystem LOCA. To -
determine which option results in the safest operation of the plant, the staff considered the followinn:
- 1. During the first cycle of operation, plants operate more frequently on the RHR system as a result of maintenance, testing and training requirements for a new plant. Thus, the period of vulnerability to a spurious RHR suction MOV closure may be greater than in subse-quent cycles.
- 2. The RHR relief valve would open to relieve pressure if a plant
- startup were attempted with both RHR MOVs open. It is not, in the staff's judgment, credible to postulate plant startups with both -
MOVs left open. The operator would have to shut at least one MOV l
to continue the plant startup.
_.. 3. Failing to close the second RHR suction MOV would not, in itself, result in an intersystem LOCA. The first MOV must aise fail. The .
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1 first MOV c'an fail in either of two ways by either the "open permissive" interlock failing along with the operator reinstating power to the valve, (it is required to be de-energized) then attempting to open the valve. The second mode of failure would be for the valve to rupture in such a way that flow between the two systems occurred. Both of these failure modes are judged to have an extremely low probability. However, the consequences of an intersystem LOCA could be severe.
4 There have been many occasions .of spurious RHR suction valve
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closures on operating plants. This has resulted in not only a loss of decay heat removal, but also an overpressure event due to the loss of the letdown flowpath.
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ACTION REQUIRED Based on the above factors, the staff believes the best course of action is to continue the current technical specification for power to be removed from the RHR MOVs during Modes 4, 5 and 6 until the low flow alarm is installed. However, the staff position that would permit the licensee to wait until the first refueling outage before installing the low flow alarm was taken over two years ago. Staff will puruse with the licensee a commitment to a schedule for accomplishink this installation at the earliest possible time.
In the interim, until the low flow alarm is installed, the staff believes that strict administrative controls should be developed and implemented to ensure that MOVS 8701 and 8702 are closed with power renoved during plant startups when RCS pressure is above the RHR design pressure.
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O . O O TO TO CONTAINMENT CHARGING SPRAY PUMPS a t d 6 TO RCS COLD LEGS : , ,
= '
1&2 i FROM
- ' AINMENT RHR HEAT 3
EXCHMGER 1 PU 1 M TO RCS ' ' ' FROM RCS u m LOOP 4
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r, 8703 870} HOT LEG g z
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TO RCS M n
I COLD LEGS C .
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- 3&4 3 '
FROM 1
"' CONTAINMENT ,
RHR HEAT SUMP EXCHANGER 2 RHR PUMP 2 4
KEY:
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NORMALLY I CLOSED TO TO SAFETY CONTAINMENT INJECTION SPRAY 1 N M VALVE TOR ERATED PUMPS f
b NORM ALLY OPEN (AT POWER)
MOTOR-OPERATED VALVE i D4 i
i AIR-OPERATED VALVE I I SCHEMATIC DIAGRAM OF M CHECK VALVE RESIDUAL HEAT REMOVAL SYSTEM
EXHIBIT 5 - TECHNICAL SPECIFICATIONS, SECTION 3.4.9.3 REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 The following overpressure protection systems shall be OPERABLE:
- a. RHR system isolation valves 8701 and 8702 open with power removed from the valve operators when the positive displacement charging pump is in operation, and
- b. Two power operated relief valves (PORVs) with a lift setting of less than or equal to 450 psig, or
- c. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2.07 square inches.
APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 323'F, MODE 5 and MODE 6 with the reactor vessel head on.
ACTION:
- a. With the positive displacement charging pump in operation with the RHR isolation valves closed, within one hour either open the RHR isolation valves or secure the positive displacement charging pump.
- b. With one PORV inoperable, restore the inoperable PORV to OPERABLE
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status within 7 days or depressurize and vent the RCS through a 2.07 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- c. With both PORVs inoperable, depressurize and vent the RCS through a 2.07 square inch vent (s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- d. In the event either the PORVs or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within
-30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient, and any corrective action necessary to prevent recurrence.
- e. The provisions of Specification 3.0.4 are not applicable.
i DIABLO CANYON - UNIT 1 3/4 4-32 l -_ _. __ _ _ _ . - _ _ _ _ _ _ . _ . _ _ _ . .. . . - - _ . _ _ - - _ _ _ . _
l REACTOR COOLANT SYSTEM i l
, SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
- a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE.
- b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months.
- c. Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection.
- d. Testing pursuant to Specification 4.0.5.
- e. Verifying the RHR isolation valves 8701 and 8702 are opened with power removed from the valve operators when the positive displacement charging pump is in operation at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4.4.9.3.2 The RCS vent (s) shall tie verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- when the vent (s) is being used for overpressure protection.
"Except when the vent pathway is provided with a valve which is locked, sealed, !
.or otherwise secured in the open position, then verify these valves open at least once per 31 days.
DIABLO CANYON - UNIT 1 3/4 4-33
-[ EXHIBIT 6
. ;, + . . . . ..
- j. PGwE PoR e8tTRA-coedPANY USSS ,
$3'u*EE SENIOR VICE PRESIDENT
% vou no me LaTesa er ev ,ser Nuclear Facilities April 29,1976 TO PGandE ENGINEERS, TECHNICIANS AND OTHERS DIRECTLY INVOLVED IN THE, COMPANY'S NUCLEAR FACILITIES -
This letter is written to the members of Pacific Gas and Electric company who are, or have been, directly involved in the design, construction, and operation of our nuclear facilities. Its purpose is to reaffirm the Company's strong cossaitment to the protection of its employees and the general public against any unsafe situation C.th respect to these facf.11 ties and, further, to assure that you have every oppor-tunity to communicate freely to your company any views you right have on the safety of nuclear facilities.
we believe that you appreciate your right and obliga-( tion to express yourselves on matters of safety and that you have the dedication and -individual initiative, insofer as your responsibilities are concerned, to see that our nuclear facilities are designed, constructed, and operated in a safe manner.
As nuclear development has accelerated there has been an intensifying discussion on nuclear issues. To give you added opportunity to ask questions or ta express your views on any aspect of the safety of nuclear facilities, including those outside your own sphere of responsibility, we encourage you not only to talk to your supervisor, but also, if you wish, co any one of the following people who have recently been designated t a review team to answer questions and to evaluate the views of I
any employee who wishes to express any concern whatever about the safety of nuclear facilities:
[
l Company Telephone l Wallace B. Allen Director, Environmental Quality Department 1675 Clinton P. Ashworth Supervising Mechanical 3305 Engineer, Mechanical &
Nuclear Engineering Dept.
Alfred W. Medcalf Steam Ginneration Engineer, 1292 e Steam Generation Dept.
,-r--.- -,,--,,,- ,,
s*
Page'Ttro . . .
April 29, 1976_ .
The individuals on this review team have broad nuclear knowledge, and they have full . latitude for drawing upon the expertise of others and stimulating action if it is necessary.
You may contact any one of them and arrange for a discussion with him alone or he will arrange for a joint meeting with other review team members. .
. tklyl-J. D. WORT 31NGTCat f.',
t &
JDW:BJD -
cc: Officers ~
Department Heads ' ' !-
Division Managers '
o
[ "L l
e'- ;* ,
U
, e 6 .
l .
1 l
l 1
, EXHIBIT 7 PACIFIC GAS AND ELECTRIC COMPANY STATION CONSTRUCTION DEPARTMENT '
DIABLO CANYON PROJECT July 25,1980 l
- MEnonANDUM TO
- DIABLO CANYON GENERAL CONSTRUCTION PERSONNEL RE: Personnel Participation l
1 To provide you with additional opportunity to participate in the Project, a new avenue of communication is now open. There are three locations where you can submit written questions and/or suggestions. Boxes have i
been bested in the following places: The Project office mail room, General Construction tool room and the Welding Crew change roca, i
You are all invited to ask any questions and to offer any suggestions you i any have regarding the Project. This includes anything from project administration, policies and rules to technical concerns such as plant design, construction and operation.
I an interested in your suggestions for improving the efficiency of our work at Diablo including safety, quality, productivity and cost control.
I as equally interested in answering any questions you any have now or later about all aspects of working on this Project or for PG&E in general.
l Your questions and suggestions will be responded to as fast as possible to you directly or on the bulletin boards if no name is giveu. If your suggestions appear to deserve consideration and/or award, via ths formal Employee Suggestion Plan you will be advised.
I as looking forward to your participation, questions and suggestions.
R. D. Et:1er Project Superintendent l
_-~___,__.,----_._-,,,_y. ,_,,.r., _ . _ , _ , - - - . - _---
l
{** ?**5
- EXHEBVT 8
..PGwE
, E9ft INTRA-COMPANY USES
,,,,,,,,,, VICE PRESIDENT g perme.seet NUCLEAR POWril CENERATION MaNo.
Rt LETTam er !
susssev Nuclear Facilities August 7, 1980 TO: PGandE ENGINEERS TECHNICIANS AND OTHERS DIRECTLY INVOLVED IN THE COMPANY'S NUCLEAR PACILITIES This letter is written to the members of the Pacific Cas and Electric Company who are, or have been, directly involved in the design, construc-tion, and operation of our nuclear facilities. Its purpose is to again l reaffirm the Company's strong cosunitment to the protection of its employees and the general public against any unsafe situation with respect to these facilities and, further, to assure that you have every opportunity to communicate freely to your Company any views you might have on the safety of nuclear facilities.
We believe that you appreciate your right and obligation to express yourselves on matters of safety and that fou have the dedication
" and individual initiative, insofar as your responsibilities are concerned,
' to see that our nuclear facilities are designed, constructed, and operated in a safe manner.
Since the Three Mile Island accident, there has been an intensi-l fying discussion on nuclear issues. To give you added opportunity to ask questions or to express your views on any aspect of the safety of nuclear facilities, including those outside your own sphere of responsibility, we encourage you not only to talk to your supervisor, but also, if you wish, to any one of the following people who have been designated a review team to answer questions and to evaluate the views of any employee who wishes to exposs any concern whatever about the safety of nuclear facilities:
Company Telephone Wallace B. Allen Director Environmental 1675 Quality Department Clinton P. Ashworth Supervising Mechanical Engineer 3305 Mechanical & Nuclear Engineering Department Alfred W. Medcalf Sr. Nuclear Generation Engineer 1292 Nuclear plant Operations
- Department
1 .
F:33 2 August 7. 1980 t
The individuals on this review team have broad nuclear knowledge, and they have full latitude for drawing upon the expertise of others and stimulating action if it is necessary. You may contact any one of them and arrange for a discussion with him alone or he will arrange for a joint meeting with other review team members.
J. O. 7CHUYLER JOS(3096):ar cc: Officers Department Beads Division Managers i
l e
n
4 ... .. i v m i EXHIBIT 9
. POwE ron mena-cowamm
,,,,,, o,,,,,,,,, NUCLEAR PLANT OPERATIONS one-w=at Diablo[anyonPowerPlant FtLa No. . i Rt Laffte or suescer Adherence to NRC Regulations .
Open Door Policy REVISED FEBRUARY 2, 1984 UT," PGandE Quality Hotline February 5, 1982 TO ALL EMPLOYEES:
Since we have so many new employees at Diablo Canyon, it is a good time to reiterate PGandE's policy concerning adherence to governmental rules and regulations. All employees are urged to follow approved plant procedures and plant technical specifications. In doing so, the plant staff will neet all governmental rules and regulations, including those issued by the Nuclear Regulatory Commission (NRC).
Should any employee feel that such rules or regulations are not being followed, he should pursue the matter with his supervisor.
If the matter cannot be resolved through his supervisor, he should pursue it via the "Open Door Policy", which has been previously stated and is repeated on the following page.
Should any employee feel that NRC rules or regulations are not being followed and plant management is not correcting the situation, NRC regulations allow him the right to discuss the problem with NRC inspectors (see NRC-3 notice posted on the bulletin board). Two NRC inspectors, Marvin Mendonca and Mark Padovan, are located in the office trailer adjacent to the Administration Building (plant extension 2439). Should an employee prefer not to talk to the onsite NRC inspectors, the Region V NRC office address and tele-phone number is:
U.S. Nuclear Regulatory Commission Region V 1450 Maria Lane Walnut Creek, CA 94596-5368 (415) 943-3700 This memorandum does not authorize any employee to leave his assigned job or interrupt his work assignment to meet with NRC inspectors during work hours. It is preferred that, if necessary, employees contact NRC inspectors during nonwork hours. However, if a meeting must take place during work hours, an employee may request permission from his super-visor. Any employee who requests time to meet with NRC inspectors will be granted such time off at a scheduled, convenient time.
k
TO ALL EMPLOYEES February 5. 1982 OPEN DOOR POLICY I would like to take this opportunity to reiterate our policy concerning the employee's right and obligation to inform manage-ment of safety concerns. Any employee who feels he has identified a concern or problem affecting plant safety or reliability should discuss this matter with his immediate supervisor. Following such a discussion, if the employee feels that his concern is not being adequately addressed, he should pursue the matter with progressively higher levels of supervision, up to and including the Plant Manager. '
While it is desirable to keep minor complaints from reaching this level of management, I would like all employees to feel free to approach the Plant Superintendent and the Plant Manager with his concern if he feels it is necessary.
Voicing concerns regarding plant safety or operation will not be documented in the employee's personnel record and will never be used for any type of disciplinary action. I give my firm guarantee of that. .
This policy also applies to any situation in which the employee feels he is being required to do any work which would jeopardize his personal safety.
PGandE QUALITY HOTLINE I wish to call attention to signs which have been posted around the plant and project concerning the " Quality Hotline". This is a PGandE hotline and NPO employees are urged to use it if they deem it necessary. The hotline number is plant extension 3567 (541-7567 if calling from a pubite or outside telephone).
SOf R. C. THORNBERRY t
RCT(3350):ws xc RDEtzler LERosetta JDShiffer
' ~
enNS=$wv oens
- I g PACIFIC GAS AND F.I.ECTRIC COMPANY l orracm or raz cRAIRMAN I
March 22, 1982 ;
i
, TO: PGandE OFFICERS, ENGINEERS, TECHNICIANS AND OTHERS DIRECTLY INVOLVED IN THE COMPANY'S NUCLEAR FACILITIES This letter is to reemphasize the Company's long- I standing commitment ,to design, build, and operate safe nuclear power plants and in achieving this commitment to require all employees to practice fundamental honesty and to adhere to Nuclear Regulatory Commission ("NRC") rules and regulations.
This is.also to reemphasize that our communications -
l with the NRC must be open and allow a free flow of information . We must be ever alert to any possible misleading or ambiguous statements made either in oral or wfitten communications. Any such misstatements must be corrected immediately upon discovery. Nothing less than full and open communication between the Compainy and the NRC can be tolerated.
In October 1977, PGandE formalized its general policy concerning employee conduct (Standard Practice 735.6-1).
The statement of policy establishes a Company philosophy regarding work conduct emphasizing that: .
l "It is the policy of this Company that '
l employees shall at all times continue to practice fundamental honesty. Employees shall not, nor attempt to: deceive, defraud, or mislead the Company, other employees, or those with whom the l Company har business or other relationship; ... misrepresent the Company or its employees; ... withhold
, their best efforts to perform their work l ~
to acceptable standards; ... violate applicable laws; or conduct themselves
! at any time dishonestly or in a manner which would reflect discredit on the Company."
This policy is particularly important to.all employees engaged in work concerning nuclear power.
4
,rf-- r n- anr,- , ema,,, w--,--,-,---o,------m-- , -- , , . - - ~ , - - - - - - - , - - - - - - , - , - + - - - - - - - - - ~ ~ - - - - - - - - - , ~ - - -
To All AddrossGd -
2- March 22, 1982 4
l In April 1976, Mr. J. D. Worthington, and again in 1980, Mr. J. C. Schuyler, issued a memorandum to all personnel involved in the Company's nuclear power work which described a program to permit such personnel to discuss their concerns regarding nuclear power. The August 1980 -
letter stated that:
"[our] purpose is to again reaffirm the Company's strong commitment to the protection of its employees and the general public against any unsafe situation with respect to these nuclear f acilities and, further, to assure that i you have every opportunity to communicate freely to your Company any views you might have on the safety of !
nuclear facilities.
"We believe that you appreciate your right and obligation to exprass yourselves on matters of safety and that you have the dedication and individual initiative, insofar as your responsibilities are concerned, to see that our nuclear facilities are .-
l designed, constructed, and operated in a safe manner. *
"To give you added opportunity to ask questions or to express your viewt on any aspect of the safety of nuclear facilities, including those outside your own sphere of responsibility, we encourage you not only to talk to your supervisor, but also, if you wish, to any one of the following people who have been designated a review team to answer questions and to evaluate the views of any employee who wishes to express any l concern whatever about the safety of nuclear facilities:"
We are proud that the application of these policies of l ~
openness in finding and evaluating safety. issues led l directly to the discovery by PGandE personnel of the " mirror image" error at Diablo that otherwise might have gone l undetected.
S
1 To All Addrococd M6rch 22, 1982 i Recently, in February of this year, Mr. R. C.
Thornberry issued a separate memorandum to Diablo Canyon Power Plant employees which reiterated the Company's policy concerning adherence to government rules and regulations.
We must strive for perfection in design, construction, and operation of ocr nuclear ~ units. To attain this goal, it i
is necessary that we all exercise our best efforts to resolve problems we encounter in our work. When problems are encountered, they must be immediately identified, clearly defined, and brought to the attention of your supervisor. This approach should facilitate the evaluation of, and formulation of timely and effective solutions to, any problem. Constructive recommendations are encouraged at all levels.
Our goal is to design, construct, and operate our nuclear facilities with full margins of safety and full compliance with NRC requirements. Strict adherence to the above policies will provide added assurance that this goal .
will be met.
t .
F. W. MIELKE, r. . B. W. SHACKELFO cc: Officers
, Department Heads -
l Division Managers All Concerned Personnel I
t i
l
, o',< O '
O navc A QdAUTY CONCERN? HOTLINE.
i Checklist for using if you have a V 4 the HOTUNE qmlity concem-1 ,
When calling the HOTUNE, please i be prepared to provide the following *
, he answering service recorder is ~
- activated wtan you call, plve the date 4
and tinw of your call, TELL YOUR SUPERVISOR.
l GGive as cornplete a description of the quality concem as you can, for l 1natance: N o R e s ufI s . . . .
i . " " " * *
+
- e as /
TELL YOUR MANAGEMENT 1 -
- c ,i e j ,,,,,, '
m Still no results -
= w u nn ms g
- l. p o.a ami. USE THE HOTLINE j a., u e. w m 4
see=us me s a OR STOP BY THE HOTLINE OFFICE 3 l Il possible, Identify individuals who may be able to provide additional or '
PG&E EXTENSION corroborative it: formation.
12I Give other methods by which the concern was expressed previously.
(Es s one y sw y ,e .,
3567
,j -i w w. . e
! E OPTIONAL Identify yourself and a PUBUC PHONE m.ihod by which w. can r.ach vaa i
a) to obtain rnore information or b) to provide results of the Investigation 541 - 7567 l to you.
Alternately,if you choose not to identify yourself, you rnny phone the HOTLINE again !n several weeks during business hours and inquire as to the status of the quality concern.
j NOTE: f.very effort will be made to protect the identity of those callers h
who identify themselves.
.'69-018 (7/01) EMBU 12 ,
w NUMBER FAP C-12 !
pggg Pacifi 8 sadElectricCs pss! uvlsion 89W-122 i I
~
NUCLEAR PLANT OPERATIONS DATE 6/15/83 DEPARTMENT PAGE'1 0F 42 TITLE OVED Y 8 IDENTIFICATION AND RESOLUTION OF PROBLEMS AND N0HCONFORMANCES MANAGER I. SCOPE This procedure describes the Nuclear Plant Problem Report (NPPR) system, the General Office Problem Report (GOPR) System and the Nonconfomance Report (NCR) system utilized for the identification of equioment, material, procedural, or other problems and nonconfomances which occur at or affect a nuclear facility. The requirements for identification, documentation, routing, evaluatim and determination of cause, msolution, and corrective action are also discussed as they relate to problems and Nonconfonnances.
In this procedure the major factors involved in identifying, re:olving and reporting problems are discussed in general terms, and detailed instructions and suggestions for completing and handling the forms are provided.
II. DISCUSSION The NPPR, GOPR, and NCR systems are used to ensure that conditions adverse to plant safety, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, abnormal occurrences, discrepancies in work activities, and nonconfomances are promptly identified, evaluated and corrected. The systems also ensure that the appropriate levels of management are notified and that independent myiews are conducted as required. -
These systems also satisfy several other objectives, including:
A. Document significant problems involving only one department.
B. Document prob 1ces which are not necessarily directly involved with operating plant equipment, such as procedural problems and spare parts problems.
C. Document the detennination of the cause and the specification of the corrective action (steps to prevent recurrence) as part of the myiew and resolution of major problems.
D. Provide a reminder that many major problems require mporting to the Mtt, and provide the required fomal procedure specified in 10CFR21 to assure that such reporting is acconplished within established time limits.
5PG01 1
, CM2B (7/E1)
C'UCLEAR PLANT OPERATIONS DEPARTMFNT NUMBER IFAP C-12 .
NUCLEAR PLANT ADMINISTRATIVE PROCEDURE 112 .
REVIS10h IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 2 0F 42 l E. '
Provide formal documentation of the final resolution of the Technical Review Emup for most major problems, i F.
- Provide for initiation DataSystem(NPRDS). of reporting to the Nuclear Plant Reliability The NPPR and GOPR systems also serve as a written method for requesting intradepartmental thus insur'ng that:work on any item, whether it is safety related or not, A.
There requested.
is no misunderstanding as to precisely what is being B. Scheduling and planning of work is facilitated.
C.
Affected departments can keep track of whether the work has or has not been done.
D. A record of the work performed is provided.
1 E.
A mechanism is provided for keeping management informed of the plant status and department activities. ,-
~
F. Information' is provided on completed werk.
The NPPR, GDPR, and NCR systems are implemented respectively by the use of Form 61-4516. the " Nuclear Plant Problem Report". Form 69-025 the
" General Office Problem Report", and Fors76-286 the "Nonconformance Report." The NPPR and GDPR forms describe the problem, document preliminary, decisions which are made regarding its significance, serve as a work request, and document the basic resolution of the problem. The ICR form documents the findings 'of ths Technical Review Group (s) and provides for follow-up on items which are judged to be nonconformances.
II
I. PROCEDURE
g A. Nuclear Plant Problem' Reports (NPPR)
- 1. General Requirements for Initiation of Nuclear Plant Problem Reports, Appendix C.
- n. Each plant and/or department may specify criteria for the initiation of Problem Reports subject to the following minimum requirteents:
- 1) Any malfunction of permanently installed plant l
systems or equipment important to safety..
- 2) Any item which must be reported as a Nonconformance (see Section C of this procedure), because NPPRs
_ initiate NCRs.
MPG 01 2
,.,,-,.._.--__,.A_~._. _ . . , . , . , , . 1.*._....~._,-,_-w,,...,. , , _ - , , _ , . . , , , , , ,%. - , . _ , , , _ . - - . , . _~,--w,,--,_---,,_.,-_,,,._,,,-.,m- ,
> . O-02 8 (7/h) .
NUCLEAR PLANT OPERATH)NS DEPARTMENT NUMBER IFAP C-12
~
l NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS10g0A W-112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 1
TITLE PROBLEMS AND NONCONFORMANCES PAGE 3 OF 42
- 3) Any incident involving equipment defects,'
malfunctions, personnel overexposure, effluent release in excess of limits, or other similar
, occurrences which represents a violation of established procedures and/or may require a special i report to the NRC (see NPAP C-11/60AP W-111). .
- b. When a plant is under construction or is undergoing major modifications, such that the plant staff is interfacing extensively with outside groups (engineering, construction, consultants, contractors), it is permissible for the various parties involved to adopt mutually agreeable alternative problem reporting methods which meet or exceed the requirements of this procedure. Such alternative methods shall be described in appropriate Temporary Procedures, or in the QA/QC manual, for the outside group.
- c. Any NPO individual who identifies a plant problem may initiate a NPPR, but as a minimum shall notify supervision (preferably his immediate supervisor). The supervisor
(., notified is responsible for notifying the supervisor in 1
i L charge of the work, if other than himself, and for
- l assuring that a NPPR is initiated if twquired. '
NPO-General Office initiated NPPRs shall be coordinated l with and transmitted to the Plant Superintendent for I further processing. l
- 2. Guidelines for Conducting Required Work
- a. Occasionally, plant management has to correct an item innediately without waiting for submittal and approval of a NPPR. An example of this type of a situation would be finding a protection channel out of calibration in the unsafe direction, which is a reportable occurrence.
Obviously, the instrument should be innediately recalibrated if possible. The NPPR would still be prepared for the purpose of informing plant management and others of what had transpired, and giving them a chance to approve the accomplished work and decide whether additional corrective action is appropriate. Although the work may be complete, the problem is not officially resolved until all sign offs are complete.
- b. Except as prohibited by 2.c. below, the decision as to whether an activity should be performed pending submittal of a NPPR is left to the supervisor in charge of the work. ,
, c. Items which shall not be conducted prior to receipt of plant management approval include:
NPG01 3
l * . C ,-018 (7 M1)
NUCLEAR PLANT OPERATIONS DEPARTMNT NUMBER NN 0-12, NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIO i IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES lPAGE 4 0F 42
- 1) Activities which may involve an unrtviewe'd safety question as defined in 10CFR50.59, or a change in the Technical Specifications or Final Safety Analysis
, Report. An activity is considered to ir.volve an unreviewed safety question if:
a) the probability of occurrence or the consequences of an cccident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or b) a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or c) if the margin of safety defined in the basis for any technical specification is reduced.
- 2) Activities which involve a design change.
- 3) Placing nonconfoming spare parts and materials in '
service without the appropriate reviews.
- 3. Design and/or Procedure Changes If the work involved on a NPPR results in a design and/or a procedure change, the administrative controle described in Nuclear Plant Administrative Procedures C-1 and E-4 shall be followed.
B. General Office Problem Reports (GOPR)
- 1. General Requirements for Initiation of General Office Problem Reports, Appendix D.
- a. Each group within Nuclear Plant Operations may specify criteria for the initiation of Problem Reports subject to the following minimum requirements:
- 1) Any General Office deviation and/or deficiency in work activities from approved procedures or from department implementing instructions.
- 2) Any item which is discovered which may require a special report to the NRC (see NPAP C-11/GOAP W-131).
- 3) Any item which must be Nported as a,Nonconfomance (see Section C of this procedure), because GOPRs initiate NCRs.
NPG01 4
l O-018 (7/h)
WUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER fFAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE 112 REVISIO IDENTIFICATION AND RESOLUTION OF DATE 6/B/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 5 0F 42
- b. When a plant is under construction or is underhoing major undifications, such that the General Office staff is interfacing extensively with outside groups (engineering,
, construction, consultants, centractors) it is permissible for the various parties involved to adopt EJtually
- agreeable alternative problem reporting methods which meet l
or exceed the requirements of this procedure. Such alternative methods shall be described in appropriate Temporary Procedures, or in the QA/QC manual for the outside group.
- c. Any NPO individual who identifies a General Office problem may initiate a 60PR but as a minimum shall notify supervision (preferably his innediate supervisor). The supervisor notified is responsible for notifying the supervisor in charge of the work. if other than himself, and for seeing that a G0PR is initiated. NPO-plant initiated 60 prs shall be transmitted to NPO-QC for further processing. NPO plant initiated NPPRs requiring General Office action shall be transmitted to the responsible
., Supervising Engineer for further action.
- 2. Guidelines for Reviewing 60 prs Supervisors are responsible for reviewing and conducting additional investigations as necessary to make one of the following determinations:
- a. The problem could be a potential Nanconformance as defined in Section C, unless a Technical Review Group has already detennined ctherwise. If it is a potential Nonconformance, the responsible supervisor shall process the matter according to Section C of this Procedure.
- b. The problem meets the requirements for a Problem Report as defined herein, in which case the supervisor shall process the problem in accordance with this Procedure.
- c. The problem is neither of the above. The supervisor shall dismiss the matter or handle it through nonnal work practices. No additional documentation is required. For example, spelling errors, and other clerical errors in documents which clearly do not affect the obvious intent of the document should not be considered discrepancies; but they should obviously be corrected.
O
. C%018 (7/61)
C:UCLEAR PLANT OPERATIONS DEPARTMENT NUMBER fFAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION W-112
! IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES 6 PAGE 0F 42
- 3. Design and/or Procedure Changes
- If the work involved on a GOPR results in a design and/or
- procedure change, the administrative controls described in the appropriate General Office Adninistrative Procedure shall be followed.
C. Nonconformances (NCR)
- 1. The Company's Quality Assurance Pmgram (QAP 10.1) requires that "Nonconformances" be reported in writing and resolved in a formal manner with appropriate management review. The administrative steps involved in the identification and msolution of Nanconformances. Appendix E, are sunnarized below.
- a. Responsibility for Classifying Problems as Nanconformances The Manager, Nuclear Plant Operations has delegated the responsibilities for determining whether a problem identified in a 60PR is a Nonconfonaance to individual General Office Supervisoring and Senior Engineers. The t Plant Manager 1 and plant department heads have been delegated the responsibility for determining whether a problem identified in a NPPR is a Nonconformance. In addition, other individuals may be delegated this rwsponsibility, provided this delegation is in writing.
When such an item is identified it is the responsibility of that individual to initiate an NCR.
- b. Criteria for Classifying Problems as Nonconformances
- 1) Definition A Nanconformance, as defined in Quality Assurance Procedure (QAP) 10.1, is a discrepancy or departure from requirements in purchase specifications, drawings, approved practices, established Quality Assurance policies or procedures, or NRC regulations which requires resolution and may require measures to be taken to prevent its recurrence. Nonconfonnances consist of conditions adverse to quality which, if left vncorrected, could have resulted in any of the following:
1 For those plants without a Plant Manager, all responsibilities assigned to this position are transferred to the' Plant Superintendent.
