ML20205A166
| ML20205A166 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/28/1986 |
| From: | Gridley R TENNESSEE VALLEY AUTHORITY |
| To: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 8608110316 | |
| Download: ML20205A166 (80) | |
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TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 374ot SN 157B Lookout Pla'cN h [ g U.S. Nuclear Regulatory Commission JUL 28 886 A/l *. 8/
Region II ATTN: Dr. J. Nelson Grace, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
Dear Dr. Grace:
SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 - NRC-OIE REGION II INSPECTION REPORT 50-327/86-27 AND 50-328/86 RESPONSE TO DEFICIENCIES AND UNRESOLVED ITEMS Enclosed is our response to your letter to S. A. White dated April 22, 1986, which transmitted Notice of Violation Nos. 50-327/86-27 and 50-328/86-27 for our Sequoyah Nuclear Plant. Jim Domer of my staff discussed an extension of the response date until July 29, 1986 with Lee Spessard of your staff.
In addition, it should be noted that we have received your letter dated June 17, 1986 requiring information on the generic applicability of these items to the Browns Ferry, Watts Bar, and Bellefonte Nuclear Plants and TVA plans to submit a separate response addressing this issue in the near future.
If you have any questions, please get in touch with G. B. Kirk at (615) 870-6549.
To the best of my knowledge, I declare the statements contained herein are complete and true. ~
Very truly yours, TENNESSEE VALLEY AUTHORITY R. Gr dley, Dir ctor Nuclear Safety nd Licensing Enclosure cc (Enclosure):
Mr. James Taylor, Director Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mr. S. P. Weise, Chief Reactor Projects Branch 1 Division of Reactor Projects U.S. Nuclear Regulatory Commission Region II - Suite 2900 101 Marietta Street, NW Atlanta, Georgia 30323 h f An Equal Opportunity Employer b
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SEQUOYAH NUCLEAR PLANT RESPONSE TO J. NELSON GRACE'S LETTER TO S. A. WHITE DATED APRIL 22, 1986 50-327/86-27 AND 50-328/86-27 e
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i <D2.1-1 (Deficiency) Furmanite Leaking Valve Bonnet DESCRIPTION: ECN L6317 authorized the use of Furmanite to stop steam leakage of check valve 1-VLV-3-891 in one of the steam supply lines to the unit 1 turbine driven auxiliary feedwater pump. Furmanite is a commercially available material that is injected into the vicinity of the gasket space to stop excessive leakage until the valve can be repaired or replaced in the next outage. The "Furmaniting" procedure can be done without isolating the valve.
The Furmanite procedure, N-84533, requires drilling into the bonnet flange to inject Furmanite into the space between the flanges outside of the gasket and inside a dam formed by inserting wire into, or peening, the seam between the flanges. In the case of valve 1-VLV-3-891, the injection of Furmanite was accomplished by drilling radially inward from the outside of the bonnet flange to a point that is slightly outside the flange stud holes. The holes are then tapped for a 3/8-inch adapter as a guide, a 1/8-inch drill is used to drill into the bolt clearance hole to form a path for injecting Furmanite. The procedure requires drilling into points near the center of the bonnet flange, to inject Furmanite into the clearance annuli between each closure stud and its hole. Drilling into studs is avoided through the skill of the mechanics doing the work.
The safety concern is primarily for the pressure integrity of the closure after performance of the Furmaniting procedure, namely:
The valve is an ASME III class B component as classified in the work plan.
The drilling procedure involved metal removal from the bonnet flange and possibly from the closure studs. Stresses can increase due to metal removal and stress concentration factors.
If Furmaniting is effective, there could be a shift in gasket loading and resultant stresses in the components of the valve.
Neither the ECN, Unreviewed Safety Question Determination or the work plan address the need for a stress analysis or otherwise dispose of this issue.
Other ECNs authorizing Furmaniting have included a stress analysis (References 4 & 5).
i BASIS: TVA has committed to apply ASME code requirements to piping and components used in a safety-related systems (Reference 5) and ANSI-N45.2.11 (Reference 6) to the design process for plant modifications.
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' D2.1-1 (Deficiency) Furmanite Leaking Valve Bonnet (continued)
The ASME Code addresses stresses in Class 2 valves by providing rules for minimum thickness (Reference 7) including the standard design of ANSI B16.34 (Reference 8). The code does not appear to anticipate that valve closure) may be modified after installation in order to facilitate continued operation. Some guidance is provided by ASME Section XI (Reference 9) that is applicable to in-service performance of repairs and replacement of components. Article IWC-4000 of Section XI provides rules for weld repair of cavities created by removal of defects in order to restore the component to its original strength and configuration. Weld repair of the holes created by Furmaniting is impractical in most situations. However, an engineering evaluation of stress increases should be included for Furmaniting ECNs applied to components covered by the ASME Code under Sections III and XI.
REFERENCES
- 1. ECN L6317, Aur Feedwater System 03, Furmanite and Peen the Bonnet on Valve 1-VLV-3-891, 10/14/84.
- 2. Furmanite Procedure N-84533, 12/18/84.
- 3. ECN L6157, 5/21/84.
- 4. ECN L6169, 6/4/84.
- 5. TVA Design Criteria No. SQN-DC-V-3.0, 12/12/75.
- 6. ANSI N45.2.11-1974, " Quality Assurance Requirements for the Design of Nuclear Power Plants."
- 7. ASME Boiler and Pressure Vessel Code,Section III. Nuclear Power Plant Components, Subsection NC, Class 2 Components.
