ML19256D969
| ML19256D969 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/09/1968 |
| From: | Boyd R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | HEALTH & HUMAN SERVICES, DEPT. OF, U.S. PUBLIC HEALTH |
| References | |
| NUDOCS 7910230917 | |
| Download: ML19256D969 (1) | |
Text
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W E-UNITED STATES ATOMIC ENERGY COMMISSION l
WASHINGTON, D.C. 20545
' inn e l Docket No. 50-289 D-} } J
'}D D
h IN REPLY FIEFER Tot stribution:
APR 9 1968 AEC Doc. Room Formal Suppl.4 Surgeon General, U. S. Public Health Service DRL Reading Departament of Health, F/,ucation, and Welfare RPB-3 Reading Attention: Nuclear Facility Analysis Section Orig: FWKaras 1901 Chapman Avenue R. S. Boyd Rockville, Maryland 20852 T. F. Engelhardt, OGC Gentlemen:
In accordance with the understandings reached Augus t 1,1961, between the Atomic Energy Consission and the Departament of Health, Education, and Welfare, attached for your information are the following docussenta:
MET 30POLITAN EDISON COMPAhT (Thraa Mile Island Nuclear Station Unit 1)
Report from the Advisory Counittes on Reactor Safeguards dated April 6,1968, concerning the Caumittee's review of the application for a license to construct a nuclear povar reactor in Dauphin County, Pennsylvania.
Sincerely yours,
(,
.2. v ger S. Boy 4cATTistant Dir ctor for Reactor Projects Division of Reactor Licensing 1454 198 J
- -lo.utes' As stated above RPB 3/DRL o,ricr >
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4/9/68 om >
Wurm ABC-318 (Rev. 6-43) e, s. esemesent marine emes to-et et-e Gro,2 2 4?C
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NUCLEAR REGULATORY COMMISSION REGION 11 e ;J b;#
101 MARIETT A ST., N W, SUIT E 3100 r
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AT L ANTA, G E ORGI A *,0103
.....f In Reply Refer To:
RII:CEA 50-369/78-20 50-370/78-15 Duke Power Company ATTN:
W. O. Parker, Jr.
Vice President, Steam Production P. O. Box 2178 Charlotte, N C 28242 Gentlemen:
This refers to the inver*igation conducted by M. V. Annast of this office during April 11 - December 6, 1978, of activities cuthorized by NRC Construction Permit Nos. CPPR-83 and CPPR-84 for the McGuire facility. The results of this investiga-tion were discussed in correspondence sent to you on August 1, 1979 by the Director, Division of Reactor Construction Inspection, Office of Inspection and Enforcement.
In accordance with Section 2.790 of the NRC's " Rules of Practice", Part 2, Title 10 Code of Federal Regulations, a copy of this letter and the enclosed investigation report will be placed in the NRC's Public Document Room. If this report contains any information that you (or your contractor) believe to be proprietary, it is necessary that you make a written application within 20 days to this of fice to withhold such information from public disclosure. Any such application must include a full statement of the reasons on the basis of which it is claimed that the information is proprietary, and should be prepared so that proprietary information identified in the application is contained in a separate part of the document. If we do not hear from you in this regard within the specified period, the report will be placed in the Public Document Room.
Should you have any questions concerning thin letter, we will be glad to discuss them with you.
Sincerely,
\\
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F Long Acting Deputy Director
Enclosure:
Investigation Report Nos.
50-369/78-20 and 50-370/78-15 1454 199 cc w/ enc 1:
(See Page 2)
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Duke Power Company AUG 0 71973
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cc w/ encl:
M. D. McIntosh, Plant Manager
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P. O. Box 488 Cornelius, N C 28031 J. C. Rogers, Project Manager P. O. Box 33189 Charlotte, N C 28242 1454 200
UNITED STATES
[pte 88 C oq'c NUCLEAR REGULATORY COMMISSION
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IE Investigation Report Nos. 50-369/78-20 and 50-370/78-15
Subject:
Duke Power Company McGuire Nuclear Plant Units 1 and 2 Docket Nos. 50-369 and 50-370 Allegation that certain formal written statements made by Duke Power Company to the NRC were incorrect and matetrolly affected the staff's review and acceptance of a revision to the McGuire Final Safety Analysis Report (FSAR).