W - - - - - - - --
MPG 01 6
. . C.-ola (7/61) t WUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER IFAe C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REV1510t[.0A6 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 7 0F 42 a) Degradation or loss of the integrity of the reactor coolant pressure boundary;
- b) Reduction or loss of the capability to shutdown the reactor and maintain it in a safe condition, including the compraise of design objectives during construction and/or modification activities; or c) Lack of effective control over items or activities (including quality program implementation) that could reduce the capability to prevent or mitigate the consequences of
. accidents that may result in potential off-site exposures comparable to the guidelines set forth in 10CFR100. " Reactor Site Criteria."
- 2) Examples of Nonconformances Appendix A is a compilation of examples to aid in determining if a problem or potential problem is a
{ c.
Nonconformance.
Role of Technical Review Gmup in Resolution of i Nonconformances In accordance with Quality Assurance Procedure 10.1. once a problem has been classified as a Nonconformance by one of the individuals assigned such authority per paragraph C.I.a. above, a Technical Review Group must be convened within 30 Calendar days to review the Nonconformance, detersinc its cause, approve the proposed resolution and ,
establish the corrective action to prevent recurrence. '
Appendix B discusses the makeup and functioning of the Technical Review Gmup.
- d. Role of Plant Staff Review Comittee in Resolution of Nonconformances
- 1) The Plant Staff Review Comittee shall review, at least monthly, the status of all NCRs that were issued on-site or which affect on-site safety-related items or activities. In most cases, this review will ,
take place after the Nonconformance has been l
resolved. When a Nonconformance Report involves a '
system design change it shall be reviewed b the PSRC prior to reliance upon the affected system (y)s for a safety-related function. (Refer to Nuclear Plant Administrative Procedure C-1).
NPG01 7
. O,-018 (7/61)
C UCLEAR PLANT OPERATIONS DEPARTMENT NUMBER fFAP C-12 NUCLEAR E. ANT ADMINISTRATIVE PROCEDURE -112 REVIS10 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 8 0F 42 I
The Secretary of the PSRC should prepare a short I
' summary of ea:h NCR which was resolved during the month and include it in the announcement package for
, each regularly seneduled monthly meeting. The NCR shall be reviewed as part of the PSRC's normal review of operating, maintenance, and surveillance testing experience.
- 2) The Technical representative on the Technical Review Group is responsible for assuring that NCRs are referred to the PSRC(secretary) prior to implementation of the resolution where such referral is required or desirable.
- 2. Determination of Reportability When a problem arises,'it is extremely important that knowledgeable individuals promptly evaluate the problem and determine whether or not it is reportable to the NRC or other government agencies; and, if it is, assure that such reports are made in a tinly manner. The initial evaluation is R performed by the appropriate department head or other assigned 4 individual and if the problem is considered to be reportable per the guidelines of NPAP C-11/60AP W-111, then that incividual or department head is responsible for convening a Technical Review Group meeting, if time permits. It is
' emphasized that in some cases the initial verbal report may have to be made so quickly (1 haur or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) that it may not be practical to convene a TRG. In these cases, the department head should promptly notify the Plant Manager and/or Manager of Nuclear Plant Operations, who will then make the verbal notification. The function of the TRG in such cases is to tuview the occurrent.e and assume that followup written reports are made.
I According to Standard Practice 420.3-1, each site shall establish a Site Review Group for the purpose of investigating potentially reportable items. This Site Review Group shall be the Plant Staff Review Connittee or a designated subcomittee thereof. Since most reportable items also involve a Nonconformance, it is most convenient to consider the Technical Review Group which investigates the Nonconformance a subcomittee of the PSRC for purposes of making the initial determination of its reportability. Therefore, the membership and functioning of the Site Review Group shall be as described for the Technical Review Group in Part C.I.c above and in Appendix B. '
NPT,01 8 l
' C.-028 (7/h )
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIOg0 W112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 9 0F 42
- 3. Audits
- a. Nonconfonnances The Quality Assurance Department, as stated in QAP 10.1, shall maintain a listing of all nonconfonnances, issue periodic reports of NCR status, and auditing the nonconformance report system annually for trends or other indications that practices should be improved. The Quality Assurance Department is responsible for issuing a report of such annual audits which sets forth the findings and recommendations.
- b. Reportable Items The General Office Nuclear Plant Review and Audit Connittee (GONPRAC) is responsible for reviewing the implementation of Standard Practice 420.3-1 at least annually.
- D. General Requirements
- 1. Problem Report Resolution Process and Instructions for
. Completing and Handling NPPRs, GOPRs and NCRs The NPPR, 60PR, and NCR forms have been developed so that they follow the general processes for hantiling problems - from discovery through resolution and signoff. Appendices C, D, and E to this procedure provide general instructions for completing these forms. Appendices D and E also give general instructions for routing of 60 prs and NCRs. Because of the differences in l the organizations of the individual plants, each plant shall
- prepare a supplementary procedure specifying how NPPRs are l handled at the plant. Also, the suggested general routing instructions for GOPRs and NCRs may be altered if desired by preparation of appropriate supplementary procedures.
- 2. Maintenance of Records j a. Records which are related to an actual, substantial safety hazard shall be retained for the life of the affected basic components or the affected licensed facilities.
These records include:
- 1) Records of the activities and meetings of review groups.
- 2) HRC notifications.
- 3) Documented reviews and evaluations.
I ,
NDCA1 0
. O-018 (7/6.1)
CDUCLEAR PLANT OPERATIONS DEPARTMENT ' NUMBER NN c-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE ~1 REVISIO h IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 10 or 42 l
I l b. Problem Reports which involve equipment important to i
safety shall be retained for five years.
- c. Preblem Reports which do no't involve equipment important to safety should be retained fer one yeat.
V. REFERENCES A. ANSI /ANS 3.2: " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants."
B. PGandE Standard Practice 420.3-1: " Defects and Noncompliance:
Reporting of to the Nuclear Regulatory Comission."
C. Quality Assuran'ce Procedure 10.1: "Nonconformances and Corrective Actions."
D. Quality Assurance Pncedure 10.2: " Verification of Reporting of Defects and Noncompliances."
E. Title 10, Code of Federal Regulations. Part 21: " Reporting of Defects and Noncompliances."
F. Title 10, Code of Federal Regulations, Part 50.72: " Notification of Significant Events."
G. NRC Regulatory Guide 1.16. Revision 4 August 1975: " Reporting of Operating Information - Appendix A Technical Specifications."
H. NRC Regulatory Guide 10.1, Revision 4. October 1981: " Compilation of Reporting Requirements for Persons Subject to NRC Regulations."
NUREG-0302, Revision.1:-* Remarks PresentedTQ0estions/ Answers I. --
Discuss'ed) At Public Regional ~ Meetfn-gsTo 1)iscuss Regulations (10CFR21) for Reporting of Defects and Noncompliance."
J. NPAP C-11/GOAP W-111. "Nonroutine Notification and Reporting to the Nuclear Regulatory Commission (NRC) and other governmental agencies."
K. Nuclear Plant Adninistrative Procedure C-1, " Design Changes."
VI. APPENDICES A. Appendix A - Examples to Aid in Determining if a Problem or Potential Problem is a Nonconformance.
B. Appendix B - Makeup and Function of a Technical Review Group in the Restoration of Nonconformances.
_ t NPG01 10 f
1
O-038 (7/61)
C'UCLEAR PLANT OPERATIONS DEPARTMENT NUP BER PFAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE 112 REV1510 IDENTIFICATION AND RESOLUTION OF DATl; 6/15/83
- TITLE PROBLEMS AND NONCONFORMANCES PAG'E 11 0F 42 I
C. I Appendix C - Description of Nuclear Plant Problem Report Resolution 1 Process and Instructions for Completing and Handling Form 61-4516.
. D. Appendix D - Description of General O'ffice Problem Report Resolution Process and Instruction for Completing and Handling Fonn 69-025.
E. Appendix E - Description of Nonconformance Report Resolution Process and Instruction for Completing and Handling Fonn 76-286.
VII. ATTACHMENTS A. Form 61-4516: " Nuclear Plant Pmblem Report."
B. Form 69-025: " General Office Problem Rcport."
C. Fonn 76-286: "Nonconformance Report" l
C 1
l
~
9
, _ _ ----w- - - -
NPG01 11
, 0-028 (7/61)
NUCLEAR PLANT OPERATIONS DEPARTn4ENT NUMBER NPAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REyls!O W-112 DATE 6/15/83 IDENTIFICATION AND RESOLUTION OF TITLE PROBLEMS AND NONCONFORMANCES PAGE 12 0F 42 APPENDIX A -
Examples to Aid in Detemining if a Problem or Potential Problem is a Nonconfomance Nonconformances represent problems or potential problems in items or !
activities important to safety as defined in this procedure. Nonconfomances ;
may include physical defects, test failures, incorrect or inadequate '
documentation, or deviations from prescribed procedures (i.e. inspection, test l or processing).
The following are some examples to illustrate when an item should or should not be considered as a Nonconformance. These examples should not be considered as all inclusive:
A. Procedural Problems l 1. While performing a procedure, a person identifies some changes which will make it more workable. A temporary change is issued pending a revision. No Nonconformance is involved. Such changes would be considered as part of the normal review process. In fact, this item l need not even be reported on a Pmblem Report inasmuch as the l 1 temporary change sheet provides a written reminder to plant w u management that the procedure requires revision,
- 2. While performing a calibration check on a reactor trip circuit it is noted that due to an error in the procedure. the instrument has had
!- an incorrect setpoint for several months. A definite Nonconformance is involved which is indicative of a weakness in the procedure l review process and represents a " lack of effective control" over activities important to safety.
- 3. Repeated failure to follow approved procedures or to provide required doctanentation after a discrepancy or problem has been identified and reported qualifies as a Nonconfomance because it represents a " lack of effective control".
B. Maintenance Problems
- 1. Normal repairs involving expected deterioration or wear need not be considered as Nonconformances but such repairs shall be documented in maintenance records.
l 2. During routine testing or preventive maintenance on a piece of equipment important to safety which was operable at the time it was cleared, component parts are found to be deteriorated but still operable, and are mplaced. No Nonconformance is involved unless the problem is an unexpected one which indicates a basic design -
deficiency or unexpected wear rather than the normal wear and tear.
Basically, in this example the testing and preventive maintenance NEG01 12
.- . . . - . ._ . _ _ =
l
. C%318 (7/h)
UCLEAR PLANT OPERATIONS DEPARTMENT NUMBER MPAP C 12 i
~
NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION
. IDENTIFICATION AND RESOLUTION OF DATE 6/15/83
- TITLE PROBLEMS AND NONCONFORMANCES PAGE 13 0F 42
~
APPENDIX A (Continued) program has done its job - i.e., discovered a deficiency before it
, produces a failure. If the parts had deteriorated to a point that the equipment was inoperable, a Nonconfomance would exist because a reduction in safeguards capability was involved (see paragraph C.1.b.1)b).
- 3. A piece of safeguards equipment fails in service, but is reported and repaired within the allowable license interval. A Nonconfomance is involved if the equipment would not have done its job, because the problem may be indicative of a deficient testing or preventive maintenance program (i.e. lack of effective control as per C.1.b.1?c) of this procedure), and a reduction in safeguards capability (C.1.b.1)b) of this procedure).
C. Test Equipment Problems
- 1. A gauge is marked as requiring calibration immediately prior to each use, and on one occasion is found to be out of calibration. No Nonconfomance is involved because the calibration program was
/' adequate to identify the problem in time to avoid using an L instrument which is out of calibration and there was ho lack of effective control.
- 2. Test equipment (including transfer standards) which has been used for quantitative work on equipment important to safety is found out of calibration to the extent that subsequent results and/or calibrations are questionable. A nonconformance is involved which indicates lack of effective control and which may require shortening the frequency or reevaluating the accuracy specifications. There l are many instruments which are calibrated routinely which are not so used.
D. Material Problems
- 1. A spare part for a system important to safety is received and is j rejected during receipt inspection. No Nonconformance is involved -
the inspection program has done its job. The problem would be documented on the standard receipt inspection form and the defective item should normally be returned to the supplier as part of the procurement process.
A " defect" of this nature, if it could create a substantial safety hazard if used, may be reportable to the NRC under the provisions of 10CFR21 (see NPAP C-11/GOAP W-111) although most spare parts are
" commercial grade" items which are exempt from this regulation unless they have been " dedicated" for use as a " basic component."
However, as long as the part is returned to the supplier, the NPG01 13
_ - ~ . _ . _ . _ _ . _ _ . . _ _ _ _ _ _ .
y y-M8 (7/E1) r NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER AP C 2 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS10[ 6 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROB; EMS AND NONCONFORMANCES PAGE 14 0F 42 APPENDIX A (Continued) supplier is responsible for making this report, if one is required.
- On the other hand, if the defective item is not returned to the supplier then an NCR is initiated (i.e. it is repaired on site for subsequent use), and the reporting burden shifts to the purchaser -
(Company).
If the received item is obviously defective or incorrect and is so identified by the materials facility prior to conduct of a QC receipt inspection, the item may be returned as part of the procurement process without filling out either an inspectior report or a problem report (i.e. the item is, for all practica' purposes, not considered to have been received for possible use). However, a defective item must still be returned or discarded in order to eliminate the requirement for filing a report under 10CFR21.
- 2. A spare part is received without adequate documentation, and it is necessary to use the part anyway. A Problem Report should be filed and the item classed as a Nonconfonnance since this condition creates a deviation from the purchase specifications and it requires review and approval, s
- 3. A spare part is received but prior to receiving inspection adequate documentation or some other part of the order is found missing, and the part is not needed to be used soon. The part may be tagged and .-
placed in the QC hold area until the missing items are received. No Nonconfonnance is involved, nor should a Problem Report be filed.
Obtaining the missing items is a part of the normal procurement process. If the part was subjected to the receipt inspection and approved and later found to be defective because something was missing, a Nonconformance would be involved.
- 4. A part degrades during storage to a point where it would no longer perform its function, and it is discovered during the established system for inspecting pe*ts in storage prior to the time it is installed. A Problem Report must be filed (see Nuclear Plant Administrative Procedure D-503). The problem may or may not be a .
Nonconformance, depending upon the circumstances. If there is reasonable assurance that the defect would have been discovered ,
prior to use by either the materials facility or the group which installs the part, then there has been no loss of control and no Honconformance is involved. On the other hand, if it was only through good fortune that the defective part was not placed in -
service (i.e. effective control was lost de facto), or if the deficiency appears to be indicative of a substantial breakdown in established quality control procedures for conduct of the materials facility operations, then a Nonconfonnance is involved. ~
NPG01 14 l'h I
i l
O-MB Pt/u) !
eduCLEAR PLANT OPERATIONS DEPARTMENT N M ER W^
NUCLEAR PLANT ADMINISTRATIVE PROCEDURE O 12 REVISION 6
~
l IDENTIFICATION AND RESOLUTION OF DATE 6/15/83
, TITLE PROBLEMS AND NONCONFORMANCES PAGE 15 0F 42 APPENDIX A (Continued)
E. Design Problems A discrepancy or departure from requirements in design documents or activities shall be identified and documented as a Nonconformance if:
- 1. It was identified after the design document had been approved and it was issued for use as a basis for further design; or l
- 2. There was a basic design error meetin I Nonconformance (not a design change) in agdesign the definition documentofwhich a was approved and issued for use in construction or modification of j existing facilities; or
- 3. There was a failure to confom to the approved procedural and quality program requirements which were comitted to in the Safety Analysis Report, Construction Permit, requirements issued by the ,
Nuclear Regulatory Comission subsequent to the Construction Pemit, '
or Operating License.
F. Miscellaneous
. 1.. Discrepancies or problems of a relatively insignificant nature but
(- which, due to their repetition, require action by management should be considered as Nonconformances.
- 2. Repeated failure to correct by the mutually agree-upon comitment date those discrepancies or departures which were identified in audits, surveillance reports, or inspections, when such a delay is determined to have a significant impact on quality, should be considered as a Nonconformance.
- 3. If a deficiency in an item is noted at an operating plant during an inspection performed to identify such deficiencies so as to allow t
correction prior to relying on the item to perform its safety function, a Nonconformance Report may not be required.
- 4. If a deficiency is noted during a construction process and checking <
and correcting is part of the routine normal course of work prior to sign-off and acceptance, it need not be reported as a Nonconfomance.
, 5. When administrative, test, maintenance, or other such procedures identify items which must be classified as Nonconformances, these t
procedures govern.
L 6. In most, but not all, cases items which requim special reporting to the NRC will involve a Nonconfomance. Appendix 1 to NPAP C-11/GOAP W-111 provides a sumary of all the generic items which require a special report to the NRC, and identifies those which should be documented on a Nonconformance Report.
O NPG01 15
0 -032 (7/61 )
NUCLEAR PLANT OPERAYlONS DEPARTMENT NUMBER NPAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIO 112 l IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 16 0F 42 APPENDIX B Makeup and Function of a Technical Review Group in the Resolution on Nonconformances.
i This Appendix describes the makeup and functioning of a Technical Review Group 1
in the resolution of Nonconfonnances.
A. Membership
- 1. The Technical Review Group shall, as a minimum consist of the following personnel:
- a. Technical Representative (Chairman)
- 1) The technical representative shall be or be selected by the department head of the affected department (i.e.
normally the individual who made the determination that the problem was a Nonconformance) or either the Manager, Nuclear Plant Operations, or the Plant Manager /
Superintendent. The technical representative shall be a supervisor responsible for or affected by the item or >
activity. The department head may choose to appoint more than one technical representative to the group, in which I case he will select one to serve as chairmen.
- 2) For all NRCs at a site, regardless of the organization which initiates them, the TRG shall be chaired by a technical representative from NPO.
- b. Quality Assurance Engineer or authorized delegate.
In the event that the Quality Assurance Engineer or his authorized delegate (s) is not onsite, the review group may meet without him and obtain his concurrence orally. The oral approval shall be noted in writing in the review group's written decision.
- c. Quality Control Representative.
- 2. Others should be included in the Technical Review Group to assure j
representation by all affected departments (i.e.. those affected by l
the NCR and those who may be affected or participate in deterniining l
the cause, resolution or corrective action.)
l a. A representative of NP0 shall be included on each TRG involving a nuclear plant.
1 i
j NPG01 16
. sNolB (7/E1) i
~
C'UCLEAR PLANT OPERATIONS DEPARTMENT NUMBER FAP C-12 NUCLEAR PLANT ADMINISTRATIVE PP.0CEDURE GOA REVISION 6 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 17 0F 42 APPENDIX B (Continued) -
- b. For the duration of the Diablo Canyon Project, a Project
- representative should be included on the TRG if the item involves the project or project schedule.
B. Conduct of Meetings
- 1. The time and location of the meetings shall be determined by the Chairman. Meetings should be timely, consistent with the significance of the problem and the availability of personnel, but need not be held imediately upon identification of a Nonconfomance. For example, the Chairman should prepare a meeting l agenda or may choose to investigate the problem prior to convening a meeting. However, the Chairman is responsible for assuring that 1
meetings are held in time to assure that Reportable Occurrences are indeed reported within the required time period (see Appendix 1 in NPAP C-11/60AP W-111).
- 2. A written record shall be kept of the review group's findings. Form 76-286 is used for this purpose, although supplemental sheets may be attached as required.
L C. Responsibilities and Authorities l 1. The technical review group is responsible for:
- a. Evaluation of the Nonconformance,
- b. Determining its reportability (see NPAP C-11/GOAP W-111).
I NOTE: An item may already have been reported prior to the time a TRG is convened, in which case the TRG is concerned only with followup reporting.
- c. Determining the probable cause, if practical.
- d. Detemining and approving the resolution of the Nonconformance and any appropriate corrective action to prevent recurrence,
- e. Establishing a schedule for implementation of the resolution and corrective action.
- f. Verifying the implementation of the resolution and corrective action.
l 0
, _ _ _ _ w. -- . __
NPG01 17
4 038 (7/61)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER fFAP C-12 i NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION GOA W-112 3
IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 18 0F 42 APPENDIX B (Continued)
- 2. The technical representative (s) on the group is responsible for:
i
- a. Detemining the technical aspects of the resolution and corrective action.
- b. Detemining whether any emergency measures are required to -
l achieve safe operating conditions prior to final resolution.
l
- c. Referring items to the Plant Staff Review Connittee when required or as desired.
- 3. The Quality Assurance representative on the review group is responsible for:
- a. Ascertaining the acceptability of proposed dispositions and I
l corrective actions with respect to the established Quality i
Assurance Program. The Quality Assurance Department does not have the responsibility to detemine that the proposed dispositions and corrective actions are technically correct.
- b. Verifying that disposition actions have been coupleted and d signing the Monconformance Report to so indicate.
- 4. The Quality Control representative is responsible for:
- a. Assisting the technical representative in researching problems and developing an appropriate resolution and corrective action.
- b. Acting es in'am e between a site Technical Review Group and the PSRC when itens are referred to the latter.
l c. Serving as secretary for the review group meetings.
- d. Keeping a log of all outstanding Nonconformances. Separate l logs will be maintained for General Office / Plant initiated NCRs.
- e. Conducting inspections as may be required when nonconfoming items are reworked.
- 5. All decisions of the Technical Review Group must be unanimous. If the members cannot agree, the matter shall be referred to the Manager, NPO, Plant Manager / Superintendent and the Manager. Quality Assurance.
- 6. In the event of an emergency, the Plant Manager / Superintendent or his authorized delegate may take actions necessary to restore a safe operating condition regardless of unanimity.
l NplVil 1R
o C% o)2 (7/EI)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12' NUCLEAR PLANT ADMINISTRATIVE PROCEDURE GOA W-112 REVISION IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 19 0F 42 APPENDIX B (Continued)
D. Dispositions The Technical Review Group has several options available for detemining an acceptable resolution, including:
- 1. Accept As-Is
- a. Procedures, specifications, drawings, activities, or items may be accepted "as-is". This resolution requires that technical evaluations be perfomed, documented, and approved to assure that there will be no adverse effect upon the safety, operability, or maintainability of the items or of the component or system in which it is installed. Examples of circumstances which may lead to accepting "as-is" include specifications which oriCinally were excessively stringent, i
modifications to related items and activities which render the "Nonconformances" acceptable; or the initial identification of the Nonconformance was in error and the requirements are being
., met without any change being necessary.
- b. When a supp11erl proposes that PGandE accept a nonconfoming' item, the PGendE specifying organization shall require the supplier to identify the nature and extent of the
~-
Nonconformance and the reason for proposing its acceptance, use or installation.
- 2. Rework
- a. Items or activities may be reworked or repairt.d to conform with original design and/or specification requirements if practices and procedures are used which were approved in advance of the work being done.
l
- b. Itams which have been reworked or repaired shall be inspected to the original requirements, or to criteria established by the Technical Review Group if it considers the original requirements to be no longer applicable.
- c. If the Nonconfomance involves ASME Code materials or items and the disposition is to rework or repair, then the Authorized Inspector shall be promptly notified and his concurrence obtained.
1 The tem " supplier" includes any individual or organization who can furnish items or services. Included are manufacturers, contractors, subcontractors,
, distributors, service shops, and consultants.
WPcol 14
_ _ - _ , . _ _ _ . . . _ . _ _ _ _ _ _ _ - _ _ - _ _ - _ . _ ~ _
[
, P.-03 8 (7/61)
- NUCLEAR PL. ANT OPERATIONS DEPARTMENT NUMBER -2 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION 6 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83
- TITLE PROBLEMS AND NONCONFORMANCES j PAGE 20 0F 42
, APPENDIX B (Continued)
- 3. Reject n:
Items may be rejected. Items that are rejected shall be clearly L
marked and/or identified to prevent their inadvertent use.
- 4. Repair, Revise or Modify r a. Itams or activities may be repaired, modified or revised so
- that they are acceptable for use, although not always confoming totally to original requirements.
i b. When a modification involves a design change, the requirements
" of the applicable General Office / Nuclear Plant Adninistrative Procedures shall apply.
h
' c. When a modification involves revising approved procedures or
- development of new procedures, the requirements of the b
applicable Eeneral Office / Nuclear Plant A&ninistrative Procedures shall apply.
- d. Items which have been repaired, revised or modified shall be
- inspected to the original requirements, or to criteria established by the Technical Review Gmup if it considers the
[
E original requirements to be no longer applicable.
f The Technical Review Group may detemine that all or part of the issues raised on a Nonconfonnance Report should be reworded, b
processed separately, or submitted to a different Technical Review Group. In such instances, the disposition of these issues may be handled by issuing additional nonconformance reports, as applicable.
[s E. Resolution of Nonconformance by Suppliers E
If a supplier has an approved quality assurance program for handling k nonconforming items, the P6andE specifying organization may delegate the E
authority to the supplier to perform a technical evaluation of those
& nonconforming items which are under the direct control of the supplier.
PGandE retains the responsibility to assure that the resolution of Nonconfonnances is acceptable and satisfactory. The organization
- delegating the authority shall establish procedures and controls and monitor the supplier's activity to assure that nonconformances are E handled correctly and affectively. When these conditions are met. PGandE is not required to issue a separate Nonconformance Report for each g nonconfomance report that a supplier may issue. Where such authority is delegated specifically in the contract or purchase order, the supplier is w responsible for establishing that:
b e
i NPG01 20 N ,
. s',-028 (7/E1)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER IFAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION GOA W-112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 i . TITLE PROBLEMS AND N0hCONFORMANCES PAGE 21 OF 42 l
~
APPENDIX B (Continued)
- 1. All procedural actions comply with the requirements or
, specifications approved by PGandE;
- 2. Personnel perfoming the evaluation are qaalified. PGandE reserves j the right to review the qualifications of contractor perscanel perfoming such evaluations.
C NPG01 21
C.-028 (7/E1)
NMNNNT E E NUMBER REVISION WAP C-12 112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 22 0F 42 APPENDIX C
- INSTRUCTIONS FOR COMPLETING hUCLEAR PLANT 61-4516) PROBLEM REPORT Yhis Appendix describes the various sections of the Nuclear Plant Problem Report and the type of infor1 nation to be included on the form. Pmeedures i which specify who completes, myiews and approves the various portions of the
! form; and how forms are distributed and tracked shall be developed by each plant as a supplement to this procedure. Quality Control Departments at each plant should further action by provide an annual trend analysis of NPPRs which may require NPD Management.
A. INITIATING DEPARTMENT
- 1. Identification '
The person who identifies a problem shall report it promptly to his supervisor.
The supervisor is responsible for assuring that a Nuclear Plant Problem Report (From 61-4516) isinitiated(except that the person who identifies the problem may choose to defer filing a Problem Report until after conferring with his supervisor).
If a NPPR was written ar.d a problem does not exist, the supervisor .
must sign the form and state why the problem is not valid. The form must then be routed to management for their concurrence and normal i
processing to close the item.
Upon notification of a potential problem the supervisor of the individualiswho problem identified the problem will decide whether or not the valid. If not, the process stops at this point. If l valid, the INITIATING DEPARTMENT portion of the form shall be completed and signed off.
- 2. Instructions l
- a. IDENTIFICATION This is a four part twelve charactor description code that is to uniquely identify each Problem Report.
l
- 1) The first three characters identify the applicable site / plants and the unit; authorized codes are:
DC0 l Diablo Canyon, General (affecting all units)
! DC1 Diablo Canyon. Unit 1 DC2 Diablo Canyon, Unit 2 HBO Humboldt Bay, General (affecting all units) 21 Humboldt Bay, Unit 1 NPG01 22
C s 0 H (~i/ h )
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER IFAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE
' REVISIOh IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 23 0F 42 APPENDIX C (Continued)
HB2 Humboldt Bay, Unit 2
, HB3 Hinboldt Bay, Unit 3 HB4 Humboldt Bay, MEPP SPO Stanislaus Nuclear Project, General (affecting all units)
NGO Nuclear Plant, General
- 2) The second two characters are the last two digits of the year. 81 is for the year 1981.
- 3) The third pair of characters identify the department which initieted the report.
Plant Staff (Nuclear Plant Operations) PG t
Plant Staff Review Committee PS i
Operations OP l Instrument and Controls TI
, Chemistry and Radiation Protection TC
. Plant Engineers TN Mechanical Maintenance 904 Electrical Maintenance EM Technical Support ST Security SE Quality Control QC Materials Facility MF Office (Personnel and General Services) 0F l l
Training TR l Document Control DC l Suggestion System SS ISI and NDE Services IS I Bioassay (DER) BE Quality Assurance QA On-site Safety Review Group SR Station Construction SC t
l t
Resident Electrical RE Resident Mechanical RM Resident Civil RC Resident Startup RS
! Although not normally found on Nuclear Plant Problem Reports, the following are found on Nonconformance Reports issued by other Company departments.
W Df.'111 92
(wo35 (7/h)
CIUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER WAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS!0f0A W112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 i TITLE PROBLEMS AND NONCONFORMANCES 24 PAGE 0F 42
~
APPENDIX C (Continued)
Engineering, General EN Electrical EE Mechanical and Nuclear ME Civil CE l
i Engineering Quality Control EQ Engineering Services ES Design Drafting DD General Construction GC Materials - ML Engineering Research ER Nondestructive Testing NT Standards Laboratory SL Siting SI Nuclear Pmer Generation NG Nuclear Plar.t Operations (General Office) NO Nuclear Projects NP Quality Assurance QA Meteorology Office MO
- 4) The fourth part of the identification is a four digit l
unique number which is assigned by the organization which initiates the report. These numbers will start at P0001 l
each January first and increase sequentially through the year.
