ML20238C160
| ML20238C160 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Sequoyah |
| Issue date: | 06/16/1987 |
| From: | Myers H HOUSE OF REP. |
| To: | Harold Denton NRC |
| Shared Package | |
| ML082310219 | List: |
| References | |
| FOIA-87-726 NUDOCS 8712300090 | |
| Download: ML20238C160 (4) | |
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S June 16, 1987 To Harold Denton From Henry Myers gg Our ongoing review of NRC's regulation of TVA's nuclear program has raised the following questions.
1.
What actions has TVA taken to address deficiencies specified in the March 5, 1987 letter to Steven White from James Taylor?
2.
What are the status and/or results of NRC review of the findings of the Design Baseline and Verification Project?
3.
The Division of Nuclear Engineering (DNE) Engineering Assurance (EA) Audit 87-09 concludes that the reviews of design calculations by the various DNE disciplines "when completely implemented would adequately satisfy the stated objectives" of the " calculation review effort."
Audit 87-09 also stated that within three of the four disciplines that were audited (nuclear, mechanical, and civil), a list of individual essential calculations was not available.
Since the list of essential calculations was not available to the auditors in three of the four audited disciplines, what was the auditors' basis for concluding that the design review would adequately satisfy its stated objectives?
Does the one essential electrical calculation missing from the essential list, constitute the only missing essential electrical calculation?
Was this missing calculation found by the 87-09 auditors or by EEB?
I f.
the former, why did EEB not find it?
4.
Does NRC believe that a significant number of mechanical calculations are (A) missing, (B) superceded, (C) out-of-date, or (D) based on data that was not design basis data?
If so, what has NRC and/or TVA done to assess the effect of such deficiencies cpon TVA's ability to determine whether original or post-licensing design and construction complied with NRC requirements?
5.
Audit 87-09 states:
No specific list of missing (civil) calculations was available for the audit team to review.
In conversations with CEB personnel, approximately 950 calculations are still missing, Since no specific list of essential calculations exist, no verification of the existence of all essential calculations or their retrievability could be performed.
What is the significance of this finding?
8712300090 871222 PDR FOIA WANN87-726 PDR i
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[ Note. The March 5, 1987 letter to Mr. White from 3
Mr. Taylor stated:
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Reinforced concrete allowable stresses were exceeded in the floor of the main stean valve room of the auxiliary building as a result of pipe rupture loads.
No justification for this condition was provided in the i
calculations.]
6.
What are the Sequoyah restart criteria?
Does NRC accept TVA's position with respect to criteria TVA is using to determine which issues need be resolved prior to restart and which after restart?
What are the criteria for determining whether an issue may be " resolved" by a comnitment to make design and/or hardware modifications after rather than prior to restart?
7.
EA-OR-001 states that the Engineering Assessment (EA) findings resulted in the issuance of "38 CAQs over and above those" identified by the Design Baseline & Verification Program i
(DB&VP).
How may CAQs were identified by the Design Baseline &
Verification Program (DB&VP)?
8.
What are the 38 CAQs issued by the DB&VP as a result of the EA-OR-001 review?
Why does EA-OR-001 list by identification i
number only 4 of such CAQs?
j 9.
Does EA-OR-001 and/or the Design Baseline & Verification Program (DB&VP) take adequate account of missing and/or incomplete calculations described in 87-097 l
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- 10. What is the Restart Design Basis Document?
(See EA-OR-OO1, p.
4-9. ) Has it been reviewed by the NRC7 1
- 11. Has NRC reviewed the adequacy of the EA-OR-001 procedure for designating the extent and significance of the EA-OR-001 findings?
Has NRC reviewed the adequacy of implementation of the l
j EA-OR-001 procedure for designating the extent and significance of EA-OR-001 findings?
[See EA-OR-OO1 p. 8-11 and Table 8.4-7. ]
- 12. Does EA-OR-001 contain a separate listing of Action Items indicative of generic deficiencies in the DB&VP7 Where does any such listing appear 7 Has the NRC reviewed the EA-OR-001 Action Items in order to determine the adequacy of the categorization of such items with respect to generic applicability and significance?
- 13. At the April 10 meeting between TVA and NRC staf f, Mr. John i
Cox, who was assigned overall direction of the DB&VP, stated that:
.... none of these deficiencies (found by the DB&BP), if left uncorrected, would or could have jeopardized the health and safety of the public from the operation of the
\\.
4 facility." (Tr., p.18.]
Is it TVA's position that none of the deficiencies found by the DB&BP, if lef t uncorrected, would or could have jeopardized the health and safety of the public from the operation of the facility?
If so, does NRC staff agree with this position?
Is
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such a position consistent with the findings of the Engineering Assurance Oversight Review Report (EA-OR-OO1)7 Is such a position consistent with the findings of Engineering Assurance Audit 89-07?
14.
Attached to a May 8, 1987 memorandum from R.O.
Barnett to J.A. Kirekbo is a listing of documents pertaining to Sequoyah pipe support analyses.
Documents containing many such calculations are missing and may have been destroyed.
What is NRC's understanding as to the whereabouts of a.
I these calculations?
If-calculations were destroyed, l
why'was this done?
i I
b.
Has TVA presented a listing of missing pipe support l
calculations to the NRC7 How nany of the missing pipe j
support calculations had been reviewed prior to their
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having destroyed?
]
c.
Af ter issuance of the Sequoyah Construction Permit, was was the design basis earthquake for Sequoyah changed?
If so, on what date or dates was the change made?
What design and hardware changes were made as a result of any post-CP changes in the design basis earthquake?
What reviews have been undertaken by the NRC to determine whether such design and hardware changes were in fact an adequate response to any post-CP changes in the design basis earthquake?
Where are any such reviews documented?
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