Period of Investigation:
April 11-13, May 3-5, July 6-28, September 10-11, November and Decembe
-6, 1978 I
Investigators:
)
W2
\\1 Aq gj M. t. Annast, Investigator Late Office of the Director h,F-1la.. E vbil15 M. D. Hunt,'7rincipal Inspector Batel Projects Section Reactor Construction and En 'neering Support Branch
///;tS/7 Reviewed by:
f f
D. Thompf n M eputy director Date' 4
Office 6f the Director 1454 201
TABLE OF CONTENTS
- 1. INTRODUCTION II. SCOPE OF INVESTIGATION III. CONCLUSION IV. DETAILS OF INVESTIGATION A.
Persons Contacted B.
Allegation, Discussion and Finding C.
Exhibits 1.
Copy of Disputed Calculations 2.
Specific NRC Questions 3.
Duke Power Company Response to NRC Questions 4.
NRR Memorandum, dated September 14, 1978 4
202
rf I. INTRODUCTION The Nuclear Regulatory Commission in sd[eral telephone conversations and by letter received information iydicating that safety related design calculations were inadequate and certain design analyses were incorrect or misleading.
The Region II Office of Inspection and Enforcement conducted an inves-tigation into the allegations and reported the findings in IE Investi-gation Report Nos. 50-369/78-10 and 50-370/78-4 dated August 2, 1976.
This report concluded that of twelve separate allegations, eleven were not substantiated; however, one allegation resulted in an unresolved item. This allegation was stated as follows:
" Duke Power Company (DPC) has an existing accident analysis which indici.tes that a dropped fuel shipping cask can fall into the spent fuel pool during certain conditions. DPC representatives have info,med the NRC and stated in the FSAR the fuel cask cannot fall into the pool.
Region II continued the investigation of the unresolved item regarding the cask drop accident analysis as authorized by Part 1.64, Title 10, Code of Federal Regulations.
II. SCOPE OF IhTESTIGATION The scope of this investigation included the following:
A.
Interviews with present and former Duke Power Company employees.
B.
An examination of documents and records related to the allegation.
C.
An evaluation by NRR of the FSAR submittal with regard to a material false statement.
III. CONCLUSIONS Duke Power Company (DPC) performed several accident analyses, requested by NRR, involving the spent fuel shipping cask. One of the calculations, based on conservative assumptions, indicated that the cask could enter the spent fuel pool under certain circumstances.
DPC engineering management reviewed the facts and believed that this calculation was hypothetical and unrealistic. The supervisory engineer reviewed this issue and, based on his experience and professional judgment, prepared formal FSAR revision to the Office of Nuclear Reactor Regulation a
(NRR). This response to NRR questions concluded, without any reserva-tions, that the cask would not enter the spent fuel pool. Review by NRR resulted in the conclusion that the DPC submittal was material to the FSAR review process because it omitted discussion regarding the disputed calculation, which if known to exist would have resulted in the need for additional NRR review.
I454 203 nt
IV. DETAILS OF INVESTIGATION A.
Persons Contacted L. C. Dail, Vice President, Design Engineering J. R. Wells, Corporate QA Manager L. R. Barnes, QA Manager, Construction E. Braf ford, Design Engineering C. L. Ray, Jr., Design Engineering R. B. Priory, Design Engineering In addition to the above, three fermer DPC employees were inter-viewed.
B.
Allegation, Discussion and Finding Allegation: DPC has an existing accident analysis which indicates fuel that a dropped fuel shipping cask can fall into the spent pool under certain conditions. DPC representatives have informed the NRC and stated in the FSAR that the fuel cask cannot fall into the pool.
Discussion:
The licensee had been asked by NRR to provide the results of an evaluation demonstrating that dropping of the spent fuel shipping cask in its vertical or tipped position would not result in the cask entering the spent fuel storage area.
The investigators examined DPC calculations, including three postulated hypothetical 100-ton cask of the maximum accidents assuming a The envelope dimensions of possible casks to be used at McGuire.
first case assumed that the cask breaks loose from a height of two feet when the crane hits the end-of-travel stops at maximum speed. The second case assumed that the crane is traveling at maximum speed with the cask at floor level when the cask breaks loose and strikes the forward edge of the cask storage pit. In these two instances, the calculations indicate the cask could not tip over into the spent fuel pool. The third calculation assumed that the stationary cask is dropped from a height of 2 feet 11 inches, with the back edge landing on and pivoting about the back edge of the cask storage pit. This calculation indicates that the cask would fall into the pool.