As an example the fourteenth NPPR identified by the Diablo Canyon operators on Unit 1 in 1977 would be identified as:
l DC1 77 OP P0014 The prefix P is used to distinguish Problem Reports from Nor.conformance Reports (Form 76-286) wnich use a similar
! numbering system except that they use the prefix N.
- b. ITEM OR ACTIVITY Describes the general system or activity involved (e.g.,
reactor coolant pung, containment leak rate test).
l NPG01 24
t'.'.-0] 8 ( 7 /s.1 )
WUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER IFAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE 112
, REVISION IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE "R03LEMS AND NONCONFORMANCES PAGE 25 CF 42 APPENDIX C (Ccntinued)
- c. PROBLEM
- 1) Describe the discrepancy or departure in sufficient detail to illustrate the problem.
- 2) A suggested resolution may be made. This can serve two purposes:
a) In many cases, the person reporting a problem has a good idea for what should be done and his suggestion i
could be helpful in detennining the disposition.
.This is optional, and trivial casunents (" repair it")
are not appropriate or necessary.
b) The second use for suggested resolution is to indicate work which has already been done for those cases where work has preceded the issuance of the fonn. Although the work may be complete, it still only has the status of a " suggested resolution" until the appropriate approvals have been received.
- d. STATUS The use of this section of the form is to infor:n those people who will review the fonn, and to whom it will be transmitted for resolution, what the status of the problem is.
In the first portion of this section, the originator should
! indicate the status of the suggested resolution, if any. That j
is, has it already been implemented and the purpose of the report is simply to indicate what has been done, is the work in progress, or has the suggested resolution not yet been started?
In the remaining portions of this section more detailed infonnation on the status of the item can be given. As a minimum, the appropriate portion (s) should be filled out if the resolution has not yet been detennined, or if it has not yet been started or is not complete.
- 1) For plant equipment problems, indicate the operational and/or clearance status; 1.e., in service, cleared and tagged to Shift Foreman, partial clearance (electrical, mechanical),notcleared. Provisions are also provided for indicating whether the item is tagged with a Maintenance Work Tag or Information Tag; the type of clearance re appropriate); quired whether(circle a SWP " plant" or " dispatcher",
is required; and any otheras special conditions or cautions that are appropriate.
NPGOI 25 i
C.-03B (7/h )
C UCLEAR PLANT OPERATIONS DEPARTMENT NUMBER l@AP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS10@0A W-112 IDENTIFICATION AND RESOLUTION OF 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 26 0F 42 APPENDIX C (Continued)
- 2) For material problems, such as nonconforming spare parts,
. the originator should indicate the location of the material and whether it is tag ed (see Nuclear Plant Administrative Procedure D-500 .
- 3) For procedural or administrative problems, the originator should indicate whether the item is complete, continuing, or halted pending resolution. Pmvision is also included for amplifying remarks as appropriate.
- 4) Finally, a remarks section is provided where explanatory notes' can be entered or types of problems can be discussed which are not adequately covered by the previous sections.
- e. PRIORITY This section is so the originating group can provide their assessment of the priority which should be placed on the item.
This is for information purposes only, and does not represent a binding order on the department which must perform the work.
The final priority is determined by plant management.
"N/A" should be used in the case where scheduling is not a factor, such as when the work is already accomplished and the form is being submitted for reporting purposes only.
"Begin ASAP" should be assigned to work which is necessary to restore plant equipment for which allowable repair times hate been specified in the Limiting Conditions for Operation in the Technical Specifications. This priority should also be assigned to work which is intended to correct a problem which is actually causing a Unit curtailment or which results in a
- significant decrease in Unit reliability.
"Begin when schedule pemits" priority should be assigned whenever the previous pricrity considerations are not a factor.
This priority should be spe:ified whenever possible because it provides the affected department maximum flexibility in scheduling its work.
" Unit outage" priority should be used when radiation levels, considerations of unit availability and reliability, or other factors dictate that a Unit outage is required.
"STS/LC0" If an item is a Technical Specification requirement and involves an LCO, this fact should also be so indicated.
.8 Sh d4 db . Sb dP
e 0 -028 (7/61)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER NPAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIOgDA W-112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 l - TITLE PROBLEMS AND NONCONFORMANCES PAGn 27 OF 42 l APPENDIX C (Continued)
- f. INITIATING DEPARTMENT SIGN OFF The initiating department should sign off in accordance with plant procedures.
B. INITIAL MANAGEMENT REVIEW
- 1. Problem reports shall be distributed to an appropriate level of plant management for an initial review of the item to determine if it is important to safety, environmental quality related, a nonconformance, or is potentially reportable. The review should also consider the appropriateness of proposed resolutions and use of procedures.
- 2. ITEM / ACTIVITY IS
- a. If the item is Q-Listed, it should be so checked.
- b. .If an item is Safety Related. it should be so checked. The
{, criteria for identifying safety related items is given in Nuclear Plant Administrative Procedure D-1.
- c. If an item is a potential Nonconfonnance, it should be so noted and an NCR initiated. Criteria for identifying Nonconformances are given in Appendix A of this procedure.
- d. If an LER is required, it should be so noted. Guidance for submission of LERs is contained in the Tech. Specs., Reg. Guide 1.16 and NPAP C-11/GOAP C-111.
- e. If a report to the NPRDS is required it should be noted,
- f. If the item needs to be placed on the outage schedule it should be noted.
- 3. RESOLUTION
- a. If a resolution has been proposed and the reviewer concurs, then this should be checked.
- b. If an activity has been halted (or proposed corrective action has not been started) the reviewer can indicate whether work may or may not proceed by circling the appropriate statement.
O
.eSete 66
.?.-02 ( *t/i1 )
s C:UCLEAR PLANT OPEPATIONS DEPARTMENT NUMBER W AP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIOg0A W-112
% IDENTIFICATION-AND RESOLUTION OF DATE 6/15/83 m TITLE PROBLEMS AND NONCONFORMANCES ,PAGE 28 0F
- 42 x x APPENDIX C (Continued)
- c. If the reviewer chooses to modify a proposed resolution
- _ or to offer an alternative resolution, he may so indicate under
~
"Other".
s
- 6. In some cases the reviewer may be unable to connent in the j
i s
' preliminary myiew regarding the resolution, in which case N/A
- can be checked. This often occurs on equipment problems when w the appropriate resolution is unknown prior to the time the
- . msponsible maintenance group takes the equipment apart and diagnoses the problem.
- e. If the problem is well understood and other departments will be involved in addressing the problems the routing of the NPPR s
should be so noted.
- 4. RESPONSIBLE DEPARTMENT l
i This?i s the department to whom the Problem Report is being '
l m
transmitted for action and resolution. The department checked here
- represents the opinion of the reviewer as to who should handle the j problem, and the purpose of this section is mostly for assisting the clerical staff - i.e.. into whose mail box should the form be placed. 'No= commitment is intended that the checked department must actually resolve the problem, if later determined by the department head (s) that another department is more appropriate.
The blanks may be used to specify a sequential routing order if the problem involves 'several departments.
' ~^ '
~
Final priority, if other than that already specified, can be
. assigned.
N' s i
I s
- 5. PROCEDURE REQUIREMENTS
\
Tht ' reviewer may ~ choose to specify the requirements for use of specific procedures. The use of procedures for various jobs is dis:;vssed in the Adninistrative Procedures applicable to the type of work' being performed.
7 If procedural requirements are nnt specificall "To be Detsnnined by Responsible Supervisor". yThe known, latterchecksimply N/A or means that the foreman or other first line supervisor will make this determir,ation, based upon his assessmeat of the nature and l 3 complexity of the task.
s -
s i# g 4 s-g, \s g'
f 1
I
, . _ , , _ . . . . _. ,,_..._,.__._....._-..___._,,_.,___...,__.m_,_._._______.,.._..___-_,___,,._._.____.-_..__m
. s 023 (7/il)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER NPAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE W-112 REVIS!O 1DENTIFICATION AND RESOLUTION OF lDATE 6/15/83 TITLE PROBLEM 5 AND NONCONF0fo4ANCES 29 0F
( PAGE 42 i
APPENDIX C (Continued)
- 6. POST-MAINTENANCE TEST On equipment problems where operability must be proven or equipment needs to be tested to diagnose the probism, the reviewer should check that a test is required.
- 7. SIGN OFF The person (s) who completes the management review should sign off.
Although there are two sign off blanks, this does not mean that two signatures are required - only one is. However, if two (or more) people review the form, they should both sign.
- 8. Convening Technical Review Group If the department head identified the problem as either a nonconfomance or potentially reportable, he will be responsible for assuring initiation of the hCR and convening the Technical Review
{. C.
Group in accordance with Appendices B and 0 of this procedure.
IMPLEMENTATION OF RESOLUTION
- 1. The supervisor (s) who receives the Problem Report will perform the i resolution (assuming, of course, that there is no restriction placed l on him by the management reviewer. He should also complete the Nuclear Plant Problem Report fom.
l 2. SUPV: ASSN. TO l The supervisor, and if appropriate, the person doing the work, and the inspector (if there is one) should be identified here.
- 3. RESULTS OF IhVESTIGATION CAUSE/ ACTION TAKEN The results of investigation by the department doing the work and l
i any action taken should be discussed in detail here. Efforts should be made to determine the cause of the failure. If a separate report is warranted, it should be referenced to and included as an attachment to the Problem Report fom. If after working on the problem, the Supervisor later determines that the problem should be i resolved by roother department, it should be passed on to them for l
resolution.
l l
l
,,nens nn
1
( .-02 5 (7/h ) .
C:UCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIO W-112 ZDENTIFICAT M AND RESOLUTION OF DATE 6/15/83 TITLE- PROBLEMS :n NONCONFORMANCES PAGE 30 0F 42
, APPENDIX C (Continued)
Once the work is complete and post-maintenance testing is required, the supervisor in charge sends the form to the Shif t Foreman or an appropriate planner / coordinator who then would detennine and arrange therequiredtest(t}-
The supervisor and person wha performed the test should be identified in the designated bic:k. <
- 4. REFORT CLEAR When the job is complete, the tests evaluattd and the system again operable, the supervisor in charge should sign off as " reporting clear
- and date the form. If a piece of equipment is involved and he reports clear to some other supervisor, he can so indicate or mark N/A in the space marked "To".
D. FINAL PLANT MANAGEMENT REVIEW, SIGN OFF AND DISTRIBUTION After the work *s complete, the form should be routed for final -
management review and signoff in accorcance with plant procedures.
\
, . I di
-- w 00000% SA
NbCLEAR PLANT OPERATIONS DEPARTMENT NUMBER NPAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REvlslog0A
' 6 IDENTIFICATION AND RESOLdTION OF DA E 6/15/83 TITLE PAGE 31 PROBLEMS AND NONCONFORMANCES OF 42 p%
APPENDIX D DESCRIPTION OF GENERAL OFFICE PROBLEM REPORT RESOLUTION PROCESS AND .
. .c IbSTRUCTIONS FOR COMPLETING AND HANDLING FORM 69-025.
4.%
This Appendix describes the general process to be followed from the time a problem is identified until the problem is resolved. 60-Quality Control V")fh should provide an annual trend analysis of initiated G0FRs which may require
.i^w
- 1. ,
further action by NPO Management.
Q A;
~ .y A. INITIATING DEPARTMENT ydjj
. a e:, .
- 1. Identification The person who identifies a problem shall report it promptly to his W 3.. %
- Y supervisor (typically a senior engineer). The supervisor or the identifying person is responsible for assuring that a General Office Problem Report (Forin 69-025) is initiated (except that the person who identifies the problem may choose to defer filing a General Office Problem Report until after conferring with his supervisor).
If a GOPR was written and a problem does not exist, the supervisor
(',
must sign off on the form and state why the problem is not valid.
The forin must then be routed to management, as instructed in L paragraph D, for their concurrence and normal processing to close the item.
- 2. Upon notification of a potential problem, the supervisor of the individual who identifies the problem will decide whether or not the problem is valid. If not, the process stops at this point. If valid, the person who identified the problem or the supervisor should complete the INITIATING DEPARTMENT portion of the form.
- 3. Instructions
- a. IDENTIFICATION This is a four part twelve character description code that is to uniquely identify each Problem Report.
- 1) The first ttree characters identify the applicable site / plants and the unit; authorized codes are:
DC0 Diablo Canyon, General (affecting all units)
DC1 Diablo Canyon, Unit 1 DC2 Diablo Canyon, Unit 2 HBO Hunboldt Bay, General (affecting all units' El Htsnboldt Bay, Unit 1 HB2 Humboldt Bay, Unit 2 23 Humboldt Bay, Unit 3 SP0 Stanislaus Nuclear Project General (affecting all units)
NGO Nuclear Plant, General (includes General Office)
..' so)8 (7/il)
- C'UCLEAR PLANT OPERATIONS DEPARTMENT NUMBER WAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIONGOA W-112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 32 0F 42 APPENDIX D (Continued)
- 2) The second two characters are the last two digits of the
, year. 81 is for the year 1961,
- 3) The third pair of characters identify the department which initiated the report.
i Generel Office Staff NO (Nuclear Plant Operations)
Training TG Department Administration DA Huclear Safety and Engineering N5 Personnel and Environmental Safety PE Although not normally found on General Office Problem Reports issued by the General Office, the following are found on Nonconformance Reports issued by other Company departments.
Engineering,_ General EN -
Electrical EE Mechanical and Nuclear ME Civil CE Engineering Quality Control EQ Engineering Services ES Design Drafting DD General Construction GC Materials ML Encineering Research ER Nondestructive Testing NT Standards Laboratory SL Sitino SI Nuclear Power Generation NG Plant Staff (Nuclear Plant Operations) PG Operations OP Instrument and Controls TI Chemistry and Radiation Prottetion TC Nuclear Engineers TN Mechanical Maintenance 2
I e
, C.-013 (7/h )
C'UCLEAR PLANT OPERATIONS DEPARTMENT NUMBER fFAP C-12 ,
NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REV1510f0A W-11e t
IDENTIFICATION AND RESOLUTION OF DATE 6/15/83
, TITLE PROBLEMS AND NONCONFORMANCES PAGE 33 0F 42 APPENDIX D (Continued) '
Electrical Maintenance EM Security 1
SE Quality Control QC Materials Facility MF
, Office / Record 0F Bioassay (DER)
BE l
juality Assurance t QA On-site Safety Review Group SR Station Construction SC Resident Electrical RE Resident Mechanical RM Resident Civil RC Retident Startup RS Nuclear Projects NP Quality Assurance QA Meteorolony Office MD
- 3) The fourth part of the identification is a four digit unique number which is assigned by the organization which initiates the report. These nusbers will start at G0001 each January first and increase sequentially through the year. Quality Control controls the nunbering sequence.
As an exanple the fourteenth GOPR identified by the General Office QC Engineer 1982 would be identified as:
NGO 82 DA G0014 The prefix G is used to distinguish Problem Reports from Nonconformance Reports (Forn 76-286) which use a similar numbering system except that they use the prefix N.
- b. PROBLEM Describe the discrepancy or departure in sufficient detail to illustrate the problem. Indicating whether the itenVwork is complete, continuing or halted pending resolution.
NPr,01 '41
O-02 6 (7/h ) '
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER FAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION0 UII2 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 34 0F 42 APPENDIX D (Continued)
- c. A Recommended Solution may be made. This can serve two
, purposes:
- 1) In many cases, the person reporting a problem has a good idea for what should be done and his sug helpful in determining the disposition. gestion This iscould be optional, and trivial comments (" correct it") are net appropriate or necessary.
- 2) The second use for Recomended Solution is to indicate work which has already been done for those cases where work has preceded the issuance of the form. Although the work may be complete, it still only has the status of a "Recomended Solution" until the approval of the Technical Review Group has been received.
- d. INITIATING DEPARTMENT SIGN OFF The person who reports the problem and/or originttes the form should sign off. In addition, his supervisor should sign off and date the form, indicating that he is in agreement. ,
- a. DISTRIBUTION Following the completion of the INITIATING DEPARTMENT, the form should be distributed as follows:
- 1) Original to the supervising / senior engineer of the department which will hardle the resolution in accordance with General Office procedures.
- 2) Information copy to Quality Contrcl Engineer, where it will be entered into NPO's tracking system.
B. INITIAL MANAGEMENT REVIEW
- 1. Upon receipt of a Problem Report, the appropriate supervising / senior engineer is responsible for making an initial review of the item to determine if it is nuclear-related, environmental quality related, a nonconformance, or is potentially reportable. He should also determine the priority level and indicate his views regarding the appropriateness of proposed resolutions and use of procedures, and should document these decisions on the Problem Report form.
- a. RESOLUTION
- 1) If a resolution has been proposed and the supervisor concurs, then this should be stated.
- 2) If the supervisor chooses to modify a proposed resolution or to offer an alternative resolution, he may so indicate.
a . -_
\
WDcn1 1A
_ _ _ _ _ ]
. t-028 D/ill NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER @AP 'C-12 NOCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS!g0 W112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND N""CONFORMANCES PAGE 35 0F 42 i APPENDIX D (Continued)
- b. CAUSE OF PROBLEM i
. Define how and why the problem occurred.
- c. CORRECTIVE ACTION TO PREVENT RECURRENCE List any measures taken to preclude repetition of significant or recurring discrepancies, departures, or conditions adverse to quality from reoccurring again.
- d. IMPLEMENTATION ASSN. TO The supervisor, as well as the person doing the work (if other than supervisor himself), should be identified here. The responsible department represents the opinion of the originator and his supervisor as to who should handle the problem.
- e. SIGN OFF The person (s) who completes the Initial Management Review should sign off. However, if two (or more) people review the ferm, they may all sign.
- f. Distribution i
l Following the initial review by management, the form should be distributed as follows:
- 1) Original I Transmit the original to the person who has been assigned the work.
- 2) Copy Transmit to Quality Control Engineer. This will serve to inform him of problem status. He should also maintain an active file of outstanding problem reports and , assure that resolutions are timely. If the problem has been identified as a potential NCR, the QC Engineer should send an infonpation copy with an assigned NCR number to the QA depa rtment.
m --- - --
O -018 (7/61)
C'UCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIOg0A W112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 36 0F 42
~
APPENDIX D (Continued)
- 2. Convenir.g Technical Review Group If the supervisor identified the problem as either a Nonconfomance or potentially reportable. he will be responsible for convening the Technical Review Group in accordance with Appendix B of this procedure.
C. IMPLEMENTATION OF RESOLUTION The person (s) who receives the Problem Report will carry out the appropriate resolution (assuming, of course, that there is no restriction placed on him by the supervisor).
l
- 0. FINAL MANAGEMENT REVIEW u
- 1. The results of investigation by the department doing the work and j any action taken should be discussed in detail here. Efforts should l be made to determine the cause of the failure. If a separate report is warranted, it can simply be referenced to as an attachment to the Problem Report form. If after working on the problem, the supervisor later detennines that the problem should be rest,hed by 1 aaother department, he should state his findings on the form and return the 60PR to his supervising engineer. The supervising engineer signs off the FINAL MANAGEMENT REVIEW block before he passes on the form to the other department.
- 2. Sign Off and Distribution
- a. The supervising engineer reviews the form and signs off,
- b. After sign off, the supervising engineer should pass the original and a copy to the QC Engineer for processing.
E. QUALITY CONTROL 1
The QC Engineer should complete and distribute the copies as follows:
- 1. Coglete verification.
, 2. Copy and any attachments will be sent to the supervisor who t originated the Problem Report for his information.
l 3. Route copy to the Manager, Nuclear Plant Operations for his information.
- 4. Make additional copies of the raport and distribute as necessary.
- 5. File the original in the Central File Room and RMS.
N Df1A1 %
l l _ __ .,_ . .
. - - . . , . - - - - - - - - - - - - - ~ - - - - - ~ ~ - ~ ' ' - - - - - ' ' ' ' ' ' ' ~ ~ ~ ~ ~ ~
C.-028 (7/61)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12 REY!sIOgDA W-112 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE IDENTIFICATION AND RESOLUTION OF DATE 6/15/83
. TITLE PROBLEMS AND NONCONFORMANCES PAGE 37 0F 42 APPENDIX E DESCRIPTION OF NONCONFORMANCE REPORT RESOLUTION PROCESS AND INSTRUCTION FOR UMPLETING AND HANDLING FORM 76-286.
This Appendix gives instructions for completing Nonconformance Report (NCR)
Form 76-286. This form is used by Nuclear Plant Operations to identify and process nonconformances as required by Quality Assurance Pmcedure 10.1.
Forn 76-286 "Nonconformance Report", is used as the basic ' minutes" of the Technical Review Gruup meeting when a problem has been classified as either a nonconformance or potentially sportable. It is also used to document reporting and the verification of the completion of the resolution.
Completion of the form ir the responsibility of the Chairman of the review group.
A. Item 1. IDENTIFICATION This is a four-part, eleven-character description code that is to uniquely identify each NCR. '
[' 1. The first three. characters identify the applicable site / plants and
%. the unit; authorized codes are:
DC0 Diablo Canyon, General (affecting all units)
DC1 Diablo Canyon, Unit 1 DC2 Diablo Canyon, Unit 2 l HB3 Humboldt Bay, Unit 3 SPO Stanislaus Nuclear Project, General (affecting all units)
NGO Nuclear Plant, General
- f. The second two characters are the last two digits of the year.
81 is for year 1981.
- 3. The third pair of characters identify the PGandE department most appropriate to provide the chairman of the Technical Review Group.
GENERAL OFFICE ggineering, General EN Civil Engineering CE Design Drafting DD Electrical Engineering EE Engineering Quality Control EQ
- Engineering Services ES l , Mechanical and Nuclear Engineering ME l
. _- m __ --
h k k
." -022 (~t/i1i C'UCLEAR PLANT OPERATIONS DEPARTMENT NUMBER NPAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION GOA W-112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 38 0F 42 APPENDIX E (Continued)
Engineering Research ER General Construction GC Materials ML Meteorology Office M0 Nuclear Power Generation NG Nuclear Plant Operations NO Nuclear Projects NP Quality Assurance QA Siting
~
SI SITE Bionssay (Engineerino Research BE Plant Staff, General (Nuclear Plant Operations) ~PG '
Chemistry and Radiation Protection TC Electrical Maintenance EM Instrument and Controls TI Materials Facility MF Mechanical Maintenance MM Nuclear Engineers TN Operations OP Quality Control QC Security SE Tecnnical Support ST Trainir.g TR ISI and NDE Services IS l
Station Construction, General SC Resident Civil RC Resident Electrical RE Resident Mechanical RM Resident Startup RS i 4. The fourth part of the identification is the letter N followed by a l
three-digit, unique number which is assigned by the department
! identified in c. abwe. These numbers will start at 001 each l
January first and increase sequentially through the year. Each department which may issue a nonconformance report shall establish controls for assigning sequential numbers for Item I for each l calendar year.
l unem so 1
i.-018 (7/E11 NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER RAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS10h W 112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 39 0F 42 APPENDIX E (Continued)
As an example, the fourteenth NCR assigned to the Diablo Canyon
, Plant Staff for a nonconformance on Unit 1 in 1980 would be identified as DC1-80-PG-N014 The prefix N is used to distinguish Nonconfomance Reports from Plant Problem Reports and General Office Problem Reports which use a similar numbering system except they use the prefix P and G, respectively.
i B. Itams 2 through 5 shell be completed by the person issuing the report.
- 1. Item 2. ITEM OR ACTIVITY Describes coolant pump, thecontainment general system or activity) leak rate test . involved (e.g., reactor
, 2. Item 3. REFERENCES / REQUIREMENTS This identifies the basic reference document which contains the requirement (e.g., drawing number, specification number. Technical l Specification reference) which was not met. Other pertinent l
documents, audit reports, inspections, departmental problem reports,
( etc., may also be referenced here.
- 3. Item 4. DESCRIPTION Describe the discrepancy or departure in sufficient detail to illustrate the problem.
- 4. Item 5. ORIGINATED The issuer lists his department, the date, and signs the form.
Distribution shall be made in accordance with Item 15 - '
I DISTRIBUTION.
C. The Technical Review Group must first determine whether the problem is actually a Nonconformance meeting the definition in Quality Assurance Procedure 10.1. If it is, the group completes Items 6 through 12. If it i
is determined not to be a Nonconforinance, it must be so stated in the i disposition and shall be referred to the responsible supervisor.
I
- 1. Item 6. CAUSE OF NONCONF0mhNCE A brief explanation of the cause of the Monconforiaance shall be
- stated in this section. In some cases the explanation may involve a study which can be referenced. If the cause cannot be specifically determined, the most probable cause(s) shall be described and entered in this section. _- - -- - __
l unem w a__ , _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ -
MB (7/h)
C'UCLEAR PLANT OPERATIONS DEPARTMENT NUMBER PFAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION N'II2 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 40 0F 42 APPENDIX E (Continued),
- 2. Item 7. RESOLUTION In this section, the Technical Review Group specifies the disposition of the Nonconformance.
The resolution shall be described as one of the following four options: accept as is, rework, repair or modify, or reject, with specific amplifying information. When procedures, practices, or administrative controls are involved in a Nonconformance, the most likely resolution will be " modify." If an NCR must be r; vised, a new Nonconformance Report shall be written. The new Nonconformance Report shall reference the one it supersedes under " Reference Requirements" and an entry shall be made in the " Additional Remarks" section of the superseded NCR referencing the number of the new NCR.
- 3. Item 8. CORRECT!YE ACTION TO PREVENT RECURRENCE Whenever possible, the steps necessary to prevent recurrence of the nonconformance shall be described. If the Nonconformance is a random incident, an isolated cate, or for some other reason it is determined that no corrective action is merited, then the basis for '
this determination shall be stated.
- 4. JJem 9. SCHEDULED C(MPLETION The proposed schedule for completion of the resolution and corrective actior shall be entered.
- 5. Item 10. REPORTABILITY AND NOTIFICATION
- a. The detemination of whether or not an item is reportable, whether or not an item is a Stbstantial Safety Hazard, the bases for these determinations, and the applicable reporting requirements by which it was reviewed shall be stated; for t
example, 10CFR20, 10CFR21, 10CFR50.36, 10CFR50.55. Technical Specification Requirement 6.9.1.12.a. etc. If the review group cannot unanimously determine reportability, this section shall state the matter was referred to the next level review group for determining reportability; for example, Site Review Group and/or General Office Review Group. (If there are any, identify those reporting requirements under which the item was reviewed for reportability, even if it turns out that the review group does not feel the item is in fact reportable.
This will assist the General Office Review Group if they have to reconsider a controversial item.)
Guidance on determining reportability is contained in NPAP C-11/GOAP W-111.
NPG01 40
( 038 ( //61) 9dUCLEAR PL. ANT OPERATIONS DEPARTMENT NUMBER W AP C-12 W-112 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISI
^
IDENTIFICATION AND RESOLUTION OF DATE 6/15/83
- TITLE PROBLEMS AND NONCONFORMANCES PAGE 41 0F 42 APPENDIX E (Continued)
- b. The type of report (s) required by the NRC or other agency and their timing should be entered. For the initial report, the reporting method (telephone, written, etc.) should also be indicated. (Follow-up reports are always in writing.)
Provision is also included for entering the information regarding the actual reports when they are made. This information should be provided to the department holding the original NCR by the department making the actual report.
- c. The Plant Maiiager and/or Plant Superintendent and the Manager.
Nuclear Plant Operations, must be promptly informed of all potentially reportable items. Likewise, the Manager. Quality Assurance, must be promptly notified if a Substantial Safety Hazard may be involved.
- d. In many cases. items which are reportable to the NRC are also reportable to other agencies, such as the Regional Water Quality Control Board on NPDESI permit violations and the
(' Department of Transportation (DOT) on radioactive material L shipment anealies. For completeness, a place is provided where these other agencies can be identified.
- 6. Item 11. ADDITIONAL REMARKS Provision is included for entering additional comments and remarks that are appropriate.
- 7. Item 12. REVIEW GROUP APPROVAL
- a. Each member of the Technical Review Group shall sign and date the report. All review group decisions must be unanimous. The Quality Assurance representative shall sign approving the quality aspects of the cause and disposition for conformance to the quality assurance program. The Quality Assurance representative does not have the responsibility to detemine that the proposed dispositions and corrective actions are technically correct,
- b. All nonconformances pertaining to an operating plant must be reviewed by the PSRC. The date of the PSRC meeting at which the report was reviewed should be entered here.
D. Item 13 is to be completed by the organization (s) implementing the resolution and corrective action.
I NPDES - National Pollutant Discharge Elimination System WDCA1 A1
1
- - O-03 8 (7/h)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER WAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS!gGA W-112 DATE 6/15/83 IDENTIFICATION AND RESOLUTION OF TITLE PROBLEMS AND NONCONFORMANCES PAGE 42 0F 42 APPENDIX E (Continued)
Item 13. IMPLEMENTATION
- a. sign and date when the resolution is complete
- b. sign and date (if reciuired) when resolution has been inspected
- c. sign and date when the corrective action is complete
, d. sign and date (if required) when corrective action has been inspected E. Item 14 is to be completed by the Quality Assurance Department when all the preceding items have been completed.
- 1. Item 14. VERIF.ICATION The method of verification is dependent upon the type and extent of the resolution and corrective action and may include such activities as document reviews and inspections of physical work. The Manager, Quality Assurance, shall designate appropriately qualified individuals to verify and sign-off Nonconfonnance Reports. When Item 14 is correctly signed and dated, the report is considered -
closed.
s
- 2. Item 15. DISTRIBUTION The distribution blocks at the bottom of the fonn shall be checked when the form is initially issued by the issuing department. If the issuing department is not the responsible department, the original of the Nonconfonnance Report must be transmitted to the responsible department for resolution. The Quality Assurance Department and the appropriate Plant Manager and/or Plant Superintendent shall always be checked. If Engineering Department is involved, the appropriate discipline chief shall be written in the " Engineering" block, and Nuclear Prtjects and Nuclear Plant Operations shall be written in the " Nuclear Power Generation" block. Other departments may be checked as deemed appropriate. Informational copies shall be so marked or identified so that they cannot be confused with official i
working copies. After verification, completed copies shall be distributed to the appropriate departments by the Quality Assurance Department. The Quality Assurance Department is responsible for maintaining the file of closed Nonconformance Reports. T originating department is responsible for maintaining file,he copies of the Monconformance Report fonns which they initiate, NOTE: The previous discussion applies to the distribution of completed NCRs. In addition to the above, a preliminary copy shall be telecopied to the Manager, NPO when the NCR is first initiated. The Manager NPO should send an infonnation copy to the Project Completion Manager.
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EXHIBIT 13 PACIFIC GAS AND ELECTRIC COMPANY GENERAL CONSTRUCTION QUALITY ASSURANCE PROGRAM Procedure GCP-12.1 Rey, 3 Documenting Discrepancies and Change Notice No. 10 TITLE: Assessing Reportability Page 1 of 11 APPROVED: Date: >% io 14 '
8%
Tice President - General Construction (/
1.0 SCOPE 1.1 This procedure establishes the method for documenting discrepan- !
cies and assessing reportability. l 1.2 A discrepancy is any departure from the requirements of specifi-cations, drawings, procedures, codes, or other applicable docu-ments.
2.0 RESPONSIBILITY 2.1 The Senior Site Representative is responsible for implementing
( this procedure. !
2.2 Quality Control is responsible for maintaining a record of and monitoring the status of discrepancy reports.
2.3 In all areas of this procedure, supervisory personnel who have been assigned responsibility for a task have the authority to delegate performance of that task to subordinates. This dele-gation shall be by written job description or by specific written
- assignment.
-- 3.0 APPLICATION 3.1 This procedure applies when a discrepancy is identified in an item, work activity or documentation which is important to safety or environmental quality, is reportable, or requires quality assurance; or when a supplier or contractor violates P G and E quality assurance requirements. This does not include insignifi-cant departures which can be corrected in the nomal course of work. unless the Technical Specifications of a plant with an oper-ating license have been violated.
4.0 PROCEDURE 4.1 Discrepancies shall be documented on either a Nonconformance Report (Quality Assurance Form 76-286), a Minor Variation Report (General Construction Fom 776-101), a Nuclear Plant Problem Report (Nuclear Plant Operations Form 61-4516), or a contractor's discrepancy report.
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Proced:;re GCP-12.1 Rev. 3 Page 2 of 11 4.1.1 A nonconformance is a departure from requirements in pur-chase specifications, drawings, codes, standards, proce-dures or Nuclear Regulatory Comission regulations which is reportable or which, if left uncorrected, could result in any of the following:
4.1.1.1 Degradation or loss of integrity of the reactor coolant pressure boundary; or 4.1.1.2 Reduction or loss of the capability to shut down the reactor and maintain it in a safe condition, including the compromise of design objectives during construction or modification activities; or 4.1.1.3 Lack of effective control over items or activi-ties (including quality assurance program imple-mentation) that could reduce the capability to prevent or mitigate the consequences of accidents that may result in potential off-site exposures comparable to the guidelines set forth in Title
- 10. Code of Federa' Regulations Part 100, "Re-actor Site Criteria".
4.1.2 The following examples may also be considered nonconformances:
4.1.2.1 Repeated failure to follow approved procedures or
( to provide required documentation after the dis-crepancy has been identified and reported.
4.1.2.2 Discrepancies of a relatively insignificant nature but which, due to their repetition, require ac-tion by management.
4.1.2.3 Repeated failure to correct by the mutually agreed upon comitment date those discrepancies identi-fied in audits, surveillance reports, or insNc-tions, when such a delay is determined to have a significant effect on quality.
4.1.3 A minor variation is a discrepancy which is not reportar e and is not a nonconformance, but which nevertheless is a departure from specific requirements.
4.1.4 A Nuclear Plant Problem Report shall be initiated for problems identified by General Construction which require resolution by Nuclear Plant Operations. These problems may be of a material nature regarding equipment, systems, components, and structures or they may address procedural or administrative deficiencies which affect the coordina-tion of work efforts between General Construction and i Nuclear Plant Operations. A Nuclear Plant Problem Report
Procedure GCP-12.1 Rev. 3 Page 3 of 11 may also be initiated for problems at a licensed plant or unit which have been identified by General Construc-tion involving a question of acceptance or operability of systems, equipment, components or structures.
4.1.4.1 Site procedures shall be developed in accordance with Procedure GCP-5.1, " Approval and Control of Field Procedures", to provide instructions for issuance, approval and control of Nuclear l Plant Problem Reports.
4.1.5 A contractor's discrepancy report is required when a supp-lier or contractor having an approved quality assurance program violates P G and E or self-imposed quality assu-rance requirements. The discrepancy shall be documented in accordance with the problem reporting requirements of the supplier's or contractor's program, and shall be con-trolled in accordance with paragraph 4.4 of this procedure.
(
4.1.6 In some instances, items or activities which appear to be discrepent should not automatically be considered as non-confoming but should be handled by established procedures and practices. The following examples are given for cla-rification:
4.1.6.1 A deficiency is noted during a construction pro-( tess, and checking and correcting are part of the routine normal course of work prior to sign-off and acceptance.
4.1.6.2 Nonnal apairs involving expected deterioration or wear; however, such repairs shall be documen-ted.
4.1.6.3 If, prior to the completion of a receipt inspec-tion of items that require quality assurance, it is observed that the received items are damaged or do not conform to the purchase specification, such items may be rejected and returned to the supplier.
4.1.7 Discrepancies which may be reportable shall be promptly reported to the supervisor retponsible for the item or activity. Discrepancies which affect operating plant equipment shall be innediately reported to the Shift Fore-man by the individual identifying the discrepancy. With-in one working day from the time that a discrepancy is identified, a discrepancy report shall be initiated and submitted to the supervisor responsible for the item or acti vity. The supervisor shall innediately review the report for potential reportability to the Nuclear Regu-l
Procedure GCP-12.1 Rev. 3 Page 4 of Il latory Comission, and to deterritne classification of the discrepancy as a minor variation or a nonconformance. If the discrepancy is detemined to be potentially report-able, a Nonconformance Report shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with paragraph 4.2. Additional cri-teria and guidance for assistance in determining reporta-bility and classification of the discrepancy are included in the attached Nuclear Plant Administrative Procedure C-11.
4.1.8 A discrepancy which is not reportable and is not a noncon-for: nance shall be processed on a Minor Variation Report in accordance with paragraph 4.3 of this procedure.
4.1.9 A discrepancy which is determined to be a nonconformance or potentially reportable shall be processed as a Noncon-formance Report in accordance with paragraph 4.2 of this procedure.
N1.10 The Senior Site Representative or his delegate (Assistant Project Superintendent or Lead Startup Engineer) shall be promptly notified of all nonconformances that are de-temined to be potentially reportable. If the nonconfor-mance affects a plant or unit holding a construction permit, the Senior Site Representative or his delegate shall notify
( the Project Manager. If the nonconfomance or potentially reportable item affects a plant or unit with an operating license, the Senior Site Representative or his delegate shall promptly notify the Plant Manager, Plant Superin-tendent or Power Plant Engineer.
4.1.11 No work shall be performed to correct a discrepancy until a discrepancy report is approved, except during emergency situations. Emergency situations are defined as those situations in which a hazard to life or property exists.
In these cases it is pemissible to perform the work, then initiate the discrepancy report as soon as possible.
The discrepancy report must state the reason that work was perfomed prior to approval.
4.1.12 Minor Variation Reports and contractor discrepancy reports shall be approved by the Senior Site Representative. Non-confomance Reports shall be approved by a Technical Re-view Group. The Technical Review Group shall, as a mini-mum, consist of Quality Control Quality Assurance, and the Senior Site Representative or supervisor responsible for or affected by the item or activity. If the noncon-formance affects a plant or unit with an operating lic-ense, the chaiman of the Technical Review Group shall be a technical representative from Nuclear Plant Operations.
Procedure GCP-12.1 Rev. 3l Page 5 of 11 4
4.1.13 The disposition for Minor Variation Reports shall be es-tablished within 30 calendar days of initiation, with the exception that the disposition of Minor Variation Reports initiated as a result of Quality Assurance Depart-ment Open Item Reports shall be established within 15 working days of the date the Open Item Report is issued.
The Technical Review Group shall convene within 5 calendar days of the date a potentially reportable r.onconfomance is identified, and within 30 calendar days for all other nonconformances, to detemine the cause of the nonconfor-mance, establish the resolution and the corrective action to prevent recurrence, and assess rtportability.
4.1.14 Quality Control shall maintain logs of Nonconfomance Reports and Minor Variation Reports on the Nonconfomance Report Log (General Construction Fom 77G-54) and the Minor Variation Report Log (General Construction Fom 776-55). The logs shall contain the following infomation:
4.1.14.1 P G and E discrepancy report number 4.1.14.2 Contractor's discrepancy report number (if applicable) 4.1.14.3 Unit (MinorVariationReportLog) 4.1.14.4 Specification
.k 4.1.14.5 Originator 4.1.14.6 Date originated 4.1.14.7 Brief description of discrepant or rejected item 4.1.14.8 Date forwarded to Quality Assurance (Nonconfomance Report Log) 4.1.14.9 Date closed (Minor Variation Report Log) 4.2 Nonconformance Reports shall be completed as follows:
4.2.1 A four part, eleven-character description code shall be used to uniquely identify each Nonconformance Report.
4.2.1.1 The first part shall identify the applicable plant / site:
DC0 Diablo Canyon, General (affecting all units)
DC1 Diablo Canyon, Unit 1 DC2 Diablo Canyon, Unit 2 HB3 Humboldt Bay, Unit 3 SPO Stanislaus Nuclear Project General (affect-ing all units)
NGO Nuclear Plant, General
l Procedure GCP-12.1 Rev. 3
. Page 6 of _ 11 4.2.1.2 The second part contains the last two digits of the year.
4.2.1.3 The third part identifies the department or organization most appropriate to resolve the nonconformance:
GENERAL OFFICE Encineerino General EN Civil Engineering CE Design Drafting DD Electrical Engineering EE Engineering Quality Control EQ Engineering Services ES Mechanical and Nuclear Engineering ME Engineerino Research ER General Construction GC Materials ML Meteorology Office M0 Nu, clear Power Generation General NG Nuclear Plant Operations NO Nuclear Projects NP Quality Assurance QA Sitino SI SITE Bioassay (Encineerino Research) BE Plant Staff General (Nuclear Plant Operations) PG Chemistry and Radiation Protection TC Electrical Maintenance EM Instrument and Controls TI Materials Facility MF Mechanical Maintenance MM Nuclear Engineers TN Operations OP Quality Control QC Security SE Technical Support ST
PrCcedure GCP-12.1 Rev. 3 Page _
7 of n Station Construction General SC Resident Civil RC
. Resident Electrical RE Resident Mechanical RM Resident Startup R$
4.2.1.4 The fourth part consists of the letter "N",
followed by a three digit number which is ob-tained from Quality Centrol. These numbers shall start at 001 each January first and in-crease sequentially through the year.
4.2.1.5 These codes will be expanded as codes are assigned to new projects.
4.2.2 The general system or activity involved shall be described.
4.2.3 The basic documents which contain the requirements which were not met, and any other appropriate documents shall be identified.
4.2.4 The discrepancy shall be described in sufficient detail l
to identify the problem.
( 4.2.5 The originator shall identify his department, date and sign the fom, and forward it to Quality Control.
4.2.6 Quality Control shall distribute infomation copies to the Manager, Nuclear Plant Operations, Quality Assurance, the Plant Manager and/or Plant Superintendent. Nuclear Projects and the appropriate Engineering discipline chief if Engi-neering is involved, and other departments as determined by the Senior Site Representative. When Station Construc-tion is identified as the department responsible for the item or activity, Quality Control shall retain the origi-nel report until the Technical Review Group is convened.
If a department other than Station Construction is respon-sible for the item or activity, the original report shall be forwarded to that department and a copy shall be re-tained by Quality Control.
4.2.7 The Technical Review Group shall convene to detemine whether the discrepancy is a nonconfomance. If it is not a nonconfomance, the basis for the detemination shall be entered under " Resolution" and a Minor Variation Report shall be initiated. If it is a nonconfomance, the group shall detemine the disposition, evaluate, and approve the report as follows:
4.2.7.1 A brief explanation of the cause of the noncon-
Procedure GCP-12.1 Rev. 3 Page 8 of Il fomance shall be stated. In some cases the explanation may involve a study which can be referenced. If the cause cannot be specifically detemined, the most probable cause(s) shall be described in this section.
4.2.7.2 The resolution of the nonconfomance shall be established. The resolution may be described as one of the following four options: accept as is, rework, repair or mM/.fy, or reject with specific amplifying infomation. A resolution which accepts a discrepant condition "as is" shall include a basis for the acceptance.
4.2.7.3 The corrective action necescary to prevent re-currence of the nonconfomance shall be described.
If it is detemined that no corrective action is required, then the basis for this decision shall be stated.
4.2.7.4 The proposed schedule for completion of the resolution and corrective action shall be entered.
4.2.7.5 The nonconfomance shall be reviewed for repor-tability under Title 10. Code of Feeral Regu-lations. Parts 21 and 50.55(e). In addition, k if the nonconfomance affects a plant or unit with an operating license, the nonconformance
. shall be reviewed for reportability in accordance l with the requirements of the attached Nuclear Plant Administrative Procedure C-11. The basis for reporting or not reporting shall be entered.
If reportable, the applicable time requirement for reporting, the method of reporting and the time of the report shall be entered. Infomation regarding the actual report and follow-up report shall be recorded as it is received fr9m the f reporting department.
4.2.7.6 All decisions of the Technical Review Group must be unanimous. If unanimous decisions cannot be reached, the matter shall be referred to the Manager. Quality Assurance Department.
4.2.7.7 The Technical Review Group chaiman shall notify the Plant Manager and/or Plant Superintendent and the Senior Site Representative or his dele-gate of any items or activities that are deter-mined to be reportable.
4.2.7.8 The Senior Site Representative is responsi-ble for notifying the Project Manager and -
all appropriate department managers of all i I
- Procedure GCP-12.1 Rev. 3 Page 9 of 11 reportable items or activities affecting a plant
- sr unit holding a construction permit. Nuclear Plant Operations is responsible for notifying the Nuclear Regulatory Comission of all report-able items or activities affecting a plant or unit with an operating license.
4.2.7.9 The chaiman and members of the Technical Review group shall sign the report approving the tech-nical content of the disposition. The Quality Assurance representative shall sign the fom verifying that the disposition is acceptable with respect to the Quality Assurance Program.
After all the required signatures of the Tech-nical Review Group are obtained, the report shall 1 be filed by Quality Control until the resolution !
and corrective action have been completed.
4.2.8 Upon completion of the resolution, the " Resolution Completed" section shall be signed and dated by the implementing organization.
4.2.9 When inspection is re. quired to verify that the completed resolution is acceptable, the individual perfoming the i inspection stall sign and date below " Resolution Completed".
4.2.10 Upon completion of corrective action, the implementing C organization shall sign and date the " Corrective Action Completed" section.
4.2.11 When the corrective action requires inspection to verify implementation, the individual perfoming the inspection shall sign and date below " Corrective Action Completed".
4.2.12 All clarifying or verifying documentation shall be complete and attached or referenced. The report shall then be for-warded to the Quality Assurance Department for verification and distribution. .
4.2.13 If a Nonconfomance Report must be revised, a new Noncon-formance Report shall be written.
4.2.13.1 The new Nonconfomance Report shall reference the
! one it supersedes under " Reference Requirements".
L
- 4.2.13.2 An entry shall be made in the " Additional Remarks"
- section of the superseded Nonconfomance Report referencing the number of the new Nonconformance Report.
l 4.2.13.3 The super:eded Nonconfomance Report shall then be processed as a completed Nonconformance Report.
4.2.13.4 The new Nonconfomance Report shall be processed in accordance with paragraph 4.2.7 of this proce-dure.
I
F Procedure GCP-12.1 Rev. 3 i Page 10 of 11 4.3 Minor Variation Reports shall be completed as follows: l 4.3.1 The initiator shall complete the heading to include the l location, unit number, specification number, p9ge number, responsible organization, contractor's disempancy report number (if applicable), and P G and E report number (ob-tained from Quality Control). If a contractor is to re-ceive a copy, the contractor's name shall be entered on the forin in the space provided.
4.3.2 The initiator shall complete the " Description of Discre-pancy" section with a description of the discrepant item, explanation of the discrepancy, signature and date.
4.3.3 The " Disposition" section shall be completed with a pro-posed disposition. A disposition which accepts a discre-Fant condition "as is" shall include a basis for the accep-tance. The report shall then be submitted to the Senior Site Representative for approval. If desired or required, additional concurrence with the disposition shall be ob-tained from the ASME Authorized Inspector, Assigned Engineer, supplier's representative, etc.
4.3.4 The Senior Site Representative and Quality Control shall review the report for potential reportability and for classification of the discrepancy. If the discrepancy
( is determined to be reportable or a nonconformance, the Minor Variation Report shall be superseded by a Nonconfor-mance Report. The Nonconforinance Report shall be processed in accordance with paragraph 4.2 of this procedure.
4.3.5 Quality Control shall maintain the original report until the disposition has been accomplished. i 4.3.6 Quality Control shall be notified when all requirements of the disposition have been verified as being satisfac-
! torily accomplished. The " Disposition Accomplished" sec-tion shall be completed, signed and dated by the verifying individual. Minor Variation Reports issued by Quality Control shall be verified by Quality Control. If applicable.
l the Contractor's completed discrepancy report shall be
' attached. Any other verifying or clarifying documents shall be referenced or attached. All attachments shall be identified in the space provided at the bottom of the re-port.
4.3.7 The report shall then be forwarded to Quality Control for review, to assure that the " Disposition Accomplished" add-resses all the requirements of the disposition, and that the attachments listed are attached and complete. The original report : hall be filed by Quality Control.
. . . ~ . , , . - . - - . . , _ , - . , , . - - - . . , , . _ , . - . - . . - _ - . ..._- - - - --..__ -
D
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Procedure GCP-A2.1 Rev. 3
( ,
Page 11 of 11 4.3.8 If a Minor Variation Report must be revised, a new report shall be written. The original report number shall be retained and the revision number noted on the report and
' in the Minor Variation Report Log. A revision to a Minor Verlation Report shall receive the same approval, review and distribution as the original.
4.4 Contractor'.s discrepancy reports shall be submitted to the Senior Site Representative for approva) of the problem description and s concurrence with, or detemination of, the disposition. The Senior Site Representative and Quality Control shall review the s y report for potential reportability and for classification of the
.', discrepancy. If the discrepancy is detemined to be reportable s or a nonconfomance, a P G and E Nonconformance Report shall be 4 initiated in accordance with paragraph 4.2 of this procedure.
g s 4.4.1 A Minor Variation Report may,be initiated to provide add-itional control of discrepancies resulting from contractor violations.
4.4.2s' Site procedures shall be developed in accordance with s Procedure GCP-5.1 to control the processing of contrac-s tors' discrepancy reports.
x
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( 5.0 00CUMENTAJ13 ,
w.~
u S.1 All .documentatics Jequired by this procedure, with the excep' ion of completed Nonconfomece Reports, shall be maintained by (pality Control for inclusion in'%e Records Management System. Completed Nonconfomance Reports shiit' be maintained by Quality Assurance,
- t i, 9
'6.0 R_!FERENCES ; , ,
G.1 Title 10. Code 'of Federal Regulations. Part 50. Appendix 8
" Quality Assurance Criteria for Nuclear Power Plants" Criteria
_; XV and XVI. ,
6.2 American National Standards Institute N45.2, " Quality Assurance Program Requirements for Mu: lear Power Plants", Sections IS and 17.
t
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7.0 ATTACHMENTS 7.1 QualityAssuranceFoh76-286NonconformanceReport s 7.2 General Construction Fom 776-101 Minor Variation Report i
- G, 7.3 General Construction Form 77d-54, Nonconfonnance Report Log i
- 4 N\ -
\ .
s 7.4 A Ge neral Construction Form 776e55',N.r.or' Variation Report Log I'
)'s 7.5 ' Nuclear, Piant Administfative,Proce' dure NPAP C-11. "Non-Routine Notification and Reporting to the ' Nuclear Regulatory Comission
'y , (NRC) and Other Governmental Agencies" ~
, N sy
\, 1 yr
l l
l EXHIBIT 14 a Las. onesee oneer scaerJ 7,* ,ess.Tremme, sensent.Jss==ry is.tses og A test for PG&E .
l Worker, firm haggle over. record i
,, _ ~ o sr st.te , s - m. - p _ .
aus m eer a.a# and naean aemsey te e., ght said as test shese is that he is i
- vesarMs regemt h was me me. a mg addha, seM , ww i an emoispe='et me osehle (he. But stats OsNA am*N= are me- anM he danskaty is ad to pas seeinse power plant wants, men whosher the federal ins ap- drugs.
assess to ressede af a psychstegkset pIhs to psychstaghre resorts as Cooper emed be heMeses hoe f test he was furend to take k seder wsEas eher medlealneeds, PGas and Eshevtsedyne went to to homeads job.
. A nand roung from osRA'salkr- , hus, Ibs test rendes seder wraps' l se tar, John Cooper, a Gold ; anye shut how toim6erpret amlaw bessess *1r his ever gas ed R .
l maham is asseral esmatrudesa et :is espesses this ved er assa. womad enase a his ruskes and'-
l De pised, has been domind pesude. AM PG&E pond ler the pay. weeld ammherreas Embawlestyen." ,
i asco te see reentes of the test emetessent testias, nehavnerdy== cesser, e, s ,eme d d c d rety adsdehtered by Bahndantres, a vem Os nesta Eshsviergan's yndeste is electru6en engenser.
, payr 4ialagial testing Arm in Pale peber is tied test reenlis esa be in6, seM he has worked at the plant Alan, relanned enty to licamord prehnte- sit and en for about fear years.
- Poeific Gen and Elor:trie Co. - gists, sold PG&E epokeswomas No anM sehst suplayees have reentres en 134ette es> Summune G.Brees. complatsed edetty aheet the tests' sept thans wher, have timer. Sie said gestenstemal esNas et het aser coesig any athr that salpit j tasen esasecutiva ymere to take the - ' ' 3 testers prevent test cast thesa their pehm, ,
test A proposed federal sug=3* Man teamits front be given out te enid est W PGAR doesn't - l
, foretas anclemt power planta te these whoare are he'R gaat am bin own seen '
l she sed tests is espected to he ~Psydielssual test remdh ese me mov intc russistisse are adaptes seen hr the Nestser Resm- to ateismens to seasses who to proved. These tiens e .
neeery coussession. est fasamer wim the suomess" et regstre meeleer owners to .
In an sehrt to pry spen the fue ,,1 '- -* terminaissy, neewn es imister the tants and also re.
a his tad resmik, Ossyer is espisland. quire sequerwhereis keep an eye as l to use a 'ame and hthe amed Ceeper sold be spoke wie a septsynes to weed est these wla law thed sives esaplayess esposed psychelesist at Diable who an. pstealian hehoviment prehlems. .
to redistles er tense chemiente pleksed parts of his test renden. "IN, already told any hems's bass rights to see their enamant files "Sesas er the that were said ebend zay aremplaaman" said Osep.
Oneper seM he asey have.heen dartug may were just et, er. "I wee't wert for a sempegy emyssed to redsstima and is asking dandous," Cooper sold, that spies en Its own sniploymen." L 6
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l UNITED STATES OF MERICA Ci f'-
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NUCLEAR REGULATORY C0l#41SSION BEFORE THE ATOMI_C SAFETY AND LICENSING APPEAL BOARD
- 7. ,
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In the Matter of ) Docket Nos. 50-275-
/ - )- 50-323 .
I PACIFIC GAS AND ELECTRIC )
~,.
C0lFANY' ) (Design Quality Assurance)
. )
(Diablo Canyon Nuclear Power )
l , Plant, Units 1 and 2) )
)
p AFFIDAVIT OF R. C. ANDERSON, H. J. JACOBSON, M. E. LEPPKE, L. E. SHIPLEY
< lc
, STATE OF CALIFORNIA '
)
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( /f < CITY AND COUNTY OF SAN
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O FRANCISCO )
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t ' The above, being duly sworn, depose and say:
I, Richard C. Anderson, am Engineering Manager for the Diablo Canyon l IP },.J --
tr (j -f Project. j 1
.+
I, Michael J. Jacobson, am Project Quality Assurance Engineer for the L
l i, l Diabloj Canyon Project.
I..Myron E. Leppke, sm Onsite Project Engineer for the Diablo Canyon i
F.Sject.4-o s j
- s ~
,I, Le,rgy E. Shipley, am Technical .Consulta", for Piping for the Diablo e '
Canyon Prdject.
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. . . . - . =-------------------==------------------------------------
By letter dated February 7,1984 (PGandE Letter No.: DCL-84-046) we
( forwarded to the NRC a response to questions raised as a result of the recent i
NRC investigation into allegations regarding small bore piping design (attached hereto as Exhibit 1). We supervised and participated in the preparation of this response, and it is true and correct to the best of our knowledge, infomation, and belief.
Dated: March ,1984 "e = n R. G. MDERSON WO Mn M. J. J ACOB5Uff l
( h{
M. E. LEPPKE
- ~
p.I.5HIPLr.YoO '~
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Subscribed and sworn to before me this 8i d day of Harch,1984.
W '
SEAL
' l(ancy NotaryJ. LedasteW'd Public in an f9r the City and County of San Francisco, State of California.
My comission exp't es April 14,1986.
j@o00000000cxxxxxonco:X>xcoercot
- NANCY J. LEMASTER I [h NOTMY PU2tIO-CAllFORt41A y *
%?,s. f CITY AND COUNTY OF '
+ SAN FRANCISCO
( f:X: C:XXXXXXXXX:0:XXXXXXTccCMy Commission Empires Ap it 14,1986
........-..a-.-.... . . . . . :. . . . . .. ..
List of Exhibits
(
Exhibit 1. PGandE letter dated February 7,1984.
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PACIFIC CAO AND ELECTRIC COMPANY , , , , , ,
Exhibit No.1 .
I February 7, 1984 .
PGandE Letter No.: DCL-84-046 Mr. Darrell G. Eisenhut Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 l Re: Docket No. 50-275, OL-DPR-76 Diablo Canyon Unit 1 Small Bore Piping
Dear Mr. Eisenhut:
During the .ecent NRC investigations into allegations listed in SSER 2), the Staff raised several questions with respect to the design of small bore piping. These questions were discussed by the Staff at the Diablo Canyon Power Fiant exit interview on January 19, 1984 and at the January 31, 1984 I meeting in San Francisco between the NRC and PGandE.
The Staff questions and PGandE's responses are documented in the enclosure.
Kindly acknowledge receipt of this material on the enclosed copy of this letter and return it in the enclosed addressed envelope.
Sincerely, l
J. O. Schuyler by J. D. Shiffer
! CCWu/BSL/JDS/JOS:naw l
Enclosures cc: T. W. BiJhop G. W. Knighton J. B. Martin H. E. Schierling bec: Diablo Distribution 0174d/0007K i
- . 2
- e PGandE Letter No. DCL-84-46
- b. ENCLOSURE I. INTRODUCTION
- q. 1. General
.~
This suistittal is provided in response to questions raised as a result of the recent NRC investigation of allegations regarding small bore piping design by I the Onsite Project Engineering Group (OPEG). Thts submittal sets forth the I questions raised, responses to those questions, conclusior.s. and if applicable, the corrective action being taken by the Project.
To prepare this submittal, the Project reviewed the information developed by the NRC investigators and noted the explanations and conclusions provided by the investigators during exit meetings and the public meeting of January 31.
After investigating the facts giving rise to thNe concerns, basic causes of discrepancies and generic implications were carefully considered. Conclusions have been derived as to adequacy of the design, effectiveness of the quality assurance program, and needs for corrective action and for strengthening the program.
The questions appear to encompass the following issues:
o Adequacy of small bore pipe design
( , o Effectiveness of the quality assurance program for OPEG o Generic implications from discrepancies found o Corrective actions which might be necessary or desirable It is important to recognize that none of the evidence demonstrates that there were inadequate designs or that the overall quality assurance program was ineffective. At most, concerns were raised which create a need for additional information to provide requisite levels of assurance. The investigation also identifies where improvements are desirable in Project programs and practices.
- 2. Nature of Concerns The concerns raised cover a wide range of small bore piping design activities
- that are more thoroughly explained and evaluated in the individual sections or subsect-lons to follow. However, it is possible to provide some statements and perspectives regarding the review effort:
- a. Discrepancies have been found in the small bore piping design work.
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- b. Such discrepancies are of a minor nature and, when revised &
calculations or analyses were performed, all of the piping and
(- supports fully met tht licensing criteria and commitments. Thus,
'E -.
it can be concluded that there is no technical or safety concern %
with the ar designed and constructed safety-related small bore ;
pfptng.
I. c. The presence of these discrepancies raised concerns regarding the control of the engineering work within the OPEG small bore piping ?-
group and its overall level of quality. Such concerns have been addrussed by the explanation and discussion given for each specific i-concern and by the corrective action being taken, p.