The results of the study provided by DPC in Revision 6 to the FSAR, paragraph 9.1.2.3, included only the first two postulated accidents described above.
The stated conclusion was: " Based upon this evaluation it is concluded that the cask would not enter the spent fuel pool due to maximum horizontal swing or tipping of the cask." No potential limitations on casi handling such as special cask travel patterns, maximum lif t height or modifications to fuel storage area structure were mentioned. The third calculation (Exhibit C-1) was preliminary in that it was checked, but not approved. Tne calculation checker had made 1454 204 7
notations on calculation sheet numbers 7, 7a, 9 and 10 disa-greeing with the basic assumptions made by Individual A, the person performing the calculation.
The original calculator's assumptions were:
1.
No energy loss due to impact.
2.
Friction is not considered.
3.
Material deformation is not considered.
4.
Cask is infinitely rigid and does not deform.
The calculation checker, Individual B, had noted that: " Reinforced concrete has the ability to absorb energy. Therefore, energy is lost on impact. Material deformation does occur. The cencrete spalls off near the edge and the cask can be deformed if dropped."
Individual B had left DPC in August, 1977 and could not be imme-diately located, to verify his notes.
His written comments appeared, however, to be adequate to support his position.
The immediate supervisor of both men, Individual C, and the principal engineer, Individual D, agreed with the checker that the assumptions were unrealistically conservative.
No attempt was made, however, to recalculate the postulated load case using more realistic assumptions. The principal engineer prepared the input for Revision 6 to the McGuire FSAR in response to NRR Question 020.2 and supplied it to the licensing engineer for submittal to hRC. The licensing engineer confirmed this subn.ittal.
Thus, the submittal to NRC was made when persons in responsible positions were aware of the existence of admittedly conservative calculations that tended to cast some doubt on the validity of assurances included in the submittal.
Duke Power Company representatives stated that until this disputed calculation was performed all previous calculations were based on the assumption that the majority of the kinetic energy of the cask would be absorbed at impact and the patential energy of the cask would cause it to fall away from the spent fuel pool. Load cases No. 1 and 2, previously discussed, and three additional load cases, postulated by Individual B, the checker, were analyzed.
The conclusion was that the cask would not fall into the spent fuel pool under any of the postulated accidents.
Individual A, the author of load cases I and 2 was then assigned as the checker of the three additiop31 load cases calculated by Individual B.
In lieu of checking these three calculations, Individual A instead postulated a Load Case No. 3 and on June 11, 1975, performed an analysis for this new load case. Individual A used the basic assumptions previously considered for load cases 1 and 2.
Using the assumptions of no loss of energy from impact, 1454 205
S no consideration of friction or material deformation and an infi-nitely rigid cask, he concluded that the cask would fall into the spent fuel pool. The calculations were checked on June 12, 1975 by Individual B who apparently noted his objections on various calculation sheets.
Individual C, the immediate supervisor, was interzir ed by the investigators and stated that he agreed with the checker and brought the disputed calculation to the attention of his super-visor, Individual D, the Principal Engineer. They discussed the assumptions and Individual D also agreed that the assumptions were unrealistic. There was a subsequent meeting with Individual A to determine Individual A's reasons for the conservative approach in the calculations. During this meeting, Individual A said that since there was no firm information on the cask, he thought a conservative approach should be used. Individual D returned the calculation to Individual C and indicated that a final resolution could not be reached until c11 parameters identifying the cask to be purchased in the future, could be obtained. Individual C and Individual D discussed several ways to preclude the cask from falling into the pool.
These included:
(1) administrative controls on the operation of the crane transporting the cask; (2) modification of the north wall of the upper cask storage level by sloping the wall to a degree such that it would preclude the cask from dipping into the pool; and (3) modification of the concrete in certain areas between the decontamination pit and the upper level of the cask storage area. Individual C placed the calculation back in the files with the understanding that the matter would be resolved when the actual cask was identified.