The major corrective action to date involves the review of 110 i
- d. l' small bore pipe support analyses: 57 of the more complex "-
(computer analyzed) safety-related small bore pipe support designs; 25 of the simpler (band calculated) small bore pipe supports; and f the 28 calculations identified by the NRC during its f~
i; investigation. Additionally, certain strengthened training and procedural requirements and commitments have been made. ,
- 3. The OPEG Organization The OPEG is organizationally a part of Project Engineering, but is located at 4 the site and thus physically separcted from the San Francisco engineering group. It was established to meet construction's need for expeditious responses from design engineering, to provide more direct feedback to design - -
( engineering on construction and startup matters, and to perform certain engineerinil activities (e.g., small bore piping design) that are best perfomed tn proximity to the physical plant. The OPEG group functioned with substantial autonosty, because of the need for close-coupling with site construction and operations, and because its scope was rather closely ;
defined. This was intended to make it more responsive to a need for on-the-spot resolution of problems.
The scope of OPEG's responsibilities is limited by Engineering management to -
matters within its capabilities, considering such factors cs staff support.
number of people. This '
facilities, abilities of assigned personnel, andTypically, the work performed by OP scope is clearly set forth in writing.
includes design of Class I small bore pipe and supports, limited resolution of physical interferences, resolution of non-conformances, and assistance in startup problems. It serves the needs of both Units 1 and 2. By far, the greatest proportion of its work is related to design of small bore pipe and -
supports. No other major design work or analysis was performed by OPEG.
The OPEG organization is headed by an ontite Project Engineer, reporting to the Project Engineers for Units 1 and 2 and receiving assignments from them. '
The number of people has varied widely, ranging from several dozen, up to almost 300. Because of the unique requirements of this group and the nature Q j !
of their work, more than 50% of its technical personnel were comprised of l'
' 2
- di 9i ll
non-permanent engineers provided by contract firms. The engineers are, f ., however, carefuly screened for technical competence by PGandE or Bechtel, and U by the contract fire prior to hiring.
- 4. Conclusions
, It is clear that the results of the reviews completed to date establish that
. there is reasonable assurance that the as-constructed small bore piping meets all design requirements and, thus, poses no safety concerns. Strengthened controls will minimize recurrence of similar issues.
Specific conc 1'usions are as follows:
o Based on reviewing a sample of 110 piping support designs, it is coacluded that final designs were not affected by the number of approximations and minor mistakes in the calculations of pipe supports and reasonable assurance of the adequacy of small bore plaing design does exist. It should be noted that as of the time of this su mittal, 6 of the 110 support analyses are not complete.
o Because of the unique features of the OPEG Small Bore Piping Group (e.g.,
work scope and how it functioned), there is no reason to believe that similar concerns extst elsewhere.
o Compliance with NQAM requirements, including numerous audits, plus the lack of significant errors, show tha engineering quality assurance program was effective, but would benefit from strengthening in areas of
(, training, technical audits, and procedure control.
o Perjorative charges in small pipe design work cannot be supported.
( . _ . . . . _ - . __ _ ._ _ _ _ _ _ _ _ ._
II. TECHNIC /d. ISSUES
(' IRC Question: The NRC has raised a question about Code Break designations (Allegations 55, 86, and 88. SSER 21). This matter was further addressed by Dr. Hartzman at the public meeting on January 31, 1984.
. Response The tem " code break" is used to describe the section of a ipi g system where the safety-related piping (Class I) changes to nonsafety-re ated (Class II) piping (see the figure below). This " code break" section is always located on l the C' ass II piping and starts at the valve which is the point at which the !
- fluid system class changes from Class I to Class II. Within the " code break" '
section is a system of supports or an anchor that dynamically isolates the i Class I piping from the remainder of the Class II piping. The " code break" l section of the pipe ends when dynamic isolation has been accomplished. The criteria used to achieve the desired isolation, as discussed in the Phase I Final Report, require that the system of supports that provides dynamic ;
isoiation be made up of either: (1) an anchor or (2) at least two lateral supports in each direction and one axial support. The anchor, or supports, are denoted as Class II* supports and are designed to the same criteria that are used for Class 1 supports.
Class I = Safety-related
" Code break" section Class II* = Nonsafety-related
>< A but supported to achieve isolation of the Class I piping VALVE END OF
(. CODE BREAK (" Code Break" section)
Class I _ Class II* _ CJass II- Class II = Nonsafety-related
~ ~ ' -
nonseismic design In the above schematic, the length of Class II* piping is not important as long as the code break requirements are P:t by providing supports or an anchor. If the length of the Class II* section of piping can be shortened by relocating the Class II boundary closer to the Class I boundary, the system would then require fewer Class II* supports; this relocation is only '
accomplished by adding supports or an anchor to the code break section closer to the Class I boundary. As an example, assume that following the valve, the l code break section included five bilateral supports (these provide support in l
' both lateral directions at one location) and then an axial support. All these L supports would require Class 1 qualification. Two alternatives for improvement of the design that are acceptable and meet all licensing criteria are: (1) to add an anchor at the location of the first bilateral support, or (2) to add an axial support at the location of the second bilateral support.
Both alternatives reduce the length of the code break and the number of i supports requiring Class I qualification.
The allegation that the code break boundaries were relocated in violation of some engineering precept, project instruction, or licensing criteria is fallacious. While it is trua that the length of Class II* piping was -
minimized wherever possible by modification or addition of supports, there is no reason not to reduce the amount of the Class II* piping to the minimum.
( 4-L
s c Question: The RC has raised a question about including es-built gaps to reauce thermal loads (Allegations 55 and 79 SSER 21). This matter was (t '
further discussed by Mr. Yin at the January 31, 1984, public meeting.
Response
. When performing small bore piping stress analysis for thermal expansion or
- ... thermal anchor motion, actual restraint clearances or as-built gaps are sometimes included in the qualification calculations as described in Piping Procedure P-11 (Section 4.6.2). The gaps that are included are physical clearances that exist between the pipe and a structural element. Thermal loads can be eliminated by gaps in pipe supports and, therefore, the inclusion of gaps in the qualification analyses is completely appropriate. In each case where gaps are included to reduce thermal loads, adequate assurance is available that the gap can be relied on to be present throughout the plant lifetime.
Before any gaps were included in a piping stress analysis, Piping Procedure F-11 required as-built reverification. Accordingly, a plant walkdown was conducted to establish the actual gap configurat<on. The gap configuration was modeled and included in the documentation of the stress analysis calculation. This practice of including gaps to reduce thermal loads is used in the industry as a method of accounting for actual plent conditions.
As a result of this NTC question, a review of all small bore piping stress analyses was conducted. The results of the review demonstrated that as-built gaps were included ir 25 piping analyses affecting a total of 64 pipe
( supports. The 64 supports represent about 35 of the supports analyzed. As reported in the Project's supplemental letter to the Staff dated December 28, 1983, 16 of the 25 piping stress analyses involved piping with service conditions below 2000F. In these 16 analyses, thermal movements are minor and not of technical concern. The 9 remaining pipe stress analyses affect only 16 supports (see Table 1) which are less than 15 of all the small bore pipe supports analyzed.
A description of the 9 pipe stress analyses in which as-built gaps were modeled into the computer analysis and the pipina system temperature exceeds 200*F for normal thermal load cases was presented in the December 28, 1983, letter. These 9 analyses fall into two categories. Category 1 gaps were modeled to accommodate thermal anchor movement (TAM) of large bore piping.
$1 hee these gaps are caused by the thermal movement of large pipes and equipment expected to have repeatable thermal growth, the gaps are expected to be present throughout the plant's lifetime. All but one support falls in this category. Category 2 consists of gaps modeled to release thermal loads and stresses induced by two opposing supports restraining the pipe in the same directica. Because of the piping configuration that exists, it is clear that the as-built gaps will remain throughout the plant's lifetime.
b l 1
1
The consideration of actual restraint clearances, as described in the supplemental December 28 letter, is a reasonable and adequate technique for
( . '- the piping geometries involved. This method is consistent with the licensing criteria for Diablo Canyon and has gained widespread use in the nuclear industry where the more conservative approach of ignoring as-built gaps results in excessive thermal loads. Finally, the use of actual restraint clearance involved a very small part of the small bore pipe and supports that j .
were analyzed.
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Table 1*
(.
Small Bore Piping Small Bore Support Corresponding Gap Modeling
_. Calculation No. No. for which Category Gap ) .
(See Note
,. Piping Analysis was Modeled Data Point 15 1 5118 63-7 181-84 550 1 556 1 181-96 3-302A 53-1 50 1 65 1 53-1 53-1 67 1 70 1 3-3028 53-1 198 1 4-302 42-6 20 1 8-310 2152-09
( 8-312 47-19 47-24 100 175 1
1 24 1 8-314A 66-22 44 1 2185-1 66-25 58 1 1
66-24 78 66-51 32 1 34 2 8-328 2157-14 105 1 9-309 181-20 200 1 181-42 NOTE: Category 1 = Gaps were modeled to accommodate thermal anchor movement (TAM) of large bore pipe whose movements are determined to be repeatable.
Category 2 = Gaps were modeled to release thermal loads and stresses induced by two opposing supports restraining the pipe in the same direction.
- Isometrics for this table were previously submitted with letter of December 28, 1983.
( .y.
15tC Question:
The 15tC has raised questions about the use of different
(' snTTnesses for the same rigid supports in static and dynamic pipe analysis (Allegations 55 and 88, SSER 21). This issue was also addressed by Dr.
- Hartzman during the January 31, 1984, public meeting.
Response
(, Piping support flexibility was modeled in 4 of 129 analyses (total number of IE-101 analyses) to more accurately determine the actual system behavior that occurs during thermal expansion of the piping and to reduce calculated thermal loads. The nature of thermal expansion produces only static (displacement limited) loads and not dynamic loads such as the seismic loads. Inclusion of support flexibility in the thermal piping system analysis is.an acceptable mettod of more accurately predicting the load that will be produced at any given pipe support. This approach is consistent with accepted engineering practice.
The Hosgri Report, Section 8.2 states that seismic supports Sincemay thebenatural considered rigid if the natural frequency is greater than 20 Hz.
frequencies of these sup orts are greater than 20 Hz, the seismic analysis considered them to be ri id.
The support itself is qualified for the combined thermal plus seismic loads.
Further, these loads are derived from two totally different loading phenomena: one static (thermal), and one dynamic (seismic).
Even though these calculations have met all licensing criteria, we have
(- reperformed the 4 original analyses mentioned above with the support flexibilities also included in the seismic analysis to demonstrate theThe results appropriateness of the original assumption.
analyses demonstrate that the stresses and support loads and dynamic piping analyses.
In summary, the apparent inconsistent treatment of support stiffness forThe ado static and dynamic analyses is technically justified.
approach was largely dictated by the desire to consistently implement seismic -
licensing criteria which analyze supports as rigid if their natural fr is greater than 20 Hz.
criteria even if support stiffness is included in both static and dynamic analyses.
9
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NRC Question: The NRC has raised a question about computation errors and moaenng sericiencies in small bore pipe support design packaps. These
(- issues were discussed by Dr. Hartzman and Mr. Yin at the publ1c meeting held on January 31, 1984.
Response
i The following response discusses and puts into persp.ctive the calculational errors, the modeling anomalies, the engineering judgments, and the
- documentation inconsistencies found in small bore pipe support calculations.
The analytical approach is reviewed to give perspective as to significance of real and perceived deficiencies. The necessity for precision in small bore calculations is discussed and a summary of the additional review effort that has been undertaken to address the NRC's concern is presented.
Although there are discrepancies in the calculation packages, one must recognize the large number of decisions that an analysi must make, and a '
checker must review,.in a given calculation package compared to the number of '
discrepancies discovered by the NRC reviewers and by our own reviewers. Small bore pipe supports are designed with adequate precision to achieve the design function. Tie primary reason for the acceptability of this level of precision in small bore pipint design is due to conservatisms and structural redundancy in the small bore p< ping and supports completed with the low magnitude of loads which they experience. Nevertheless, the need for originating and checking engineers to more rigorously document acceptance of minor calculational errors is acknowledged.
( Some of the pipe supports reviewed by the NRC inspectors are among the most complex small bore supports in the plant. The discrepancies found in our study of the NRC review actually represent a small percentage of the total number of decisions / actions that must be performed to arrive at a complete analysis. These analyses have been reviewed by the Project in detail and it has been determined that no modifications are required as a result of the discrepancies. This review is described below. The fact that no modifications were required confirms a conclusion that the design process and conservatisms are tolerant to minor anomalies and that the engineers responsible for the design of supports nave ensured that significant errors do not exist. ,
- a. Pipe Support Design Process In the case of frame structure supports, the design generally consists of two phases. The first phase consists of the analysis of the frame ,
structure and the second phase consists of the analysis of the associated bike plates. Associated steps include evaluation of welds and qualification of standard components (struts, snubbers, U-bolts, etc.).
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During the analysis of the frame structure, the analyst must translate a 5 7- 5 y f support drawing into a three-dimensional re>resentation describing the
- L.: placement, orientation, and properties of tie steel members and the directions and combinations of the applied piping loads. Upon completion 3-s
? the analyst must perform a final check of the overall results to assure
- compliance with derign criteria. 3 5
on consists j -
A of,moderately complex for example, small bore approximately 1 gipe support iscrete steelatstructura Diablo Camembers and 11 J
4 connections. In addition, the support has many supplementary items such 5 r as U-bolts or other small members which act to restrain the pipe. The 'i model eventually developed by the engineer will contain approximately 30 9 3- g 5 joints and 25 elements. To develop the model, the engineer has had to specify 30 directional (x-y-z) coordinate points and define the =
s
- - connectivity of the elements to these joints. This means ensuring that $
p: approximately 90 numbers are correctly calculated, all digits and signs j
': are correct, and indicatint the proper numerical combinations to define 3 member connectivity are inc icated. Also, the engineer has to indicate 7 the orientation of the strong and weak axes of the member. When the i T analysis is completed, the engineer applies the loads to the support {;
g model.
_l T -
Typically, small bore supports are bilateral (supporting the pipe in two directions) and many are gang supports (supporting two or more pipes). 4 For example, consider a frame that acts as a support for two pipes. j
.A: Given the nuut>er of loads that must be specified (deadload, tributary ;
P: mass loads, normal and accident thermal loads, and three different a i_
T seismic loads), one arrives at a total of 32 individual loads that must 5
(. be correctly transferred from the piping analysis, including directional ?
f
-;; sig n. Also, he must specify parameters, such as unbraced length, for 3 coce checking purposes. The engineer then submits the input for computer =
_7 s 3 analysis. Upon receipt of the computer analysis, the engineer reviews the cuput for appropriateness of deflections and stresses. Up to this 3 point, the engineer has had to correctly develop and specify at a minimum 3
-y y .. approximately 300 numbers, assuring that all digits and signs are :
- correct. In addition, he has had to review numerous pages of computer 5 output. q a
e After the engineer has completed his frame analysis, he must now begin 1 k- the task of analyzing the base plates. For the evaluation of base y
= plates, the analyst must similarly deal with hundreds of numbers or e combinations of numbers. The engineer must choose from the many load 5 4 combinations the sets of forces and moments to be input into the d 4 i E plate / anchor bolt analysis. The local coordinates of the baseplate model
,i most be correlated with the local / global coordinates of the frame model. gi i The plate size, thickness and shape, in addition to anchor bolt location. :
=
must
? stiffness, capacity, spacing, and derated capacity edge distances,lt to r- also be reviewed and input. Taken as a package, it is not difficu j 2 conclude that the engineer in the above discussions has had to deal with
!7 and review up to 1000 numbers. l I ( d i
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i 3 _______ _ . _ _ .
The judgment and capability of the engineer throughout the design process helps assure a safe design. His engineering training, experience, and
- (_- insight are important in visualizing the model and loading conditions, as well as deciding that the results are acceptable. The engineer is responsible for assuring that the support design is free of significant error by applying his experience from performing analyses of many pipe supports.
Additionally, the reviewing engineer provides an important function in assuring that major errors do not exist by applying his general experience in evaluating the final piping system. The small size of
~
these couponents allows good visualization and a heurestic understanding )
of the adequacy of a design, even without formal calculations and '
analyses. The engineer's understanding and experience lead to the ,
identification of any major error by observing any obvious '
inconsistencies such as undersized members from that provided for other ,
pipe supports. l The broad responsibility of the reviewing engineer is to assure that the calculation is sufficiently accurate for its intended purpose, i.e., to document how the support meets the design requirements. Therefore, minor discrepancies in areas of the calculation that would not lead to a criteria exceedence would not be expected to be documented. The fact that when the discrepancies were addressed the supports were acceptable without modificatioa substantiates the adequacy of the design process.
Nevertheless, discrepancies uncovered should have been documented..
- b. Documentation of Small Bore Support Design
(
There are approximately 4000 small bore pipe supports which were designed and qualified in the field. Of necessity the process used to design and aualify these supports was a production-oriented process. The flow of Work required a receipt of a set of loads and displacements, design of the support, preparation of design calculations, checking of the design calculations, and review and approval of the as-built drawings. Both the originator and reviewing engineer focused on the parameters of primary importance to the adequacy of the support. Although satisfactory for criterion and safety considerations, the level of rigor associated with these supports was different from that achieved in other parts of the plant. In general, this variation in rigor is clear to thNe familiar with design practices in power plant and industrial plant facilities throughout the country. More importantly, the rigor of design documentation varies according to (1) the importance of the system, (2) the decree to which the system design may be challenged (large loads vs. ,
suull Toads), and (3) the conservat<su which exists <n the design.
The level of rigor of the small bore design documentation was technically consistent with the number of supports and the conservatism and structural redundancy inherent in the designs; however, compliance with quality program documentation was less than fully achieved in some instances.
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- c. Design Characteristics b The previous section described the design process and the conservatism inherent in small bore design. The fact that the margin is very large for this class of pfpintl 1s often discussed but its importance must not be underestimated. W 1 bore piping is fabricated from asterials with
., duct 111 ties into the 305 to 405 range (resulting ductfifties from the 1 design analyses are typically less than 15). The supporting systems provide for a highly redundant set of supports in which deflection of an individual support results in the transfer of load to adjacent supports.
Additional conservatisms exist and are frequently tabulated in the methods used to calculate small bore loads on supports, especially when span tables are used for calculating stresses in the supports. The l
result is that the small bore pipinfl system and supporting structures are highly conservative in design and h< ghly insensitive to variations in the details of individual support designs.
- d. Review of Supports A significant number of small bore pipe support calculation packages have been reviewed in detail. Some were reviewed prior to the January 31, 1984 meeting and many have been reviewed since then. The IDVP reviewed a total of 19 calculation packapes as documented in ITRs 60 and 61. The Project has reviewed 110 smal bore pipe support analyses: 57 of the more complex (computer analyzed) safety-related small bore pipe designs; 25 of the simpler (hand calculated) small bore pipe supports; and the 28 calculations identified by the NRC during its investigation.
( This Project review has been conducted to reverify the adequacy of the small bore piping design and to define the necessity for further improvement in documentation of the design adequacy. Each calculation package has been subjected to a detailed engineering review by the Project to identify all possible deficiencies or errors. This review
' has, of course, been far more rigorous and detailed than that performed in the original checking process.
Each of the selected calculation packages was reevaluated by a reviewer and reconfirmed by a checker. A checklist was used to aid in the review process. Results of the review were documented on the checklist and supplemental comments sheets, if required.
l The reviewers verified that the structural model was adequate and couplete, that the loads used in the calculations were properly applied, and that the structural model reflected the latest as-built drawing.
Calculations were reviewed for required documentation, such as weld calculations, anchor bolts, base plate, spring variability, frequency, and structural analysis, to demonstrate compliance with appropriate project criteria, procedures and instructions.
The results were summarized into three categories.
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The first category, " Hanger Acceptable As Is or With Minor Suppienental l f-Calculations or Comments " is used to indicate those support calculation
\.- packages that were found to contain complete and acceptable information or to indicate those support calculation packages that were found to be acceptable, but which, for example:
, . (1) Lacked certain statements needed to document the conclusions
- - reached.
(2) Did not contain documented evidence of the evaluation of certain items which the reviewer felt was prudent to include in the calculation package.
(3) Contained information from which the reviewer could not make an assessment and thus deemed it ne:essary to perform supplemental calculations in order to support his evaluation and conclusions.
It is not surprising that, due to the detail in the review, minor supplemental calculations or comments were required. Other engineers, rigorously looking after the fact, will generally always comment on some aspect of someone else's design calculation.
The second category " Hanger Acceptable With Detailed Calculations " is used to indicate those support calculation packages that were found to be acceptable, but where, for example:
(1) The reviewer believed that it was advisable to perfom additional analyses or modify and rerun the existing computer analyses.
(
The term " Hanger Acceptable" indicates acceptability to the design criteria which were originally used to qualify the supports. The methods and criteria were not modified for this evaluation. Highly sophisticated analysis, such as plasticity calculation, was not used to qualify any of these supports.
The last category. " Hanger Unacceptable " is used to indicate those support calculation packages that were found to contain errors which, upon reanalysis, showed that the hanger required modification.
There were 129 support calculations included in the review. The results are as follows:
Category 5 of Supports Acceptable with Minor Supplemental Calculations or Comments 785 Acceptable with Detailed Calculations 175*
Unacceptable 05
- Detailed calculations for 6 supports (55) have yet to be completed.
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These results are significant. Of the 129 small bore supports, some among the most complex in the plant, the fact that no modifications were C: required indicates the minor ispect of the anomalies noted.
It is also interesting to characterize the discrepancies themselves. The discrepancies noted in the review were tabulated into one of three categories. These categories error (2) modeling were or engineering (1) modeling,ified judgment (ver by subsequentinput, or calc r_ calculation),and(3)documentationdiscrepancy.
The first category includes such items as mis-modeling a beam property, having the wrong sign on an applied load, or performing a mathematical calculation incorrectly. The second category includes items which the reviewer noted as a modeling or engfreeering judgasnt, but felt that a supplemental calculation was necessary to verify the conclusion, and su>sequently performed the calculation and verified the judgment. The third category includes reference to non-Project documents and a clear engir.eering judgment made but not explicity stated as such.
The conclusions drawn from this categorization are as follows:
Category Percent of Discrepancies Modeling, Input, or Calculation Error 745 Modeling or Engineering Judgment 75 Documentation Discrepancy 195
(.
The design process for small bore piping presents a large number of opportunities for the support designer to err in both the analysis and documentation of that analysis. On the other hand, the design process provides sufficient conservatism to assure that such deficiencies do not result in supports that do not meet licensing criteria. An extensive review program of tie documentation for the design of pipe supports was conducted.
The results of this program demonstrate that, while the level of documentation.
of these calculstions should have been better, the small bore piping supports are adequate and met design requirements when the documentation discrepancies were corrected.
(- .
RC Question: The ac raised questions about the placement of new restraints acJacent to old restraints as a means of qualifying the old restraints
(': (Allegation 88, SSER 21).
Response
. .New pipe supports were added to small bore piping for many reasons; e.g., to
- c. aset code break, valve acceleration, or thermal criteria. In some cases these new supports were located near existing supports. This approach would The obviously have the effect of reducing loads on the existing supports.
small bore piping program was explicitly conducted to ensure that all supports met the licensing criteria. In some cases, conditions were modeled where a structural restraint that was not a pipe support was present. For example, there are several instances in which a penetration was modeled as a seismic restraint. When a support was modeled in the final analysis, either a support or restraint physically existed in the plant orIfaanew newsupport supportpoint was a is added, modeled in the stress enalysis calculation.
documentation number is assigned to the new pipe support and remains with it throughout the design, construction, as-building, and final engineering approval cycle. This documentation trail ensures that the support is constructed in accordance with the design.
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SC (vestion: The RC raised questions about snubbers located adjacent to
- r. rig 10 restraints being inoperative during dynamic loading (Allegation 88. SSER U 21). This question was discussed further by Mr. Yin at the public meeting held on January 31, 1984.
Response
. During a site visit, the RC identified 16 snubbers that were located in close
' proximity to rigid restraints (proximity restraints). There was concern that in the event of a seismic disturbance, the rig *id restraint would prohibit the snubber from actuating. The " lost motion" or dead band", resulting from mechnical clearances in the snubber, must be overcome before the snubber will begin to restrain the piping. These clearances are typically very small and a review of the tut results for the Diablo snubuers indicates an average dead band of 0.021 inches (roughly the thickness of 5 sheets of paper).
We agree that there are snubbers located in close proximity to rigid restraints at Diablo Canyon just as there are at other nuclear plants. It has been industry practice to ignore the dead band when performing seismic analysis. This was believed to be justified since tne non-linearities induced by the small dead band described above are not sufficient to affect the results of the seismic analysis. Further, seismic stress is induced in a piping system only when large movements of the piping occur relative to the building structure. If the piping is allowed to move 0.021 inches, the induced stresses will be of an insignificant nature. It is recognized that loads on pipe supports may change.
Therefore, in order to address the potential changes in pipint stresses and
{ support loads and to provide assurance to the NRC that there <s no safety concern, the DCP has undertaken a 1005 review of all praximity restraints.
This program is described in detail in Attachment 1. Attachment 2 describes the results of this program.
The results of this study demonstrate that in no case is a section of piping overstressed or a support overloaded when the piping movement is not sufficient to lock a snubber or engage a rigid restraint.
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ATTAC K NT 1 b (Proximity Restraints)
An issue concerning the significance of snubbers located in close proximity to other seismic restraints has been raised. In its initial form, the issue was that snubbers located close to rigid restraints may not lock up during a
,. seismic event. The safety significance of this, if any, was unknown and it was felt that it should be reviewed. The review involved removing the 4
identified snubber from the piping seismic analysis if actuation was not predicted, and reanalysis of the three seismic load cases: DE, DDE, and Hosgri.
Each of the 16 snubbers identified by the Staff were reviewed. A reanalysis of the DE, DDE and Hosgri seismic lead cases was performed to determine the l amount of movement. If actuation was not predicted for the identified ,
the snubber was removed from the piping seismic au lysis. If the :
snubber,lly seismica induced piping movement was found to be greater than the amount required for the snubber to actuate, the snubber was considered acceptable since it would function. If the movement was less than the actuation level, the snubber was assumed not to function, and additional evatustions of pipe stress, valve acceleration levels, and loading on pipe supports were performed. The results of those evaluations are presented in Attachment 2.
In this review, the actuation level, or " lock-up" movement, was taken as the average value from th2 test results of snubbers in use at Diablo Canyon. The
. actual test results for the mechanical snubbers were used to extract the " lost notion" or " dead band" movement that occurs prior to snubber actuation. This
(~ lost motion includes the effects from the minute clearances in the snubber itself as well as the ball bushing and hinge pin. These movements are typical of any snubber and are not unique to Diablo Canyon. Every plant that uses snubbers has a lost motion movement of this magnitude.
Attachment 2 shows that, independent of whether the snubber will actuate, the piping system meets all licensing criteria. This confirms the validity of the design engineer's technical judgment that specific analytical treatsent of I snubbers was not warranted.
Therefore, our subsequent review demonstrates that the systems are fully acceptable, with snuober actuation specifically included. To better appreciate why snubber actuation was not initially included in the calculations, several facts should be recognized. In actual installation, there are clearances (gaps) in the r10id restraint that are designed to allow thermal, expansion or construction tolerar.ces. These clearances allow the piping'to move sufficient distance to actuate en adjacent snubber, even though the analysis may not predict actuation. More isportantly, if a snubber cannot actvate because of a nearby rigid support, the movements of the system are so small (less than 0.021 inches) that the actual piping stress cannot be significant; i.e., the failure of the snubber to actuate util not affect the piping integrity.
(. 17
In order to provide even further assurance that there is no safety concern
(; with snubbers next to rigids and anchors, and rigids next to anchors, a U thorough review was ande of the locations of all seismic restraints in the plant. A screening criteria was developed to asses. the proximity of:
- 1. snubbers next to rigids (SR)
' . .2. snubbers next to anchors (SA) ;
. 3. rigids next to anchors (RA) '
These screeening critierta considered the piping stress that would be ,
developed as a result of the snubber " dead band". This dead band would allow l movement of the pipe prior to the snubber / rigid load acceptance. An initial screening was made using a 3-diameters (3D) spacing criterion.
In order to assess the sensitivity of the 3D criterion an additional review was undertaken of all of the snubbers within 5-diameters (SD) of a rigid or anchor. Note that the 5D criterion had been previously accepted as a method for screening snubbers next to anchors on SNLPPS. The NRC both raised this question and accepted the SD response. A summary of the results is as follows:
Proximity Restraint Type 3D SD SR 25 37 SA 2 6 RA 25 37
.(. As can be seen from the above table, the number of snubber interactions is small and demonstrates that good engineering practice was employed at DCPP.
l These proximity restraints ware reviewed using the same methodology described i previously for the initial 16 snubbers.
The results of this comprehensive study of all proximity restraints i
demonstrate that in no case is a section of piping overstre sesd when the piping movement is not sufficient to lock a snubber or engag(e a rigid restraint. With over one-half of the support evaluations completed, all design criteria have been met.
The snubber and rigid interface issue raised by the Staff is a concern of recent vintage and, while it is worthy of attention from an ALARA point of view, it is not a safety concern. This issue was not part of the )CP criteria, procedures, or instructions, nor has it been an industry practice to consider, t te gaps in riflid restraints or the " dead band" in snubbers. As a cor. sequence, the IDVP d'd not review this issue. As we have stated in several NRC meetings, PGandE will undertake a snubber optimization program.
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O 'e' A ATTACHIENT 2 !
\
DDE DISP. H0S DISP. 1:ItJ88ER MAIIEER ANALYSIS DE DISP. Col 0ENTS w/o SNU8. w/o SNUB. ACTUATION
- NO. NO./ REY. w/o SNUB.
0.090" 0.180" 0.376" Yes .,
16-473L 2-105/2 .
0.063" 0.126" 0.298" Yes 16-495L 2-105/2
--- --- --- --- This snubber was identified as a potential 16-285L 4-102/4 interference 1 robles, not as a snubber l actuation prosles.
9.042" Hosgri Pipe Stresses R 16-295L 4-102/4 0.007" 0.014" Support Loads OK Valve Accelerations MA 0.042" 0.169" Yes 16-63SL 4-102/4 0.021" L 0.253" Yes
? 16-775L 4-102/4 0.081" 0.162" 0.013" No Pipe Stresses K 4-2SL 4-135/2 0.001" 0.002" Support Loads OK Valve Accelerations MA 0.056" 0.112" 0.131" Yes 4-32SL 8-109/2 0.066" 0.132" 0.159" Yes 4-33SL 8-109/2 0.030" 0.108" DDE, Pipe Stresses OK 15-63SL 8-110/4 0.015" Hosgri Support loads OK
, Valve Accelerations OK
' O e ,
i .
ATTACHENT 2 HOS DISP. SNUBER HMEER AMLYSIS M DISP. DOE DISP.
ACTUATION
- CtNSERS (
w/o SNUS. w/o SNUS. w/o SNUS.
N0. NO./REV.
0.007" No Pipe Stresses K 15-64SL 8-110/4 0.002" 0.004" Support Loads OK Valve Accelerations OK i
0.011" No Pipe Stresses OK '
16-795L 8-116/2 0.004" 0.008" Support Loads OK ,
Valve Accelerations OK 0.099" Hosgri Pipe Stresses OK 16-675L 8-117/4 0.001" 0.002" Support Loads OK I Valve Accelerations OK 0.010" No Pipe Stresses OK 16-685L 8-118/2 0.001" 0.002" Support Loads OK Valve Accelerations OK 0.264" 0.210" Yes22-400SL 3-313 0.132" 4 Yes 0.050" 0.100" 0.054"22-401SL 3-313 '
Summary 8 of 15 9 cf 15 11 of 15 Lock Up Lock Up Lock Up
- Test results from vendors indicate en average lock up displacement of 0.021".
NRC Cuestion: The NRC has raised questions about possible improper resolution ;
of p'pe interferences (Allegation 89. SSER 21).
{;.
Response
During the course of modifying piping supports interferences and obstructions were encountered. These were identified to Engineering and dispositions
- . foguested. As an example of this process, it was noted in one case that a Unistrut beau for the support of electrical conduit was constructed near a l pipe and subsequently identified to Engineering for disposition (Allegation 89 ,
from SSER 21). In fact a walkdown program designed to identify all such ;
unintentional restraints is commonly perfonned at the end of a project. Such a walkdown was performed at Diablo Canyon and any unintentional restraints were resolved by Engineering.
In a case such as the one involving the above-mentioned Unistrut. Engineering went through the following process of qualification. First, an attempt was made to requalify the system with the added restraint of the Unistrut present. In this case it was not possible to protect the Unistrut so the addition of a support at the lovtfon of the Unistrut was investigated. This investigation showed that ?.he Unistrut was not required and it was removed from the plant. All of this was part of the iterative practice of qualifying an installed piping system and is not unique to this plant. All applicable procedures were followed in this process. Since all design criteria were met. !
there is no safety significance to this item. In fact it would appear that this situation demonstrates good communication between Construction and Engineering, sound engineering practice, and a proper solution that resulted
(. in a system that meets the design criteria.
4 I
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4 IRC Ouestion: The NRC has raised a question about the calculation of the ioas-carryin (Allegation 79 and
\,c 88, SSER 21)g capacity of the small bore piping supports Response ,
Different methods exist to qualify a piping system to design criteria. These An example of
_ methods often require iteration between engineering designers.
this can be seen in small bore piping qualification, where the pipe stress analysis produces reactions or loads on the pipe supports. After obtaining the loads on the supports, the pipe stress analyst transmits results to the
- pipe support engineer for his use in qualification or design of the supports for these loads. The pipe support engineer reviews existing as-built pipe support drawings. If the support is determined to be inadequate to sustain this initial load, the support designer and the stress analyst may well review the systan to determine if the engineering assumptions in the piping stress ,
analysis have excessive conservatism. An additional series of more realistic calculations may be performed before it can be shown that the support meets criteria. This process of recalculation may occur many times before the support is qualified. Such an approach is a logical and orderly method of qua'lifying small bore piping systems.
Another method used to qualify a piping system involves use of the maximum capacity of the pipe suppcrts for qualification. This method can be more efficient than the method discussed above by reducing the number of iterations and recomputations between the stress analyst and the pipe support engineer.
In this situation, the pipe support engineer calculates the maximum capacity of a support for each load case. This information is provided to a pipe
(- stress analyst, who compares the computer results of the piping stress If the calculated support loads analysis to these maximum allowable loads.are in excess of the allow iteration without requiring the pipe support engineer to recalculate stress in the support. This so-called technique of a " reverse calculation" is used to reduce the number of calculations and interfaces between the engineers.
However, it does not alter the final result since both the piping and the supports must be shown to be qualified to the applicable licensing criteria.
When the piping analysis is complete, all loads are transmitted to the support The engineer for final acceptanc'e or support modification and documentation.
reverse calculation technique is often used in the industry and is analogous ,
to calculating an acceptable load rating of standard supports.
This question also conveyed the implication thet intermediate Such an taplication or iterative is calculations erroneous.
were being improperly destroyed. Procedure 3.3 contai Pursuant the preservation of the final stress analysis calculation packages.
to p_rocedure 3.3 3 all final calculation packages are retained and permanen f11Vd. There is no regulatory or other Project requirement to retain the intermediate or iterative analyses.
- 22-I
, . _ . - ~ _ _ _ _ _ . _ _ _i __ a__________.
s l s IstC Question: The NRC has raiseda ' ques' tion about assumption of joint releases for rigid connection'(bliegation 88, $$ER 21).
t - '
, Response
, N 1
' Joint releasesorefers to a method oftproviding an accurate representation of
' end concoctions in structural members. An initial calculation of a pipe
' support frame might conservatively assume that welded ends at structural members are completely rigid. However, ft41s obvious that no joint is
[
- completely 1001 rigid. The structural member may have very little moment
$ resistance in some rotation axes, and assuminJ rigidity is not representative of actual behavior. An engineer may model'the jolnt to closely represent its actual physical characteristics. In many instances, the joint is modeled so that no moment resistance is offered by the steel to which the member is attached (i.e., assume that moment loads are not transmitted). This method provides a more realistic model of tie structural behavior of the frame.
The weld at the joint is still considered in the computer model, and there is no intent or need to remove it since the forces transmitted by the weld and associated stresses are evaluated and verified to be acceptable. This practice is standard in structural engineering evaluation of frame structures.
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IUtC Question: The NRC raised questions about U-bolt allowables (Allegation m3, na.x z U.
Response
During the January 31 meeting the NRC indicated that it was currently reviewing the information that had been submitted on December 28, 1983
. , - concerning U-bolt interactions. One area of review that remained was the test <
' sample size. The following information provides the justification for establishing U-bolt allowables by compliance with ASE testing requirements ASE Section III, Subsection IF-3260, provides the procedure by which U-bolt allowable ratings were developed. per NF-3260, the procedure for load ratings consists of imporing a total load on one or more duplicate full-size samples of a component support. The total load is to be equal to or less than the load under which the component support fails to perform its required function. If a single test sample is performed. NF-3260 requires the load
- ratings to be derated by 105.
The tests performed for the Diablo Canyon supports were more numerous than the <
single test permitted by the code but were less than the " statistically significant sample" allowed by the code as an alternate. The conservatisms added in the generation of allowables is at least equivaient to a derating of allowables by 105. The following is P. summary of conservatisms:
< (1) A minimum of four U-bolts were tested fcr three loading conditions for each pipe size. The loading conditions consisted of the application of .
tension loading and a combination of side and tension loads
(- side loading,llowables (45o). The e for tension and side loading were based on the lowest test load of all pipe sizes tested using a given diameter U-bolt.
The test loads used in the equations of NF-3260 repr2sent the lowest i
j tension and side test loads found for 1/4-in. and 3/8-in. diameter rod U-bolts, respectively, i
i (2) The added conservatism occurs in the interaction formula with the ,
application of both tension and side loading because the minimum tension test results and the minimum side loading test results are combined.
(3) U-bolt tension failure did not occur for any U-bolts for piping sizes greater than 1-1/4 inches in diameter. The allowables were based on the testing machine's capacity rather than the U-bolt's capacity. Therefore, substantial margin exists for the larger U-bolts.
In susunary, the load ratings for U-bolts meet the requirements of the ASME Code for qualification by type testina. The use of allowable U-bolt ratings determined by qualification testing will reliably ensure a conservative design and is consistent with all design criteria.
(
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b d i lutC Question: The NRC has raised a question about angle-shaped structural
- members (Miegation No. 95 from SSER 21).
Response
[
In this response, the following symbols are used.
~
S List of Symbols ,
- B= Length of angle leg
- t= Thickness of ar.gle 1c; e
- L= Length of span h
Fy= Minimum Yield Strength
[ ,
b= y Width of Compression Flange
?
In small bore pipe support design, angle-sectioned beams are frequently used i
for structural members because of the small loads typically encountercd in
[ small bore piping.,
(, Angle sections were vred at Diablo Canyo., prior to the verification program. )
~
Where modifications to existing supports were made during the verification l
Program, structural tubing was often substituted for the original angle L section. , l t
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~
The criteria for the use of angles as laterally unsupported beams subjected to
- bending forces were based upon evah.ations initiated in 1977.
Troject-specific critaria were required bacause the AISC Manual of Steel k Construction (Ref. 1) does not provide guidance for angles with laterally t
unsupporte:i spans greater than 76.0 bf/Fy. The ters 76.0 bf /Fy is the allowable-span for an unbraced length of a member not meeting the requirements t , of Section 1.5.1.4.6a of Reference 1. However, these criteria were developed for I beams and r.ot specifically for angles. Reference 1 does not provide
?y '.
criteria for laterally unbraced members greater than 75.0 bf /Fy. The lack p of specific guidance in this area has been recognized in the literature (see Referen(e 2). However. AISC recognizes that special investigations are necessary for angles with latertl y unsupported spans greater than 76.0 bf /
5 Fy. This is indicated on page 2-21 ef Reference 1 where a statement is provided which explains the use of angle 1 cad tables. The statement is as follows:
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"The tables are not applicable for angles laterally unsupported
(;
~ or subjected to torsion; for such members a special N investigation is necessary."
Secause the AISC did not completely address the design of laterally unsupported angles, PGandE performed a literature search in 1977 to determine -:
if other information was available which would be adequate to set criteria.
In late 1977 it was found that extensive testing of laterally unsupported angles loaded in bending had been performed in Australia. Literature which describes the testing, findings, and recommendations has been previously ..
provided to the NRC staff (References 3, 4, and 5).
In the Australian tests, various sizes of angles were characterized by different B/t ratios. Angle sections with B/t ratios between 6 and 16 (Reference 5) have been tested. 1he majority of angles at Diablo Canyon fall within this range. The only angles at Diablo Canyon not falling into this 1, range have B/t values less than 6. However, at this end of the range (beams with B/t less than 6 are less slender) the data can be used conservatively since the net effect is to allow an increase in acceptable unbraced lengths.
Based on the tests and comparison to structural theory, simple formulas were developed in Reference 5 for use in the design of 1cterally unsupported angles .__
in bending using several different methods of load application.
For all the various angle sections and load cases investigated, Reference 4 recommends that an allowable bending stress of 0.66 Fy may be used if L/t is less than 300. The Diablo Canyon Project Design Criteria M-9 limits the maximum bending stress to 0.6 Fy and a maximum L/t ratio of 270. These limits used at Diablo Canyon fall within the recommendation of Reference 4 and are
( thcrefore acceptable.
Refarences 1-
- 1. American Institute of Steel Construction (AISC) Manual of Steel 4 Construction. Seventh Edition. AISC, New York.
B. F. Thomas, J. M. Leigh, M. G. Lay, Civil Engineering Transactions. L Z.
19/3, The Institution of Engineers. Australia. _
- 3. B. F. Thomas and J. M. Leigh. The Behaviour of Laterally Unsupported Angles BHP Melbe Res. Lab. Rep. E L 22/4, December 1970. Y
- 4. J. M. Leigh and M. G. Lay, Laterally Unsupported Angles with Equal and ;
Unequal Legs. BHP Melb. Res. Lab. Rep./ EL 22/2, July 1970.
- 5. Safe Load Tables for Laterally Unsupported Angles, Australian Institute of Steel Construction, September, 1971.
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IstC Question: The HRC has raised questions abcat the calculation of runsamentai fraqtiencies for small bore piping.
{ Response The Rayleigh method for the determination of natural frequency was not used in the analysis of piping supports for small bore piping. A static equivalent approach was employed, whereby a unit force (1.0g times the tributary mass of the piping) was applied in the restraining direction of the pipe support. The corresponding deflection of the pipe support was then compared to an s'iowable limit. A deflection of less than 0.025 inches indicates a suoport that has a natural frequency of over to Hz. Simple beam theory was used to convert the desired frequency to a deflection criteria. The Hosgri report (Section 8.2, page 8-8) indicates that the support was to be assimed rigid in the seismic analyses if its natural frequency is above 20 Hz.
During the January 31, 1984 seeting with the NRC, a question was asked to clarify the loading direction in calculation W-988 for the applied unit load. A resiew of calculation W-988 indicates that the 1.0g load was, in fact, applied in the restraining direction of this particular pipe support as the horizontal plane is the restraining direction for this pipe support.
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4 sc Question: The RC has raised a question about the size of the sample vu nzea for reverification of small bore piping.
Response
The program to verify the small bore piping at DCP began in 1981 by the selection of a sample of typical piping and supports. This sample was rigorously analyzed for compliance with all appitcable licensing commitments and criteria. The results of the initial sample analysis indicated that there were several areas where incorrect or incomplete assumptions had been used in the ori inal analysis. Additionally, areas were identified where the original
- criteri had not been totally followed. These errors were generic to all small bore pip ng analysis and were, therefore, addressed by reanalysis for
[ all portions o piping where these generic errors could result in
- noncompliance with design criteria. Examples of these generic issues were r allowable active valve acceleration, considere. tion of anchor movements,
' thermal analysis of piping, and code breaks.
! The identification of these generic issues caused the original sample program r to be revised and expanded. These generic issuas would be evaluated for all l piping and a sample approach woold be used in the qualification of the 4 remainder of the small bore piping. In accordance with that philosophy, a sample size was selected by the ITP and subsequently approved by the IDVP and
- the NRC. This concept used a worst case scenario for selecting the sample piping that would be reanalyzed. For example, systems were selected in areas ..
of the plant where the response spectra were the highest. The initial sample selected in the fall of 1982 rea:.ined the " sample" throughout the small bore a
verification program. In its original form, the 5000 feet of sample piping l L
( was intended to valify 25,000 feet of a total of 43,000 feet of piping in the .
plant. The resa ning 18,000 feet required reanalysis because of the generic s issues.
w.
The tennelysis required for the generic issues proceeded by identifying all -
E piping and supports in the plant that could be affected by these generic i
issues. All small bore pipe was reanalyzed and modified if necessary for these issues, including the sample piping. As this effort proceeded, it became obvious that additional generic issues had been identified and should be included. For example, o'ne original generic issue was qualification of hot p1 Ping. Further analysis indicated that the intermediate-range temperature piping required reanalysis and should also be included as a Seneric issue.
1 Therefore, as the program evolved, the amount of piping that was analyzed as L part of the generic program grew and the amount qualified by the sample ,
B Program shrank. When all of the issues had been evaluated and the final program' completed, 28,000 feet of small bore piping were qualified by rigorous 7-
[
reanalysis and 15,000 feet were qualified by the 5,000-foot sample. It must also be reseabered that all the generic issues were also addressed even in the ..
15,000 feet qualified by the sample program.
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The sample o as was only used to qualify low temperature piping systems (c (lesstian for carbon steel and 1600F for stainless steel) without
- remote operated valves, code boundary changes, or significant anchor movements.
During the IDVP review of the ITP, the small bore program was exhaustively examined. The IDVP reviewed in detail completed samp,es of span rule application of File 44. Because the IDVP selected a portion of the san >1e
.,1 s grogram to review, they explicitly reviewed the File 44 methodology. Tsis was beca6se substantial amounts of File 44 analyses were included in the sample program. Of the 5.000 foot of sample piping. 3.400 feet had been qualified by File 44, which was the original analysis method used prior to 1981. The only hardware modifications installed on piping originally analyzed by File 44 were to address generic concerns.
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RC Ouestion: The RC raised a question that ITR No. 60 toentified one case
( wnere sne proper criteria had not been used for the review of natural frequency.
Response
. , E01 113g identified one small bore support that had incorrectly compared the talculated value of pipe support deflection (used for natural frequency estermination) to an allowa>1e of 0.0625 inches. The proper allowable was 0.025 inches. This calculation was recalculated using a more complete model and a computer solution. The results indicate i that the frequency was above the 20 Hz criteria.
It should be noted that even if the value of the natural frequency was below ,
20 Hz, an insignificant change in system response would result. The reason for maintaining the natural frequency of a pipe support above 20 Hz is to An permit consideration of a rigid restraint in the piping strass analysis.
equally acceptable analysis technique is to calculate the frequencies or stiffnesses of the supports and to analyze the piping with these stiffnesses included. Since there are many pipe supports on one system (analysis), the reduction of the natural frequency on one support to below 20 Hz would result in an insignificant effect on the piping system respotise and support loads.
To ensure that this was an isolated, random error rather than one that was generic or indicative of a programmatic breakdown in training or design control other calculations of natural frequency performed by the same originaling engineer have been checked to assure that he had not systemically
(. used the improper allowable. In all these cases the correct allowable was used. Additionally, the review being performed in conjunction with the concern for calculational errors has not uncovered any other instances where this incorrect comparison has been made. We therefore conclude that this was an isolated sistake that was not representative of a generic concern or a programmatic breakdown.
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III. N0NTEC W ICAL ISSUES
{J mc Ovestion: The NRC has raised questions about " destroyed de::umentation" TA11egation 87. SSER 21) and " altered current documentation" (Allegations 55, 87, and 79, S M R 21). These concerns were discussed further by Dr. Hartzman ,
l at the public meeting held January 31, 1984.
'. Response The verification process for small bore piping analysis is an iterative one.
l The initial analytical attempt is usually a conservative, simplified bounding l
calculation whic1, if successful, expedites the verification process. If, however, this bounding calculation does not demonstrate adequacy of design, a more sophisticated analysis is then initiated. This process is repeated until either the adequacy of design is shown or a determination is made that modifications are necessary. ANSI standard N45.2.9 (1979) does not require The only calculations required to be retention of intermediate calculations.
retained are the final calculations which reflect the analysis actually relied upon to show adequacy of design. For the situation considered, no superseded calculations are required to be retained by regulation, regulatory guide, standard, or procedure. Despite this fact, DCP procedures, based on judgment of the analyst and checker, call for retention of superseded calculational records "to the extent necessary to support and verify final designs."
' The specific calculations involved in Allegation 87 are W-988 and W-944.
These Unit I calculations were originated and checked by individuals in OPEG ,
1 who had working responsibility for small bore piping analysis. After j
(- origination and checking, but prior to approval of the calculations in The ;
question, the OPEG group was divided into Unit 1 and Unit 2 sub-groups. ;
analysts who had derived these calculations were reassigned to the Unit 2 D group. The two calculation packales were reassigned to individuals of the Unit 1 group who elected to re-perform the unapproved calculations for W-988 ,,
and W -944. The new calculations were checked and approved in accordance with applicable procedures; thus, the earlier unapproved calculations were not retained in the calculational packages.
Several factors have led to confusion and misunderstanding of the calculations in question. First,The the initial calculation of W-988 showed the support not second attempt, by a different analyst, showed the to be qualified.
support error. to be qual 1#ied but unfortunately that calculation c the second analysis. Obviously one could speculate that the second analysis was somehow dishonestly done (as opposed to an ' honest mistake") to "make the probles go away." Such was not the case. A third analysis was completed which shows that indeed the support is qualified as designed and constructed.
W-944 was a calculation that had not been approved at the time of the personnel transfers and the checkar of the original calculation became the analyst of the next iteration. Obviously that individual was aware of the
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status of the original analysis and qualified the support in the normal . .
iterative process. .j
{ Adding to the misunderstanding is the issue of a master log and an unofficial informal log which, on the surface, appear to contain conflicting ..
information. Each calculation package contains a calculation index and, in ,
. addition, there is a master log which lists the design calculation number,
- ,;. revision number, hanger number, calculation status, analyst's name and date,
- checker name and date, and approval date.
Confusion has arisen because of the existence of the vr. official informal log
- L that was kept as an aid to the Assistant Onsite Project Engineer in tracking
= enflineering activities. The informal log showed the two calculations and the L
" or'ginal assigned analysts. Other than indicating the coupletion or approval
! date, the informal log was never updated to reflect the reassignment of the :
calculations to the new analysts and checkers. The informal log was not. -
=
- however, the record calculation index or master log, but rather a management ..
tool which was not required to indicato the information contained in the
[ master log.
r :
' Both the calculation index and the saster log properly documented the approved r calculations for MP-988 and MP-944. In accordance with applicable procedures.
calculations are not indexed in the calculation index or logged on the master i
log until they are approved. Because the original calculations had never been I approved they were neither indexed in the calculation index not logged on the master log. Thus, in neither case were official calculations, nor 4 i
calculations "necessary to support and verify final design," destroyed. jh h .
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( Based on comments made by the NRC Staff at the January 31,1g84 public .
- meeting, Allegation 55 seens to be based on two calculations MP-072 and /1 E
W-345. Calculation W-072. Rev. O, analyzed hanger 2171-16 and showed that a
- U-bolt would be overstressed. The originator suggest2d the use of a cut plate ;
r_.
bracket instead of a U-bolt. The recommended design modification was checked
[ and approved according to written procedures. Prior to issuance for i
constr9ction, the stress analysis was redone and new loads were issued. An F analyst was given the hanger to review. Our investigation has not positively determined who wrote the phrase "too costly to fabricate" on the original -
- desian but it is believed it was the analyst who also did Rev.1 of the y
calculations which also indics 'd overstress of the U-bolt. Thereaf ter, Rev.
2 of the calculations was performed analyzing the support showing an angle iron in lieu of the U-bolt. This analysis was also performed by the originni i
= analyst. The calculation was checked, approved, and issued for construction.
During construction the r.upport was further modified and an as-built was issued 4yConstruction. That as-built condition was approved pursuant to
[ applicable procedures. Our review indicates that all design and construction activities concerning W-072 met all procedural requirements and criteria.
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2182-74. The originator of l The second calculation. W -345, analyzed han pr i this calculation proposed a design modificat'on to the support because 049243. theThe Rev. 11. i h~ axial therinal movement exceeded that allowed by drawingoad. The group leader support was otherwise capable of accepting the piping approved the calculation as ' preliminary" without modification, but noted at i the end of the calculation that a modification was not required due toThis an note /
insig *?fcant uploading in the support (less than 45 of allowable). '
!*.' was signed and dated. At the time of his decision, the group leader was aware of a pending revision to drawing 049243. shich would support hit decision. !
Thus, the calculation indicated the design adequacy of the hanger in accordance with the to be-approved revision of drawing 049243. This calculation was subsequently reviewed to verify its compliance with the revised drawing and was then approved. Again, we are unable to discern any In each of ;
- the altered current documentation" in our review of this calculation.
above instances there was some initial iteration of design approaches after which the final design was derived, reviewed, and approved in accordanc with applicable procedures.
In conclusion, our analysis of Allegations 55 and 87 does not indicate any dtstruction of documentation that was required to be retained nor does it show any instances of alteration of documentation in the pejorative sense,
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15tc Question: The NRC has raised questions about the ettent and timeliness of training of onsite pipe support engineers (Allegation 82. SSER 21). Concern C' was also raised that the responsibility and authority of small bore piping group personnel did not appear to have been delineated in writing. .
Response
- >. The Project provides formal training in the Engineering Manual Procedures
("EN") which implements Project QA requirements. Those requirements meet QA Criterion II of 10 CFR 50 Appendix B, and are set forth 1.1 Nuclear Quality Assurance Manual. ("NQAM") and Bechtel Quality Topical Report, Rev. 3A .
("SQ-TOP-1") which has been approved by the NRC for the Project. Each engineer assigned nuclear safety related work receives indoctrination and training in EMP in accordance with Procedure 2.1 of that manual. This course for the engineers identifies and describes the procedures applicable to their I, work. It includes a review of procedures on design criteria memoranda, design t i
calculations, design changes, drawing control, discrepancy reports and I nonconformance reports. ;
PEI-15 specifies that the indoctrination and training are to be given within 30 days of assignment to the Project. Training records indicate that '-
l approximately 705 of all DPEG design engineers on the current OPEG roster j received Engineering Manual training within 30 days of assignment as required. Approximately 955 received such training within four months of assignment. The majority of those instances where an engineer did not receive training within 30 days of assignment occurred early in the Project. Project Audit 28.4 conducted in February 1983 and closed in May 1983, resulted in the correction of most of these discrepancies. Since May 1983, only five OPEG
! ( design engineers have exceeded the 30-day traininsi requirement by more than a few weeks. As 1005 compliance is required, administrative changes are being made to assure that all engineers receive required indoctrination and training within the prescribed times.
~
The training program covered by EMP 2.1 is consistent with QA Criteron II and is directed at the process of design control, design change, design calculation, discrepancy and nonconformance procedures. EMP 2.1 is not addressed to the professional qualificatian of engineers and designers, and tiierefore does not encompass'the technical education necessary to enable an engineer to properly perform design work. To ensure technical competence, pipe support engineers are hired in large part on the basis of interviews, educational qualifications, and previous experience. For permanent or temporary employees, the professional credentials of all are required to be ..
verified by either the Personnel Departments of Bechtel or PGandE. For contract employees, such verification is a contractual requirement for the j contract firm. This process is detailed in Table I. A thorough review of the engineer's work experience is confirmed through technical interviews conducted by senior Engineering gersonnel. Additionally, the engineer's first assignments are carefuily selected to provide an adequate opportunity for the designer to gain familiarity with project calculation format and methods, and I
( 34 :
his work is closely monitored to assess the designer's capabilities. Future
(, assignments are determined on the basis of assessing the engineer's L performance on these early assignments.
A review of the technical background of the engineers in the small bore pipe support group at the site shows that experienced, technically qualified
. engineers had been hired, with little or no need for additional instruction in
.. small bore piping calculations other than that normally provided to familiarize tum with the proper design critsria and Project calculational methodology. Of all the pipe support engineers employed at OPEG, more than 415 (36) and greater than five years of nuclear related experience. Most of the engineers had worked on two or more other nuclear power projects, with many having worked on five or more plants. All have at least a BS in Engineering or equivalent, and their minimum professional experience is one year, the maximum professional experience is 14.5 years, and the average l professional experience is greater than five years.
In SSER 21 (Allegation 82), the Staff identified five individual engineers who had not received procedural training within 30 days of commencement of their assignment as required by PEI-15. The project has reviewed the work of those individuals along with all of the pipe support engineers. The apparent discrepancies in calculations that are currently being reviewed are t eing correlated with indoctrination and training coupletion dates for persons originating and checking the questioned ca culations. For each such discrepancy checked to date (the 23 Stokes calculations), all individuals
! completed the QA orientatfon program prior to approval of the final calculation under review.
(' While some individuals did not receive indoctrination and procedure training within the 30 day specified period, the records indicate that the discrepancies in calculations that have been observed are not related to either indoctrination and training or professional experience, but rather are random events. Consequently, the delayed completion for the training of a few J
design support engineers does not appear to relate to the discrepancies j detected.
In order to better implement Project training requirements, the Project proposes the following new actions for OPEG:
- 1. i Training records of all ent r.eering personnel working on the Project have bee: reviewed. Effective < smediately, any person who currently does not have the required training in QA and engineering procedures will not be allowed to continue engineering design work until such training is completed.
A weekly training sessions in QA/ Engineering procedures will begin immediately to train new arrivals. Also, a refresher course will be held hree times a year for all engineering personnel who complete or who have apleted QA/ Engineering procedures.
(.- ---
- 3. No person newly assigned to OPEG will be permitted to perform, check, or approve any calculation until the QA/ Engineering procedure training has
{ been coupleted.
- 4. Failure to complete a refresher course within 30 days of requirement will disqualify an engineer from performing, checking, or approving any calculation. ,
- 5. All training personnel will utilize a formal syllabus which shall be reviewed and approved by engineering and QA management. Initially, the training sessions shall be monitored by engineering and QA management to assure that required matters are properly addressed. Training sessions will give special attention to changes in procedures that have been implemented in the last year.
- 6. All such training requirements will be formalized End documented, and compliance will be verified by QA audits. ..
Concern also has been raised that the responsibility and authority of small bore piping group personnel did not appear to have been defineateo in writing. The small bore piping design group personnel authority and duties are delineated in writing through the DCP QA Program, procedures applicable to the engineering work, and organization charts.
OPEG is an extension of the home office project engineering organization which is located in a different geographical area. This relationship is defined in the DCP Nuclear Quality Assurance Manual (NQAM) Section 1 No. 7. As part of
(. the project engineering team, OPEG carries out the Engineering Department's <
responsibilities outlined in NQAM Section 1 No. 7, as directed by the Project Engineer to whom OPEG reports (Reference NQAM Section I No.1. Figure 7).