Individual D, the Principal Engineer no longer employed by DPC, was interviewed by the investigators. He stated that in 1975 he was the senior supervisor responsible for the review and resolution of NRC questions relative to the cask drop accident as defined in Question 020.2 of the McGuire FSAR (Exhibit C-2).
Individual D clearly recalled the details involved in the disputed calculation and stated that he had given serious attention to the difference in opinions and had discussed the whole issue, in detail, with all parties involved. He felt that the assumptions made for Load Case 3 were unrealistic. Individual D believed along with Individual C and Individual B, that if more realistic yet conservative assumptions were used, the analysis would show that the cask could not enter the spent fuel pool.
However, actual calculations were not performed to verify this belief. The calculations were to remain open for a final review when the physical details of the actual cask could be defined.
In Individual D's opinion, the dieputed calculation had received appropriate and proper attention by qualified supervisory personnel and the final conclusion was based on proper review and good engineering judgment. Individual D's position was that, based on the foregoing actions, the statement which he submitted for the FSAR revision was correct; i.e., the cask could not enter the spent fuel pool (Exhibit C-3).
1454 206
Duke Power Company arranged for a formal briefing of events involving the cask drop accident calculations and to explain the history of the disputed calculation.
Further information and documents were provided voluntarily and in full cooperation with the investigators.
Region II fo_ warded a memorandum, with enclosed statements and documents involving the cask drop issue to Headquarters, Office of Inspection and Enforcement on July 28, 1978 requesting review of the case. The facts submitted by Region II were summarized as follows:
1.
Duke Power Company had information that, using conservative assumptions, the cask could conceivably enter the spent fuel pool.
2.
This information had not been provided to the Office of Nuclear Reactor Regulation (NRR).
3.
No recalculation was made of Load Case 3 using more realistic assumptions.
4.
Duke Power Company's decision not to include the disputed Load Case 3 as part of the FSAR submittal was a conscious act.
5.
No formal review procedure had been established which specifically required that later calculations, using actual cask design parameters, would consider the load case in question.
The Office of Inspection and Enforcement reviewed the case and forwarded tne documents to the Office of Nuclear Reactor Regulation (NRR) for a determination of materiality. NRR concluded that the omitted information would have been material in the FSAR review process and could have affected the staff's acceptance of the FSAR revision.
NRR concluded that the response to the staff question was a material false statement by omission (Exhibit C-4).
Region II was asked to provide additional information and docu-mentation concerning the cask drop accident analysis.
This included further interviews with the principals directly involved in the cask drop analysis. The following additional effort was requested in support of Headquarters review of the case:
1.
Determine various individuals' background and experience.
7.
Obtain individuals opinions regarding the performance of the calculations.
1454 207 4
e 3.
Reverify all facts and statements. Obtain additional state-ments.
4.
Attempt to locate Individual B, the checker of the disputed calculation.
Region II continued the investigation and reinterviewed Individual A, no longer employed by DPC, who performed the original calculation.
At this time Individual A was given a copy of the calculation to examine. He confirmed that he was the author of the calculation and explained his rationale in reaching the conclusion that the cask would f all into th_ spent fuel pool. He readily admitted that the assumptions he had used for Load Case 3 were ecnservative but pointed out that the same assumptions were used for Load Cases 1 and 2.
Individual A stated that he did not recall the notations made in the calculations by Individual B, the checker.
Individual A said that Individual B was a young inexperienced engineer, recently graduated from college, at the time he checked the calculation.
According to Individual A, the checker only perfomed a mathematical verification of the calculation and "did not express any strong opinions about the assumptions." Individual A commented that he would have initialed any additional sheets which would have been added to the calculation by the checker.
This was not the case with page 7.a which contained Individual B's disagreement with the assumptions.
Further examination of the document revealed that calculation sheets (Form 184), used for the calculations in question, were a version of Revision 12-74 i.e. the sheets were printed in December, 1974.
The calculation was performed on June 11, 1975 and was checked on June 12, 1975. Page 7.a was also dated June 12, 1975, however this page was a version of Form 184 revised in May 1976.
This meant that Individual B could.not have recorded his objections to the assumptions on page 7.a. on June 12, 1975 because this particular page was not in existence.
Individual B was located in Springfield, Virginia and was inter-viewed on October 10, 1978. Af ter initially denying any knowledge of the questionable page, Individual B admitted that he had recorded his backdated comments approximately one year later, apparently in order to please Individual C, his former supervisor.