The specific duties, responsibilities, and authority of OPEG at the Diablo Canyon jobsite are delineated in procedure PEI No. 9. Rev. O. The accomplishment of these duties and responsibilities is delegated thrcugh the organizational chain from the Onsite Project Engineer / Assistant Onsite Projact Engineer to lead discipline engineers, then to the discipline group engineers. Assigment of these duties and responsibilities is made by the OPE /AOPE and lead discipline engineers. The organizational chain within OPEG is defined both in PEI No. 9 and in a written organization chart maintained by the Onsite Project Engineer.
The authority and duties of personnel shown on tte established organization chart are delineated in writing as followu a'.' ' Onsite Project Engineer /0nsite Assistant Project Engineer responsibilities and authorities are defined in PEI NO. 9 Paragraphs 3.3 and 3.4. Signature authority of the OPE /0 APE is defined in PEI No. 9, paragraph 4.3, and responsibility for
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approval of design changes initiated by OPEG is defined in PEI No. 9, C paragraph 4.2.4. Additional duties are defined in other procedures app 1' cable to design of piping and piping supports, consisting of Engineering Manual procedures; Piping Group Controlled Procedures, Instructions and Criteria; and Project Engineer's Instruction (Reference -
PEI No. 9. Paragraph 4.2.1). .-
./ . ' b. Lead Discipline Engineers are jobsite re$resen5atives from the Office Engineering Group Supervisors (EG ). The Lead Discipline Engineers receive technical direction from the Home Office EGS and administrative direction from the Onsite Project Engineer. These authorities and responsibilities are documented in PEI No. 9 Paragraph 3.5. Authority for sign-off of OPEG originated design changes is documented in PEI tio. 9. Paragraph 4.2.4.
In representing the EGS for activities within 0FEG's scope, additional duties of the EGS/ Lead Discipline Engineer are defined in other procedures applicable to design of piping and piping supports as listed in item (a) above. For example, Engineering Manual Procedure 3.3 Rev. 5 and Pi ng Procedure P-6 Rev. 2 require the engineering discipline group 1 der or supervisor to approve design calculations for pipe supports. For OPEG pipe support calculations, the Lead Discipline Engineer has this duty as described above.
- c. Area Leaders and Squad Leaders are responsible to assist the Lead
- Discipline Engineer in the performance of his duties and to work
(- under his direction. This organizational responsibility is delineated in the OPEG organization chart.
- d. OPEG engineers work under the direction of the Lead Discipline Engineer as defined in the OPEG organization chart. All work performed by the OPEG discipline group engineers is coordinated and supervised by the Leed Discipline Engineer. The discipline engineers do not have any other authority and duties except to follow the direction of the Lead Discipline Engineer in accomplishing the assigned task. Their specific authorities and duties with respect to assigned tasks are delineated in the procedures that apply to their work. The procedures applicable to design of piping and pipe supports are defined in PEI No. 9 Paragraph 4.2.1. For example, an engineet assigned to check a calculation has authority to require corrections to calculations, as delineated in Engineer Manual Procedure 3.3, Paragraph 4.2.6,
'- and he has the duty to perform checking in accordance wtth Engineering Manual Procedure 3.3, Paragraph 4.2.2.
rformed by personnel The more gen 2ral authorities and duties emected to be
- assigned to 0 PEG in specific positions wit 11n the disci Ifne group are defined and delineated in accordance with established Bechtel practices. Generally,,
_g
they cover three categories of personnel: (1) permanent employees. (2)
( contract each is summarized (job shop)in Table I. personnel, and (3) temporary personnel. The process for In light of the foregoing it is evident inat the onsitt small bore piping design group authority and duties are established, and are described in writing to the extent necessary to fulfill the requirements of Criterion I to
- . 30 CFR Part 50. Appendix 8.
Attachments:
Table I Attachment A - Example Job Description h
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( .
TABLE I b A. permanent Personnel
- 1. Opening is identified and and related to Job Description (e.g.,
Attachment A), by Project.
'l 2. Chief Engineer either provides a prover indiv1' dual from elsewhere in the organization, or finds a new employee through Personnel Department.
- 3. In hiring a new employee, the Chief Engineer makes selection based e
upon personal interviews, reviews of experience and educational background, other credentials, and as much inquiries of former employers or supervisors as he can make.
- 4. After hire, the Personnel Department confirms key parts of employment and educational buckground to the extent practical.
- 5. Three (3) months after hire, the employee is given a formal performance evaluation, followed by another in nine (9) months, and thereafter one every twelve (12) months or upon change of supervisor.
B. Contract (job shop) Personnel
- 1. Same as A-1, above.
( 2. Chief Engineer requests Personnel Department to have contract agencies provide resources of candidater.
- 3. Chief Engineer reviews resumes, conducts interviews, and selects most suitable candidates (typically one out of eight candidates).
- 4. Personnel Department executes agreement with contract agency to provide selected personnel, which includes responsibility of contract agency for accuracy of background information and credentiels.
- 5. Personnel are initially indoctrinated and closely supervised. They are also periodically ranked, and those with lowest rankings are replaced.
C. Temporary Personnel
- 1. Same as A-1.
- 2. Chief Engineer identifies personnel for temporary status from among contract personnel, having made selection es above.
(- l
- ATTACHMENT - A
- JOB DESCRIPTION (Example)
( *
/
ma SENIOR ENGINEER
~
-ones.umeen 300A,190A assaovso anuev ennon evem ,w coo.
EST h= 35, 26 ENGINEERING SUPERVISOR apreenva oes r _.,
3ely 3, goes ENGINEERING amoucas senaurnou enroo Of73CE ENGINEERING SUROSARY:
Plans and conducts independent work requiring ludgment in the evolustion, eclection, application and adaptation of
@ zing techniques, procedaros and criteria. Devis:s new approaches to problems.
por antary pede determination, see attached addendum.
JOS DDEENSIONS:
A. Supervielen Resolved a performa most assignments independently with instructions as to the general resutts 1xpected. Receives technical guidance from Engineering Specialists or Supervisors on unusual or compiox problems and supervisory aporova pecposed protect pians. .
- s. Supervesien Emeressed
- Provides technical direction and assigns work to engineers, designers, drafters, technicians and others who assist in performing specific assignments, however is not responsible for staff planning or salary actions.
C. Centacts
- Independently contacts vendort representatives and project field personnel to gather or give information. Cont alient counterparts as directed.
l
( PRINOPAL RESPONSIBILITIES:
- 1. Plans, schedules, conducts, and coordinates detalled phases of onelf' sering work usually in one discipfine i or staff group. Performs work which involves conventional engineering practice but may include complex fea as resolving conflicting design requirements, unsultability of conventional materials and/or difficult coordination seguirements.
- 2. Plans, ocordinates or' prepares equipment or work specifications, bid evaluations and award recommendat equipment.
- 3. Coordinates engineering efforts in assigned areas hetween specialty and other en0lneering groups or discipl the ellent, suppliers, and contractors and between $ther divisional groups.
- 4. When delegated, assumes a lead role over oiher en0ineers or project out> groups for comple.Ing specific ta S. Assists in on the job treinine of assigned personnel and provides input for their performance evaluations.
S. Prepares lettere to vendors and clients.
l
- 7. Reviews bid analyses and makes recommendations.
S. Prepares or assists in preparation of conceptual studies, designs, reports or proposals.
S. Performs or assit ts in the performance of problem analycis and original design.
- 10. Reviews project controls, cost estimates, quantity take offt and manpower requirements for proposats change enters
- 11. Reviews and checks work of subordinate engineers.
JOBENOW12DGE A thorough knowled0e of oneinsering techniques, th9 design of engineered systems, and engineering a tions. A broad knowledge of the application of en0ineering to plant constructability as appiled to const motorials.Up'.odtate knowled0e of computer applications to oneinsering and design.1Riorking knowled plannine and control methods including computettaed methods.
(..
- A Isoad knowledge of precedents in the specialty area and a Good knowledge of principles and practices technical areas.
A knowledge of rotated construction practions and the economics involved
N * **
vera
(*;~ SENNNt ENGINEER 3,M A current knowledge of industry or regulatory standards and desig,n ortleria portinent to the particular engineering discipline.
Skill in oral and written communication.
The above is normally acquired through-e A rooognited degree in an engineering or eclentific discipline from an accredited college or university OR e A pe?::"rel liconee in an approgvlate enginetting discipline from a recognized licensing boerd.
OR e Sufficient number of speciallred courses in relevant germal engineering or appropriate engineering disciplines to meet job ttMautrements.
AND e Practical work experience in design enginesi.ng or relevant equivalant experience in allied types of engineering outficient to demonstrate competence as a trained engineer.
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e
(, SENIOR ENGINEER y ADDENDUM Salary Grade Determination for SENIOR ENGINEER Grade N Plans and coorenstes independent work requiring lud0 ment and emperience in the application and substantial adaptation of engineering techniques. Devises new approaches is technical problems.
Provides technical direction for specific tasks and assigns work to subordinate senior engineers, en0heers, des 1 0ners, drafters or project subgroups.
f Requires superience and demonstrated skill in handling professional work at the grade M level and a broad knowledge of precedents in the industry.
Gredo N Plans and conducts independent work requirin0 judgment in the application of en0ineering techniques. Normally uses conventional approache6 to technical problems encountered.
Provides technical direction and assigns work to en0 ineers, dssigners, and drafters who assist on specific assignments.
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15tC Question: The NRC raised questions about various aspects of document control for saall bore pipe support design (Allegations 55, 79 and 84, $$ER
{ 21).
Response ' ,.
w The DCP QA Program requires formal control of implementing procedures.
Detailed requirements are contained in Engineering Manual Procedure 5.2.
Implementing procedures are required to be logged into a control system by title, date of approval'and revision number. All holders of implementing
. procedures are reoutred to formally acknowledge receipt of revisions by returning a signec acknowledgement.
Spe:tal implementing procedures, instructions and criteria for the small bore iping troup,designverif'cationeffortwereauthoredbytheProjectTeamPiping and the control of their distribution was managed by the Project Administration Group using a system of signed, returned receipts.
A master document distribution matrix was prepared to establish which manual holders receive specific documents in accordance with the requirements of their job assigissent. A specific set of defined documents is assigned to a pipe support engineer; a different set of documents is assigned to a pipe stress engineer, and so forth, a) Out-of-date Procedures The staff identified three instances of out-of-date procedures contained
( '- within the controlled procedure manuals maintained in the OPEG. As a result, a discrepancy report (DR 83-47-5) was issued by Project Engineering. This DR addresses corrective action, impact on final design and actions to prevent reoccurrence.
A 1005 review of all control procedures, instructions and criteria assigned to OPEG personnel was completed by December 15, 1983. Sixty-three (63) manuals containing 133 criteria documents 412 procedures and 451 instructions were reviewed. The results showed that 905 of the documents assigned to the manuals were correctly in place. The review results have been evaluated to -
determine the possible impact on the small bore reverification work. Most of the instances found involved doceents missing from certain controlled manuals, in which case the appropriate requiresents are available to the engineer through other controlled manuals in tha work area. Each instance of an outdated procedure or instruction was evaluated and determined to not impact the completed design work. The documents found to be outdated were characteristically documents that the assigned manual holder would not be using in~ performing his spe.cific assignments.
All 63 controlled manuals have been brovflht up to date. They now contain only current copies of those documents specif ed by the master document distribution matrix.. .
l r '. The Staff also expressed the concern that since Piping Procedure Manual B-075 C. was presumably the only controlled manual assigned to the OPEG Stress Group, there was a possibility that Stress Group engineers had been witimt access to uo-to-date edures for an extended period cf time. However, our ihvastigati has shown that other controlled copies of the menual had been assigned er avellable to members of the Stress Group since the inception of the WEG gOoup. For example, the October 14, 1982 Distribution List for pgping Grop 1Preii:edures, seders of the Stress group were assigned controlled manuals.
Instructions and Criteria for Diablo piping Design shows that Although numer of manuals assioned to the stress Group has varied, at no time were there 1ess than three conf. rolled manuals assigned to this Group.
On a broader level, the Sta'ff conce:n relates to Allegation 84 in $$ER 21 dealing with lack of management responsiveness to an engineer's request for a copy of controlled design procedures. The alleastion was discussed and resolved in SSER 21, with the Staff concluding that the " spirit of the allegation was substantiated" and t' int " management must improve its sensitivity in addressing safety concerns and improve communication with workers." In late 1982, there was an acknowledged shortage of copies of the manual, such that all angineers did not have individual copies. However, sufficient numbers of the controlled documents were available as discussed above and the engineers were able, and required, to use them. Additional copies have subseqcontly been made available, consistent with the goal of avoiding unnecessary complications in document control due to the distribution of more copies than necessary to accomplish the work.
8ecause the controlled design documents were, in fact, available to the
(. alleging engineer, there was no violation of procedures or adverse affect on the small bore pipfng analyzed. Nevertheless the Project has perceived tha desirability of improvement in this area, and,has taken several actions toward this end:
- 1. Document Control Procedures and practices are being reviewed with onsite
. Engineering personnel. They have been notified of the importalice of complying with document control procedures and of their responsibility to
update manuals and return acknowledgement forms.
- 2. Procedure P-1 was revised in Rev. 4 dated January 30, 1984 to require a monthly supervisory review of controlled manuals to assure that procedures, intructions and criteria are kept current. ,
- 3. For future revisions to design procedures, the supervisor will discuss the content of the revision with engineers under his supervision to be sure everyone ts aware of changes and how they are to be implemented.
Alternatively, procedure changes which are now routed to all manual holders will be formally routed to all engineers and will require an acknowledgment signature.
0 _u.
Also as a part of the resolution of DR 53-047-5, the ssible effect of
{' outdated design criteria documents on the final desi has been reviewed.
There were no instances found of out-of-date criteri in t% manuals. All individuals, including those missbg criteria documents, had access to current controlled copies of applicable criteria in order to correctly perform their design work. ,.,
As a separate effort, a Project QA review of configuration control of other annuals at OPEG (i.e., Engineering Manual. PEIs) has been coigleted. No l deficiences were identified in this review.
b) Use of External Documents
~
The staff questioned whether references, such as the following, in the possesion of Pipe Support Engineering personnel were used in lieu of approveo work procedures:
o An ION dated March 21, 1983 " Guidelines fcr Calculating Design of Skewed Welds" o Westinghouse Nuclear Technology Division Data for calculating double cantilever supports o Bechtel GPD STRUDL II Computer Program Users Manual CE-901 November 3, 1983 o Bechtel GPD 10tl dated November 11, 1980, "GPD Pipe Support Newsletter
( No. 5, Beta Angle" o Control Data Corporation (CDC) Bechtel National Support Manager to Civil / Structural Projects staff, " Baseplate II User Aids."
o Midland " Pipe Deflection Formula" o UE & C Pipe Support Design Standard August 15, 1979.
Experienced engineers connonly have general reference material as a part of their personal and professional library. This type of material includes textbooks and handbooks, and typically provides standard formulas and tables, code discussions, example calculations, rules of thumb and other simplified, conservative methods in cousson use in the industry. As general reference >
t material, they are not controlled and do not constitete acceptance criteria.
Project Engineering Procedures (EMP-3.3) provide for the use of references such as textbooks, catalops, monographs and other suc5 accepted industry techniques in specific ca culations. The reference must be documented when necessary to provide details of the design sufficient to allow independent review. In such cases, it is required that they be documented as formal references with the calculation in which they are used. Their use then is
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{' checked and approved via the calculation review and approval process. In the future, approvals of this material will be provided where general project standardization in their use is applicable. These materials will be fores 11 red, controlled, and included in procedures manuals with appropriate instructions, qualifications and limitations. ..
De above identified documents are references of the type normally found in an experienced engineer's personal library. We know of no instances where the references were improperly used. In one instance, a non-project document was referenced as the source of a double cantilever deflection formula used in a calculation. It was a standard engineering formula, not unique to any particular project, and need not have been referenced in the calculation.
c) Out-of-date Procedure Listings The staff also noted an instance of out of date procedure listings. An occurrence was observed where a controlled manual Table of Contents dated October 28, 1983 was in the possession of the Onsite Project Engineer, while i other supervisors had the previous version dated September 15, 1983.
This specific instance, ironically, resulted from management's efforts to improve the methods for distribution of revisions to controlled manuals.
Distribution of the October 28, 1983 revision was held by the Onsite Project Engineer upon receipt for two weeks while these improvements were being ,
formulated. The revised practices have since been incorporated into Piping Procedure P-1.
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(- Tssued my sne Project may have reflected inadequate design change procedures.
Response
,.. .:The project has in place formal procedures for requesting and approving design
'3- edhanges. These procedures do not permit design changes to be made on the
- ' basis of an interoffice memoranden (10M). The NRC's concern apparently
, relates to two identified 10Ms issued by Project Engineering. As discussed
.below, hovever, neither of the two memoranda constituted design changes.
The first ION involved the use of the weldinfl code (AWS) for calculation of skewed welds. The Pipe Support Group Superv'sor issued an ION dated March 21, 1983, for the purpose of providing guidance in modeling skewed welds in confonnance with the code. The 10M did not change any design documents, nor did it violate either good engineering precepts or approved QA procedures or requirements.
Tha. .:econd 10M of concern to the Staff was an 10M issued by Engineering on Oc % r 20, 1983, to General Construction, approving a request to revise a contractor's installation procedure. The change involved installation tolerances in the contractor's procedures which had Seen previously approved by Project Engineering in accordance with Project procedures for approval of contractor documents. General Construction and the contractor formally executed the change. Neither the request nor the 10M approving the change resulted in a change in the Project's approved design drawings or
(. specifications, thus, the issuance of a Design Change Notice was inapplicable. Project actions, including the 10M from Engineering approving the change in the contractor's procedures, were consistent with Project procedures for review, approval, and amendment of contractor documents.
9
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IstC Question: The NRC noted that design input had been received via telephone
( ano uses witnout written confirmation.
Response
Engineering Manual Procedure 6.1, Section 4.4, specifically provides that all design information provided verbally must be confirmed in writing. If the data are used prior to such confirmation, the calculations must be marked This
- preliminary,' and cannot be finally approved without such confirmation. '
requirement 10 an additional measure to assure that preliminary data are confirmed before the calculations are reviewed for final approval.
The calculations for Support 2156-200 noted the use of input loads received via telephone, but the originator failed to mark the calculation
' preliminary". When written confirmation of the input loads was received and compared to the input used, an error was noted. The calculation was performed again with correct inputs, and tiie support design remained acceptable.
Investigation and review of past audits show this occurrence to be an isolated case wh ch was clearly in violation of the engineering procedures.
IntC Question: The NRC has en ressed a concern that errors detected in several h- calculations which had been ciecked and approved may indicate that checking has not been properly performed.
Response
in[natureandsignificanceoftheerrorsfoundhavebeenpreviously
' discussed. The broad responsibility of the checker is to assure that the calculation is sufficiently accurate and sufficiently free of errors to serve its intended pur mse, i.e., to document that the support meets the design requirements. Tie minor nature of the errors detected and the fact that the calculations in question were corrected and still demonstrate support acceptability is a strong indication of the overall adequacy of the checking function.
Notwithstanding such a conclusion, the Project wishes to dispel the implication that discrepancies are " allowed" to exist or somehow disregarded, even though upon further analysis they do not affect the design adequacy.
Therefore, two actions are underway and will be completed by March 1, 1984.
First, it will be re-emphasized to Engineering personnel in writing, that calculational and documentation discrepancies will be dealt with seriously.
Originctors of documents are responsible for eliminating discrepancies.
Accordingly, they may not depend on the checker to accomplish this.
Second, recognizing that, in some cases, it is not economically justifiable to reperform an extensive calculation because of a discrepancy which will not i
affect the results or conclusions derived from the results, the En ineering
{; Procedure on calculations will be modified. This modification wil require that if the checker of a calculation detects an error which, in his judgment, can be classified as described above, the checker will identify the error, designate it as such, and initial the designation. This action is consistent with the requirements of ANSI M45.2.11-1974 which requires that analyses be sufficiently detailed that an experienced person can review them and accept the results without recourse to the originator.
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i utC Question: The NRC has raised a question regarding Licensee technical QA :
O tuaiss and surveillances with respect to the small bore piping support program. i Response :
In iglementing Criterion XVIII of 10 CFR Part 50, Appendix B the NRC has
' endorsed, with certain exceptions. ANSI M45.2 and ANSI M45.2.II. The latter i
document provides requirements and guidance for establishing a system of audits of quality assurance programs, and provides definition of various types of audits. Criterion XVIII mandates audits to verify c isnce with the QA program and to determine its effectiveness. O ne of the ove-cited references establish requirements for the performance of technical M audits.
! On the Diablo Canyon Project QA audits are conducted (in fulfillment of
. licensing commitments) to verify compliance with the project quality assurance l ,
program requirements. ,
The Project audit program has been developed and implemented to comply with requirements of the Project Nuclear Quality Assurance Manual. This program, t
1:1 turn, has been approved as being in compliance with Project requirements and Criterion XVIII of Appendix 8. It calls for a system of audits, the scope .
of which has been widely accepted in the nuclear industry, to assure that the QA program is properly functioning. Relative to the OPEG group, this audit '
scope has included all the mejor areas of design activity such as control of calculations, control of design drawings, indoctrination and training, and design change control. In addition, PGandE, as the licensee, has conducted a g series of Activity Audits covering 0 PEG activities.
Since 1982 there have been some nineteen (19) audits of DPEG to verify compliance with Project QA requirements. Closecut and corrective actions related to audits is documented in the Project audit files.
The verification of technical requirements in design output docuinents is performed by Engineering as part of the design control process. The type of verification can vary from ciecking to independent review by the Chief Engineer or an outside agency, depending on the significance of the document.
Specifically, reference is made to Procedure No. 3.4 (Design Verification),
i Procedure No. J.11 (Computer Programs), and other procedures related to .
i specific design documents (e.g., design calculations and drawings). These are all the responsibility of Engineering, are part of the design control process, j and are subject to Quality Assurance audit.
l While the Project's audit program is in full compliance with QA requirements in implementation of criterion XVIII, we believe that there is merit to the suggestion of formal, technical audits for OPEG. It is therefore olanned that a program of such audits will be immediately developed for OPEG, on the following basis:
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The audits will be. formally conducted and fully documented. They will o
- faclude all the features normally associated with QA program audits, such as entrancelexit meetings, checklists, and reports to management.
e The inftial audits will give special attention to those areas of most I sophisticated analysis, use and understanding of codes, use a:1d understanding of computer v ograms, independent checking, 6nd technical review of conventional worc. ,
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RICHARD C. ANDERSON L
E My nane is Richard C. Anderson. I an the Engineering Manager in the l Diablo Canyon integrated project organization consisting of Pacific Gas and Electric Company and Bechtel Power Corporation employees. I an a Registered ,
Mechanical and Nuclear Engineer in the State of California. I hold a BS ,
degree in Mechanical Engineering fron the University of California at Berkeley. .
I have been with Bechtel for nore than 26 years and for five years was assigned as an Engineering Manager in Bechtel's San Francisco Power !
Division, responsible for engineering work in the Pacific Northwest and f
( Japan. I have been assigned since March 1982 specifically to the Diablo :
Canyon Project to act as the Project's Engineering Manager. Prior to these Engineering Manager assignnents, I was the Chief Nuclear /Environnental {
Engineer for Bechtel's San Francisco Power Division, involved in nuclear power plant design, safety, and operation.
Prior to that, I was assigned as an Assistant Project Engineer on a =
proposed nuclear power plant project for PGandE and as Mechanical Group Supervisor, and later Project Engineer, on another large nuclear power plant project in the United States. Ther,e assignnents included supervision and coordination of design, specification, procurement, and quality control i activities.
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( I also served as Senior Mechanical Engineer for various other nuclear power facility projects in the United States and abroad, which included work in systens, safety, and equipnent engineering.
I have been an instructor in Bechtel's power plant courses for over 10 years and have given numerous talks and lectures in California on nuclear power and energy issues.
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( PROFESSIONAL QUALIFICATIONS OF f FRED C. BREISMEISTER I
My name is Fred C. Brr'isneister. I an Manager of the Research and Engineering / Materials and Quality Services (M&QS) group in Bechtel's San Francisco. Area Office. In this position I supervise and provide consulting services to the Diablo Canyon Project. I an a Registered Professional Quality Engineer in California.
My educational background is as follows: BS,1962, and MS,1964, in Metallurgical Engineering, Rensselaer Polytechnic Institute, New York.
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Prior to my duties as Manager in KaQS, I was supervisor of the Welding Engineering Section, where I was responsible for the development and technical content of Bechtel welding procedures and field fabrication standards, as well as technical support and direction to engineering and construction regarding welding, heat treatment, fabrication, inspection, and code problens.
I joined Bechtel in 1972 as a Metallurgical / Welding Engineer. I an an AWS D1.1 Certified Welding Inspector and a menber of the American Welding Society, the Str::ctural Welding Code Subconnittees 2 and 3, and the Preheat Task Force and Toughness Testing Task Group.
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PROFESSIONAL QUALIFICATIONS OF EDWARD M. BURNS My name is Edward M. Burns. My business address is Westinghouse Electric Corporation, P.O. Box 355, Pittsburgh, Pennsylvania,15230. I an employed as a Lead Engineer within the Nuclear Safety Department of the Nuclear Technology Division.
l From 1967 through 1971, I attended the Milwaukea School of Engineering and received a Bachelor of Science Degree in Mechanical Engineering. Following graduation I entered the United States Army and served as an ealisted nan, Lieutenant and Captain at several locations within the
[ United States and Europe. Fron March 1977 to August 1979, I served with the US Army Amor and Engineer Board as a project officer responsibic for the planning, conduct, analysis and reporting of operational tests of ground nobility, equipment, and ordnance.
I enrolled in 1977 in the University of Southern California night school progran and received in March 1979 a Master of Science Degree in Systems Managenent. On leaving the Arny in September 1979, I attended the University of' Wisconsin and received a Master of Science Degree in Nuclear Engineering in Decenber 1980. Additionally, from May to Decer.ser 1980, I worked as assistant to the head of the University of Wisconsin Fusion Studies-Progran. In this capacity, I was responsible for coordinating parametric studies input for a conceptual heavy ion bean fusion reactr.
-< 1270A
Following graduation, I was employed by Westinghouse' Electric Corporation in the Nuclear Safety Department. From initial employnent to Novenber 1983, I was a Senior Licensing Engineer, responsible for evaluating a
the compliance of engineered safeguards fluid systems and components with applicable safety and design criteria. Specifically, I reviewed the implementation of cold shutdown design improvements for five domestic and three foreign nuclear power plants. During this period, I also acted as the Westinghouse coordinator of licensing and safety activities related to the US NRC draft Regulatory Guide 1.139 and Unresolved Safety Issue A-45 prograns.
In December 1983, I was promoted to my current position of Lead Engineer, responsible for coordinating licensing services in support of nuclear power plants.
I an a menber of the Anerican Nuclear Society and the Anerican Society of Mechanical Engineers.
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PROFESSIONAL QUALIFICATIONS OF ,
DANIEL J. CURTIS My name is Daniel J. Curtis. I an a Onsite Project Engineering Group (OPEG) Plant Design Group Supervisor for the Diablo Canyon Power Plant. I have held the position since November 1983. My responsibilities have includod the supervision of the small bore piping qualification activities at the Diablo Canyon Jobsite under the technical direction of the San Francisco hone office. Small bore piping qualification activities include sna11 bore pipe stress analysis, small bore pipe support design, and piping isometric approval. I an a Registered Professional Civil Engineer in the State of California.
My educational background is a follows: BS in Civil Engineering, 1973, California State University, Chico.
I joined Pacific Gas and Electric in January 1974. From January 1974 to March 1976, I worked in the Design Drafting Departnent perforning structural analysis and design of niscellaneous structures. Fron March 1976 to June 1980 I was assigned to the Mechanical and Nuclear Engineering Departnent. Duties have included review and approval of pipe supports, developing design criteria for supports, coordination of work with consultants, and perforning piping analyses.
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E-In July 1980, I joined Science Applications, Inc. My duties included the seismic qualification of equipment and perfoming time history and response spectra analyses of piping.
In February 1981, I joined Bechtel Power Corporation. From February 1981 to March 1982 I worked on the Pipe Support Staff. Duties included providing technical assistance to projects, perfoming employee interviews, review and approval of project criterian, and other routine supervisory duties. From March 1982 to November 1983, I worked on the Diablo Canyon Project as the Project Large Bore Pipe Support Group Leader. My responsibility was the overall supervision of the pipe support calculations being performed on-project.
l In Novenber, I was assigned to the Onsite Project Engineering Group.
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PROFESSIONAL QUALIFICATIONS OF CHARLES W. DICK My name is Charles W. Dick. I am presently on special assignment managing engineering studies for new design concepts for nuclear power plants. Just prior to that I was Manager of Engineering ano Licensing for the Zinner Nuclear Plant with responsibilities which included nanaging design contractors for conpleting the project and for providing technical support of construction quality verification.
My previous assignnent was as a Project Manager and a nenbar of the project managenent tean of the Diablo Canyon Project consisting of the 4
integrated organization of Bechtel Power Corporation and Pacific Gas and Electric Company, and with responsibilities which included quality assurance.
I an a Licensed Professional Engineer in the states of California, New York, and Pennsylvania.
My educational background '.'s as follows: BS in Electrial Engineering, California Institute of Technology,1946; and MS in Electrical Engineering, Stanford University,1948.
I have also had additional training through Advanced Engineering Prograns in Business Administration and fron various technical and business courses..