He stated that, initially, in 1975 he checked the mathematics of the calculation and found them to be correct. Although he did feel that the assumptious used by Individual A were unrealistic and thus conservative, he made no further comments at that time, partially because he w:.s inexperienced and deferred to Individual A's experience.
In June, 1976, Individual C told him to go over the figures again, this time very carefully.
Individual B stated that he was not directly ordered to add any comments but felt that to document his comments at this time would please supervision and enhance his career with DPC. At the same time, the additional 1454 208 x
t rf year's experience had given him confidence to state that the issue at hand seemed trivial and th ' assumptions were, in fact, too conservative.
Individual B r hecked his own figures and found them to be correct. He thery added the comments on pages 7, 7a, 9 and 10.
Page 7a was a new page which he inserted at this time at.* dated this page to agree with the rest of the pages.
Individual B then returned the calculation to Individual C without comment. He recalls no further discussions on the matter.
The investigators reinterviewed Individual C on November 8,1978.
When confronted with the backdated page 7a, Individual C stated that he had asked Individual B approximately one year after the original calculation was performed and checked, to take another close look at it and to document any comments he might have.
Individual C confirmed that Individual B gave the calculation back to him with the added comments but maintained that he did not notice the backdated date and did not instruct Individual B to backdate anything.
He explained that the reason for his request for the recheck was the fact that Individual B had been transferred to another design group and was also planning to attend graduate school.
Individual C stated that he wanted the checker to review the calculation again to record his opinions before he departed. Individual C returned the calculation to the files. He stated that he made Individual D's successor in the design group, Individual E, aware of the calculation, insofar as telling him about the preliminary states and disputed nature of the document.
Further investigation disclosed that, with the exception of other DPC employee was apparently aware of Individual C, no the later addition of page 7a and the other comments on pages 7, 9, and 10.
Individual D, the Principal Engineer, was reinterviewed on October 5, 1978. He still maintained that the assumptions made by Individual A were unrealistically conservative. His basis for this was the fact that major energy dissipation would occur during any cask drop accident. He stated that concrete is commonly accepted as a good energy absorber and that the cask would also deform on impact, resulting in energy dissipation. Individual D is presently the head of his own engineering consulting firm. His background and experience include thirteen years involved in nuclear power plant design including direct experience with cask accident evaluations.
Individual D stated that, in his opinion, there were adequate internal controls withic the DPC derign engineering organization to assure future review of the preliminary calculation when an actual cask had been sele:ted.
Individual D stated frankly that, based on his experier.ce and engineering judgment, in retrospect he would still have provided the input for the FSAR submittal without reservations.
1454 209
t 9
Additional review of documentation and discussions with DPC technical staff during the course of the additional investigation revealed that DPC had still not decided on the final specifications of the spent fuel cask. During discussions about the cask selection the investigators found that Duke Power Company Quality Assurance Manual, Section PR-301, paragraph 5.9, requires formal approval of all engineering divisions for all specifications of safety related equipment which is to be purchased. The cask specifi-cations when finalized would, therefore, receive a formal review and this process would apparently include the resolution of any outstanding calculations.
Findings: The allegation is substantiated as stated. Duke Power Company formally submitted information to the Nuclear Regulatory Commission in the form of a revision to the Final Safety Analysis Report which omitted the fact that an existing preliminary accident analysis calculation indicated that the cask could fall into the spent fuel pool under certain circumstances. This constitutes a material false statement by omission. The investigation disclosed that the omission was based on engineering judgement of an experi-enced supervisor a'ter appropriate review with all parties concerned.
Several internal 1 rocedural controls exist which tend to give reasonable assurant-that the cask drop issue would be addressed when the final cask specifications are finalized. There is no evidence which indicates a willful or iatentional attempt by DPC to deceive the NRC in the course of the iSAR submittal.
The backdating of page 7a in 1976 impacts on the integrity of the individuals involved but does not materially affect the calculation results or the FSAR submittal in 1975. There was no evidence that the calculation was further modified as a result of the NRC investigation begun on April 11, 1978. This incident appears to be an isolated case since no other similar backdated documents were identified. The investigators, through interviewa, concluded that other DPC employees and management were not aware of the backdating incident.
I454 210 z