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I Prior to my assignment to the Disblo C4 fon Project, I was Manager of Division Quality Assurance at the Bechtel Power Corporation from 1980 to 1982. Ny responsibilities included fomulating the QA programs for implementing such prograns, and training QA personnel for some 14 nuclear projects, I. joined Bechtel Power Corporation in 1965 and worked as a Project Engineer on various nuclear and fossil-fuel projects. I was responsible for project engineering work for a number of different types of power projects and studies, for nuclear standards development and for licensing. Beginning in 1973, I becane an Engineering Manager and subsequently Manager of Engineering, with overall managenent responsibility for the project engineering work on nere than 20 power plant projectL Prior to my enploynent at Bechtel, I was engaged as an engineer with the General Electrical Company beginning in 1948. During that tine, I was involved in narketing and appli:ation engineering related to nuclear power facilities. Prior to that I was assigned as an electric utility applications engineer and provided consultation services involving heavy electrical apparatus.
I an a senior member of IEEE and a menber of the American Society for Quality Cor. trol. I was also a member of the industry working group for development of ANSI Standard N45.2.11 (Quality Assurance Standards for Design of Nuclear Power Plants).
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PROFESSIONAL QUALIFICATIONS OF KENNETH C. DOSS i My nane is Kenneth C. Doss. As an employee of Pacific Gas and Electric Company since 1952, I am currently Senior Nuclear Generator Engineer participating in the systematic and independent review of Diablo Canyon Power Plant activities, which iricludes the review and evaluation of the technical adequacy of procedures and review and evaluation of design changes and modifications. I an also involved in the evaluation and assessnent of Diablo
- Canyon's and similar plants' operating experience and performance as related 4 to nuclear operating safety.
My educational background is as follows: AS in Electronics, Cuesta College,1969.
I joined PGandE in 1952 as a member of a line crew in the Electric Transmission and Distribution Department.
In.1955 I was assigned to the Morro Bay Power Plant as an Instrument Repairman and participated in the Startup of Units 1 and 2.
Subsequent assignments at the plant included Test engineer and Instrunent
~ Maintenance Forenan and participation in the startup of Units 3 and 4 and pre-startup check of plant control systems.
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In 1970, I was transferred to the Diablo Lnyon Project as a member of the Diablo Canyon Task Force engaged in startup preparation work at Humboldt Bay.
In 1971, I went to the Project jobsite as Instrunent and Control Supervisor and was pronoted as Senior Instrument and Control Supervisor in 1977.
Since September 1977 I have been a Senior Nuclear Generation Engineer Instrunent and Control Supervisor on the Diablo Canyon Onsite Safety Review Group (OSRG). My responsibilities ir.cluded preparation of training naterials for operators and technicians, including description of training naterials for operators and technicians, and instructions for control systens, nuclear instrunentation, and computers. I also participated in specifying test equipment and spare parts supplies for all instrunent and control systens.
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Q'JALIFICATIONS OF RD D. ETZLER My name is Ricnard D. Etzler. I am Project Superintendent at Diablo Canyon. I have held this position sini:e September 1978. I an responsible for manajing the onsite construction and startup activities at Diablo Canyon.
My educational background is as follows: BS in Mechanical Engineering, California Polytechnic State University,1967.
Prior to my duties as Project Superintendant, a was Resident Mechanical Engineer. I held that position from March 1977 to September 1978.
As Resident Mechanical Engineer, I was responsible for managing the mechanical type of construction activities such as ir.sta11ation of piping, ventilation systens, turbine / generator components and nuclear steam supply system components.
Prior to cy duties as Resident Mechanical Engineer, I was a Field Engineer and Group Leader reporting to the Mechanical Resident Engineer. I held this type of position and level of responsibilities from 1971 to 1977. My responsibilities included supervising installation of the nuclear steam supply and turbine generator systems.
Prior to my duties as a group leader for the Mechanical Resident Engineer, I was a Startup Field Engineer beginning in December 1969. My duties as a Startup Engineer included preparing preoperational startup testing procedures and scheduling tests.
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Prior to qy assignment to Diablo Canyon, I was in training to be a startup engineer since October 1968. This training included approximately 9 months startup experience at the Robert E. Girna nuclear power plant near Rochester, NY, and 6 weeks, reacto operator training at Westinghouse's Waltz Mill facility near Pittsburgh, PA.
Prior to October 1968 I was a field engineer at PGandE's Round Mountain 500 kV Sebstation for 3 aonths. Duties included planning construction activities, "as-built" drawings, and assisting in testing
. components, My firs,t assignaeat with PGandE was as a Field Engineer on die Construction of the Moss Landing Powar Plant Units 6 and 7. This assignment started in June 1967 and continued to July 1968. My duties included assuring installation of piping systens was in accordance with er lineering specifications and drawings.
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PROFESSIONAL QUALIFIC JIONS OF HOWARD B. FRIEND My name is Howard B. Friend. I have been empleyed by Bechtel since 1952. Since 1982 I have been employed by Bechtel Power Corporation as Project Completion Manager for the Diablo Canyon Project, an integrated effort between Bechtel Power Corporation and Pacific Gas and Electric Company. My responsibilities include managing the effort required for completion of the renaining services necesary to bring Units 1 and 2 of the power plant into commercial operation. The effort includes detemination of manpower and other resources for engineering, licensing support, procurement, construction, startup testing, project cost and scheduling and related services, as required. I am a registered Professional Engineer in the State of California.
My educational background is as follows: BS in Mechanical Engineering, Heald Engineering College,1952.
From 1981 to 1982 I was employed by Bechtel as Manager of Projects for the San Frsacisco Power Division. I also served as Project Manager for the South Texas Project (two 1250 MW pressurized water reactor [PWR] units),
responsible for the takeover of engineering, procurement, construction management, and related services.
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Fron 1979 to 1981, I was employed by Bechtel as Manager of Division Engineering. In that position I was responsible for directing all engineering of the San Francisco Power Division, including the design of both fossil-fuel ,
and nuclear power plants. My department was responsible for more than 22 major design projects.
Fron 1974 to 1979. I was enployed by Bechtel as Engineering Manager.
In that capacity I was responsible for Bowline Units 1 and 2, Skagit Unit 1, Syncrude utility plant and other utilities for the Syncrude Ter Sands Project, among others.
Fron 1972 to 1974, I was enployed by Bechtel as Project Engineer on other major projects, including Peach Botton Units 1, 2, and 3.
( Earlier assignments covered a variety of fossil-fired and nuclear 6 power plants in supervisory and technical capacities and in field assignments.
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( PROFESSIONAL QUALIFICATIONS OF JOHN M. GISCLON My name is John M. Giscion. I am the Technical Manager at the Diablo Canyon Power Plant. I have held this or equivalent position!, since February 1979. I an responsible for plant staff review and approval of plant modifications. I an a Registered Professional Mechanical Engineer in kvada and 3 Reaistered Professional Mechanical and Nuclear Engineer in California.
I hold an NRC Senior Reactor Operator's license on Diablo Canyon Unit 1.
My educational background is as follows: BS in Mechanical Engineering, University of Nevada,1961.
( After graduating from the University of Nevada, I served four years l in the U.S. Navy as an officer. I joined PGandE in 1965 and was assigned to l
the Pittsburg Power Plant as Engineering Trainee.
In 1966, I was tranferred to Hunbolt Bay Unit 3 with assignnents in nuclear power plant nuclear engineering, testing, and technical operations.
In 1968. I joined Westinghouse Electric Corporation (NRF - Bettis Atonic Power Laboratory) as a Plant Engineer. I held various assignments in maintenance and modification of equipnent and systems and served as design
-liaison for the liquid radwaste disposal systen.
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In 1970 I rejoined PGandE and was assigned to Humboldt Bay for startup preparation as a member of the Diablo Canyon Task Force. As a menber of the Westinghouse startup team, I was assigned to the H.B. Robinson Power Plant for three months.
I was a Power Production Engineer (Nuclear) fron 1971 to 1974. I participated in the preparation and review of ifcensing material for Diablo Canyon Units 1 and 2, including the FSAR, Technical Specifications, equipnent description and operating instructions, testing procedures, administrative procedures, and operational quality assurance manual.
Prior to att current duties as the Technical Manager, I was a Senior Power Production Engineer (Nuclear) from 1974 to 1979. I participated in the
( startup testing program and was responsible for supervising a staff of engineers (including persons experienced in nuclear engineering instrumentation, radiation protection, and chemical engineering) engaged in preparation of materiti required for plant startup, and in perforcing tasks related to startup.
I have completed the following fomal training courses: Reactor Physics for Engineers and Nuclear Reactor Eagineering (University of Idaho NRTS Graduate Education Program) Nondestructive Testing (General Dynaaics/Convair), Nuclear Power Plant Operator Simulator Training (Westinghouse Nuclear Training Conter, Zion, Illinois), Diablo Canyon Design Lecture Series and Station Nuclear Engineering Applications (Westinghouse),
and Management for Excellence Program (University of Santa Clara).
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( PROFESSIONAL QUALIFICATIONS OF JOHN B. HOCH My name is John B. Hoch. Since January,1982, I have been employed by PGandE as Diablo Canyon Project Manager. % responsibilities include managerial and supervisory duties, and providing coordination and direction of the Diablo Canyon Project organization. I am a Registered Professional Engineer (M2chanical and Nuclear) in the State of California.
My educational background is as follows: BS degree in Mechanical Engineering fron the University of Idaho,1959; graduate studies in Engineering, University of California, Berkeley,1961 to 1962; MBA, University of San Francisco,1969.
Fron 1980 to 1982, I was employed as Manager of the Nuclear Projects Department at PGandE. My responsibilities included managerial and supervisory duties, and providing coordination and direction of the Nuclear Projects Departnent in matters related to PGandE's nuclear power plants.
Fron 1977 to 1980, I was employed in PGandE's Engineering Departnent as Project Engineer for Diablo Canyon. My responsibilities included coordination of all Diablo Canyon Engineering activities.
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C Fron 1962 to 1977. I was employed as a Mechanical Engineer and as a Senior Mechanical Engineer in PGandE's Engineering Departnent. My responsibilities included engineering design and analysis work for both fossil-fueled and nuclear power plants. In add; tion, I was responsible for NRC licensing activities for PGandE's proposed Mendocino Power Plant and for the Diablo Canyon Power Plant.
From 1959 to 1961 I was employed by PGanaE in its Department of Electric Operations with responsibilities which included engineering analysis, supervision of instrument naintenance activities, and start-up activities associated with new fossil-fueled generating units.
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l PROFESSIONAL QUALIFICATIONS OF MICHAEL J. JACOBSON My nane is Michael J. Jacobson. I an the Project Quality Assurance (QA) Engineer for the Diablo Canyon Froject consisting of the integrated organization of Bechtel Fower Corporation and Pacific Gas and Electric company. I an a Registered Professional Quality Engineer in the State of California.
My educational background is as follows: Sacramento State College, BS in Civil Engineering,1970; and Golden Gate University, Business Managenent Certificate in Managenent,1979.
( I joined Bechtel Power Corporation in 1970 as a Quality Assurance Engineer resporsible for various aspects of the design phase quality assurance on a nuclear power plant project. I was subsequently responsible for perfoming structural design and seisnic analysis activities on the project.
Later, I was assigned as Project Quality Assurance Engineer responsible for supervising project QA activities, including direction of quality audits of construction activities.
Subsequently, I was assigned as Project QA Engineer on various other nuclear power plants, where I was responsible for directing project QA prograns. I was responsible for e.'suring that project construction and site activities, as well as quality control aspects, net applicable QA regulatory requirenants.
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i I was assigned to the Diablo Canyon Project ir.1982 to direct and control the DCP QA program.
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I 30FESSIONALQUALIFICATIONSOF MYRON E. LEPPKE My name is Myron E. Leppke. I an the Onsite Project Engineer on the Diablo Canyon Project consisting of the integrated organization of Bechtel Power Coorporation and Pacific Gas and Electric Conpany responsible for direction and control of the multidiscipline Onsite Project Engineering Groups at the Diablo Canyon jobsite. Prior to that, I was the Assistant Onsite Project Engineer of the same organization with the prinary responsibiltiy for the Plant Design, Record Managenent, and Docunent Control Groups. I an a Registered Professional Mechanical and Nuclear Engineer in the State of California.
My educational background is as follows: BS in Mechanical Engineering, University of Wyoning,1970; and MS in Nuclear Engineering,
. University of Wyoning,1971.
In August 1971, I became a Mechanical Systens Design Engineer employed by Pacific Gas and Electric Company on the Diablo Canyon Nuclear Project.
In Septenber 1977, I was transferred to the Diablo Canyon jobsite to becone the onsite Quality Assurance Supervisor. I had responsibility for nonitoring quality assurance activities in Construction and Operations.
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i In August 1979, I was transferred to the Diablo Canyon Construction Organization and assumed responsibility for direction and control of the mechanical and piping construction activities.
In June 1981, I was transferred to the Nuclear Power Generation Departnent with responsibiltiy for formation of the Onsite Safety Review Group. This group was formed in order to provide independent review of operational activities and plant design with a view towards engineered safety improvements.
In March 1982, I was transferred to the Onsite Engineering Group as a Senior Piping Engineer responsible for the Small Piping Design Reverification Program.
In Septecber 1982, I becane the Assistant Onsite Project Engineer.
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. . x O A.. E T-ix E .O. E iE -i;+i. m +v: a PROFESSIONAL QUALIFICATIONS OF LEO MANGOBA Hy nane is Leo Mangoba. I have been enployed by Bechtel since 1976.
Since October 1982 I have been a pipe support group leader at Diablo Canyon where I have been responsible for managing the design of ina11 oore piping supports.
I graduated with a Civil Engineering degree from Feati University, Manila,1972.
Prior to 1974 I was an engineering e:tinator with Calderon Construction Company.
Fron 1974 co 1976 I held a variety of assignnents working in pipe support engineering.
In 1974 I began working as a job shopper for Bechtel in the capacity of Pipe Support Engineer where I worked on design calculations for both large and snall bore pipe supports. In 1976 I was hired directly by Bechtel to perfom the sane function. In this capacity I was involved with the Fast Flux Test Facility and the Linnerick and Skagit projects.
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From 1977 to 1979 I was an Assistant Pipe Support Group Leader working on the design of large and small bore pipe supports. In 1979 I becane the Pipe Support Group Leader, nanaging the design of small bore pipe supports for the Monticello, Point Beach and Susquehanna projects. I, 'letotar 1982, I accepted an assignment in the same capacity with Diablo Canpn.
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. ,; r ~ ,- 1 i PROFESSIONAL OjiALIFICATIONS OF GARY H. MOORE My name is Gary H. Moore. I am the Unit 1 Project Engineer of the Diablo Canyon Project consisting of the integrated orgnaization of Pacific Gas and Electric Company and Bechtel Power Corporation. I have held this position since January 1982. I am responsible for the project engineering work related tc the design and analysis of Diablo Canyon Power Plant Unit 1. I an a Registered Professional Mechanical and Control Systems Engineer in the state of California.
My educational background is as follows: BS in Mechanical Engineering, San Jose State University,1968; and MS in Mechanical Engine 9 ring, San Jose State University,1969.
I joined PGandE in 1969 as a Mechanical Engineer in the Mechanical and Nuclear Engineering Departnent, designing instrunantation and control (IS0) systens for conventional fossil plants.
In 1977, I was naned a Senior Mechanical Engirteer supervising the 18C Group assigned to the Potrero Unit 7 Project.
In 1979 I was named Supervising Mechanical Engineer, supervising the 16chanical and Nuclear Engineering Department's entire I&C Group, including responsibility for the 18C design of the Diablo Canyon Power Plant.
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( I have completed the following fomal training courses: Sinulator Training, Westinghouse Nuclear Training Center, Zion, Illinois; and Westinghouse PWR Infomation Course.
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ROBERT G. OMAN My name is Robert G. Oman. I am an Assistant Project Engineer on the Diablo Canyon Project consisting of the integrated organization of Bechtel Power Corporation and Pacific Gas and Electric Company, responsible for the direction and control of the mechanical, electrical, instrumentation, and HVAC engineering groups. Prior to that, I was the Onsite Project Engineer with responsibility for overall direction of multidiscipline engineering group at the Diablo Canyon jobsite. I an a Registered Prufessional Mechanical and Nuclear Engineer in the State of California.
$ educational background is as follows: BS in Naval Science, U.S.
Neval Acadeny,1966; and U.S. Navy Nuclear Power School,1968.
After qualification as a supervisor of operations of Westinghouse PWR reactors, I served for three years as an engineering officer aboard a nuclear-powered submarine where I was responsible for the operation and >
maintenance of various reactor plant electrical and fluid systems.
I joined Bechtel in 1972 as a Nuclear Engineer on the Trojan Nuclear Project, becoming Nuclear Gre.'p Leader a year later, and Mechanical Group Supervisor a year after that. W duties included perfoming and supervising mechanical system design, licensing activities, and field coordination through startup to comercial operation.
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% next six years were spent in Spain as Nuclear Group Supervisor, Mechanical Group Supervisor, and Assistant Project Engineer on the Vande11os Nuclear Project. % duties included supervision of systens design, techn
- ly transfer, and assisting g Spanish counterpart in implementing project management tools and 14.roduction controls, and developing procedures for engineering interface with construction.
In 1982 I was assigned to the Diablo Canyor. Project.
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PROFESSIONAL QUALIFICATIONS OF ROBERT PATTERSON My nane is Robert Patterson. I am Plant Superintendent and Assistant Plant Manager at Diablo Canyon. I h:ve held this position sin e April 1980.
I am responsible for directing all activities of the Maintenance, Operating, and Chemistry and Radiation Protection Dep:rtnents at Diablo Canyon.
Prior to qy duties as Plant Superintendent, I was Supervisor of Operations. I held that position from 1971 to 1980.- As Supervisor of Operations I was responsible for supervising the operating staff in the preparation of equipment operating procedures and related material prior to the startup of the plant. I participated in the preparation and review of l
licensing material for Diablo Canyon Units 1 and 2 including PSAR, FSAR, and Technical Specifications. I was also responsible for directing the operating staff in performance of preoperational tests and three separate hot functional test prograns. For the Unit 1 startup, I received an NRC Senior Operator's License.
Prior to my duties as Supervisor of Operations, I was a member of the l
l Diablo Canyon Task Force from-1970 to 1971 engaged at Humboldt Bay in Diablo
' Canyon sta. tup preparation. My duties included preparing training materials, initial loading, and low-level testing procedures for pre-startup activities.
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From 1969 to 1970 I was assigned to Pacific Gas and Electric Company's (PGandE) General Office in license preparation for Diablo Canyon.
During this period, I was assigrx:d for seven months to the R. E. Ginna Power Plant. There I conducted a training progra:n for operators taking the AEC Operator Licens'e examination and participated in the preoperational testing program and review of test results for acceptance of systems. During my R. E.
Ginna assignnent, I also participated in initial loading, low-level physics testing, and power operation testing programs.-
Prior to this I was on special assignment for preparation of PGandE power plant operator's training program and related manual. I served in this capacity from 1968 to 1969.
Prior to special assignment, I was assigned to the Potrero Power Plant for startup of a 220 MWe conventional unit. I held various other assignnents in power plant engineering and other technical operations at Potrero. During this period,1964 to 1968, I was also reassigned to Humboldt day Power Plant during refueling outages to participate as a Shift Nuclear Engineer. At Hunboldt Bay I participated in prestartup activities including preparation of training materials, initial loading, and low-level testing procedures. I directed the preparation of reactor refueling procedures 1270A
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subsequent to initial fueling and directed the prefomance of this work on shift. I was responsible for the theoretical analyses of reactor core nuclear and thermal-hydraulic perfomance plus evaluation of the perfomance of plant safeguard and other auxiliary equipment. From 1961 to 1964, I was assigned to other technical operations at Humboldt Bay and served in various assignnents in power plant nuclear engineering.
Prior to my Humboldt Bay assignnents, I was a staff engineer fron 1959 to 1961. In this capacity I was assigned to both the Vallecitos and dresden projects. At Vallecitos I observed various phases of plant operation including the initial startup of the AVBWR. At Dresden I participated in initial loading and low-level testing and half-power to full-poser testing.
Prior to Vallecitos and Drescien, I had various assignments from 1955 to 1959 involving power plant engineering and technical operations. I was involved in a conventional power plant startup.
I graduated fron Cocper Union School of Engineeing, New York, in 1953 with a BME. I an a registered Professional Nuclear Engineer in California.
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PROFESSIONALQUALIFICATIONSg R. KEITH RHODES My name is Keith Rhodes. I am Technical Services Supervisor with the General Construction Station Department Instrument and Control (I&C) Group. I have held this position since January 1,1980. I am currently assigned to the Diablo Canyon Project Startup Department and am respnsible for directing activities-of the Instrument and Control Group.
biy educational background is as follows: AS degree in electronics, Cuesta College, California,1976. ,
During the period from June 1980 until May 1983 I was assigned to the Technical Mrvices I&C Group in Emeryville, California. I was responsible for supervising the I&C personnel at various job sites on work assigned to General Construction Station Department, including the Diablo Canyon, Geysers, and Helms Projects.
-I was made a Field Engineer in 1975 and was responsible for supervising activities of the Diablo Canyon General Construction I&C Group. I was also responsible for directing contractor instrument installation and valve maintenance work.
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In 1972 I was made a General Construction Technical Subforeman and f
assigned the responsibility of directing the contractor, S&Q Construction i-performing instrument installation work.
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l From 1967 until 1970 I was self-employed. ;
1 I initially joined PGandE's East Bay Division in 1962 and was an Apprentice Instrument Repaiman at the Pittsburg Power Plant.
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PROFESSIONAL QUALIFICATION OF JAMES D. SNIFFER My name is James D. Shiffer. I am the Manager, Nuclear Plant Operations, and as such provide line management support to the Diablo Canyon Power Plant. $ organization is responsible for all operations, maintenance, operational engineering, training, security, quality control, energency planning, and radiation protection activities at the plant. I an a Registered Professional Mechanical and Nuclear Engineer in California.
My educational background is as follows: BS in Chenical Engineering, Stanford University,1960; and MS in Nuclear Engineering, Stanford University, 1961.
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l I joined Pacific Gas and Electric Conpany in 1961 as a Nuclear My Engineer assigned to the startup preparations for Hunboldt Bay Unit 3.
duties included preparation of training material, initial and low-level testing procedures; training of operating personnel for AEC license examinations; directing initial loading and testing programs as Shift Nuclear Engineer, and various other operational engineering assignments during the period between 1961 and 1969.
In 1969 I was transferred to the startup preparation for the Diablo Canon plant which included a seven-nor:th assignment to the startup and l
initial testing of the R.E. Ginna PWR plant.
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In 1980, I was appointed itanager of the newly fomed Nuclear Plant Operations Department.
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PROFESSIONAL QUALIFICATIONS OF LAWRENCE E. SHIPLEY My name is Lawrence E. Shipley. I am a Techaical Consultant to the piping program at the Diablo Canyon Project. I have held this position for sixteen months. My primary responsibility is in the review of piping systens to licensing commitments and newly developed seisnic criteria.
My educational background includes the following: BS in Mechanical Engineering, U.S. Merchant Marine Academy, New York,1965.
1 I joined Bechtel Power Corporation's Scn Francisco Power Division in 1967 in the field of piping stress analysis. My responsibilies included technical direction of 150 engineers and designers on projects that included nuclear and fossil-fired power plants and the liquid metal fast breeder reactor at the Fast Flux Test Facility at Richland, k%shington.
In 1981, I.becane the Assistant Project Engineer on the Susquehanna Steam Electric Station in Pennsylvania, responsible for engineering in the civil-stt-uctural, architectural, and piping and plant design areas. The work I directed included: structural analysis review of all Seismic Category I buildings, piping / stress analysis review of all Seismic Category I buildings, piping / stress analysis and pipe support design, valve qualification, welding and NDE, and materials selection and qualification.
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In 1982, I was appointed Technical Consultant to the Diablo Canyon Project for the piping program.
In 1983, ray duties were expanded to include those of Assistant Chief Engineer for Plant Design in the San Francisco Power Division.
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PROFESSIONAL QUALIFICATIONS OF AZRIEL SHUSTERMAN My name is Azriel Shusterman. I have 23 years of experience as a mechanical engineer, the majority of it in the design of piping and pipe supports. Since August 1982, I have been employed by Bechtel's San Francisco P,ower Division and have worked on Diablo Canyon Unit 2. In October 1982, I worked with the jobsite's small bore piping design group in a supervisory capacity.
I graduated with a Mechanical Engineering degree from the University of Riga, Latvia, in 1961.
From 1961 through 1964 I was a Mechanical Engineer employed by the f
Diesel Manufacturing Plant of Riga, Latvia.
From 1964 through 1978 I worked at diga's Special Project Institute i of 011 and Industry where I was responsible for the engineering and design of piping, piping layout, pipe supports, and pipe stress analyses as well as the i
fabrication and installation of pipe supports. I also had interim assignments as a Senior Engineer in a plant that manufactured special tools, molds, and dies.
From 1980 to 1982 I was employed by Quadrex as an engineer on the Zinmer and Susquehanna projects. In this capacity, I was responsible for pipe .
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support design and piping walkdown inspections. At Susquehanna, I was also responsible for the technical review of small bore pipe support designs.
I accepted employnent with Bechtel on the Diablo Canyon Project in August 1982.
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( PROFESSIONAL QUALIFICATIONS OF ROBERTC.THORNBERg My name is Robert C. Thornberry. I an Plant Manager of the Diablo Canyon Power Plant. As such, I an responsible for ensuring that the plant is operated in a manner consistent with the safety of the plant personnel and the general public and in accordance with the license granted by the Nuclear Regulatory Cormission. I an also responsible for direct supervision of all administrative functions. I an a Registered Professional Nuclear Engineer in California.
My educational background is as follows: BS in Chemical Engineering, 1962, and MS in Nuclear Engineering,1963, Georgia Institute of Technology.
I joined Pacific Gas er.d Electric Company in 1980 as Project Design Coordinator for the Diablo Canyon Power Plant, responsible for the project design activities, i Prior to that, in 1979, I was an engineer with Atonic Energy of Canada, Ltd., responsible for safety studies for 600 MW CANDU reactors.
In 1976, I was employed by the San Diego Gas and Electric Company as Supervisor of Nuclear Licensing responsible for all aspects of licensing, including directing the support efforts of the NSSS supplier, architect-engineer, and other project consultants in the licensing process.
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In 1972, I joined the General Atonic Company where I worked on high-tenperature, gas-cooled reactor safety analysis reports.
After gradu6 tion in 1963. I joined the E. I. Dupont Company where I j spent four years at the Savannah River Plant, monitoring the daily perfomance and safety of heavy water reactors, investigating unusual operating conditions, reviewing operating procedures, and calculating core operating parameters. For the five years subsequent to this, I was assigned to the Savannah River Laboratory where I worked en the design and analysis of fuel and target assemblies and directed a study and redesi 5n of the energency core cooling syster.
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PROFESSIONAL QUALIFICATIONS OF MICHAEL R. TRESLER My name is Michael R. Tresler. I am the Assistant to the Unit 1 Project Engineer on the Diablo Canyon Project, consisting of the integrated organization of Pacific Gas and Electric Company and Bechtel Power Corporation. In this position I am responsible for assisting the Project Engineer in directing all engineering on the unit with the exception of i
licensing-related efforts and other special activities. I have also been associated with the Project as Resident Mechanical Engineer, Project Superintendent, Assistant Station Construction Superintendent, Project Control Engineer, and Piping Design Coordinator.
- My educational background is as follows: BS in Mechanical Engineering, California Polyteche.ic State University,1964.
I joined PGatidE in 1964 and performed pipe analysis and support design, and construction inspection, design, and startup of large fossil-fired units.
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In 1969, I spent a year participating in the startup and initial l
testing of the R.E. Ginna PWR Plant in Rochester, New York.
In 1970, I became PGandE's Lead Engineer in the piping design and quality assurance areas.
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I joined the Diablo Canyon Project in 1972 as Resident Mechanical Engineer, becoming Project Superintendent in 1977.
In 1979, I spent a year as Assistant Station Construction Superintendent with responsibility for Diablo Canyon and niscellaneous fossil-fired construction work.
In 1980, I returned to Diablo Canyon as Project Control Engineer and was appointed Piping Design Coordinator in 1981 with the responsibiliy for controlling all piping and support design work on the Project.
I assuned my present duties in Dctober 1983.
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WILLIAM N. WHITE i I
My name is William N. White. I am an Assistant Project Engineer in the Diablo Canyon integrated organization consisting of Pacific Gas and Electric Company and Bechtel Power Corporation employees. My responsibilities include supervision and direction of seismic-related engineering analyses for the Diablo Canyon Unit 1 P'roject Engineering Organization. I an a Registered Professional Civil Engineer in Oregon and member of the American Society of Civil Engineers.
My educational background includes: BS, Civil Engineering, University of Idaho; MS, Civil Engineering, University of Colorado; PhD, Civil
~ Engineering, University of Colorado.
For the past seven years I have been an engineering specialist with Bechtel's San Francisco Power Division working with the Chief Civil Engineer's i
staff-in the area of seismic analysis for several Bechtel projects.
Earlier, I was a Structural Engineer with the Tennessee Valley
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Authority where I was responsible for seismic analysis of all Category I b structures for a twin-unit nuclear power plant, including seismic input for the design of the nuclear steam supply system.
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I was an Assistant Professor at Oregon State University where I taught undergraduate and graduate courses in structural mechanics and analysis and computer applications. I performed a special study for Bechtel or, soil-structure interaction for the proposed Mendocino nuclear power plant while teaching at Oregon State University.
While employed at the Bettis Atomic Power. Laboratory, I was a Senior Engineer working on shock analysis of nuclear reactors aboard submarines and was ir.volved in programs to assess the shock resistance of reactor internals subjected to long-tem irradiation damage.
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