- 8. ANSI Standard Specification B16.34, Fittings Flanges and Valves.
- 9. ASME Boiler and Pressure Vessel Code,Section XI, Rules for In-service Inspection of Nuclear Power Plant Components.
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TVA's Response D2.1-1 (Deficiency) Furmanite Larking Valva Bonnat
- 1. CAUSE TVA did not have a procedural requirement for a documented stress analysis on valves repaired by the Furmanite process. The modification was reviewed during LECN/USQD preparation and judged to not require any revision to the valve stress analysis. However, no documentation was produced to verify this conclusion.
A contributing f actor to this condition involved weaknesses in the LECN/USQD process in that special design considerations identified in USQDs were not adequately reviewed and Unplemented by the responsible discipline organizations. This contributing factor also applies to the findings documented in deficiencies D3.1-1, 3.2-2, 4.3-1, 4.3-3, and 6.1-2.
II. EKTENT TO WHICH THIS CONDITION COULD OCCUR In general, the Furmanite process has existed in the SQN maintenance program in three phases since plant commercial operation:
Phase I - All work done on a Maintenance Request (MR)
Phase II - All work done after a USQD was prepared (with no associated LECN)
Phase III - All work is done after a USQD (supporting an LECN) is prepared and a procedure is prepared by Furmanite Corporation and filed in the workplan.
Eighty-five compo:ents covered by the QA program were identified for which the Furmanite process was implemented under either MRs (Phase I) or USQDs without an accompanying LECN (Phase II).
Four ECNs have been identified which involved the Furmanite injection process (Phase III). Two of the ECNs L6157 and L6169, had a stress analysis attached to the USQD that was prepared by the Furmanite Corporation. The other two, L6317 and L6223, did not have a stress analysis performed for the valves.
III. ACTION TO CORRECT EXISTING PROBLEM The valves modified under ECN L6223 have been removed from the plant.
The USQD prepared for ECN L6317 states that the valve qualification is not degraded by the additional weight of the Furmanite and the injection mechanism. Therefore, TVA considers this modification to be acceptable in regard to seismic qualification. A stress analysis will be performed for the valve modified by ECN L6317. If the valve can not be qualified, it will be replaced before tuc' load.
D211 SQEP onsite - July 28, 1986
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III. ACTION TO CORRECT EXISTING PROBLEM (Csatinutd)
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- Fcr tha 85 componsnes whsra ths Furmanita precess was inplem1nted undsr MRs (Phess I) or USQDe (Phase II), the following actions will
, be taken to qualify the components:
- 1. Identify those components which were modified by drilling as a result of the Furmanite process.
- 2. Perform field verification of those components identified in No. I above to determine if the component is still in place or has been replaced.
- 3. Perform an evaluation on all components that were not I replaced, to determine component failure consequences, and
- 4. For all components listed in (3) above, that are determined to be necessary for plant safe shutdown, perform a detailed stress analysis on the component to qualify the component.
Components which can not be qualified will be replaced. The actions described in steps 1 through 4 will be completed prior to restart. .
IV. ACTION TO PREVENT RECURRENCE Policy memorandum PM86-04 (DNE) issued April 25, 1986, states that engineering judgment must be accompanied by technical justification.
This policy will be reflected in future revisions of the Division of Nuclear Engineering (DNE) procedures. This is an ongoing DNE activity not associated with SQN restart.
Regarding the LECN/USQD process TVA has revised this process to require re-review of the USQD at the time of LECN closure to ensure
' that any special requirements and/or design considerations referenced in the USQD have been implemented by the responsible discipline organizations. This requirement is described in NEP-6.1, " Change Control."
In support of SQN restart, TVA is reviewing the previously issued LECNs and supporting USQDs associated with the unit 2 system boundaries involved in the prerestart plant walkdowns to ensure that any special requirements and/or design considerations documented in the USQDs have been satisfied. The review will be completed prior to restart.
Safety-related components are not repaired using the Furmanite process on a Maintenance Request (MR) at this tLne. Additionally USQDs are now only issued by the Division of Nuclear Engineering (DNE) Nuclear Engineering Branch (NEB). All safety-related components which require the Furmanite process will have the USQD issued by NEB. NEB personnel who are responsible for the issuance of USQDs will be instructed in the requirement for a stress analysis evaluation when a component is to be repaired by the Furmanite injection process. These personnel will be instructed by August 8, 1986.
D211 SQEP onsite - July 28, 1986 l
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,TVA's Response D2.1-1 (Daficiency) Furminito Locking Valva Bannst (Cantinurd)
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V. OTHER RELEVANT INFORMATION OR COMMENTS Design Study Request (DSR) No. SO47 (Final Draf t Response dated February 27, 1985) addresses a wide range of potential concerns regarding nuclear plant leak sealing, including the use of the Furmaniting process.
Results of this DSR will be incorporated into a controlled standard practice document or other specification which would be readily available at the site level. DNE anticipates incorporation of the final results of the DSR into a general construction specification by September 30, 1986.
d D211 SQEP onsite - July 28, 1986
- D2.3-1 (Deficiency) Long-Term Unincorporated ECNs DESCRIPTION; ECN L 5320, dated 10/23/80, authorizes replacement of overcurrent relays for the essential raw cooling water pump motors with larger capacity overcurrent relays. The original relays were provided for 600-hp essential raw cooling water pumps.
A new essential raw cooling water pumping station was constructed using essential raw cooling water pump motors rated at 700 hp. Preoperational testing of the new pumps using the original relays in the 6.9kV shutdown board identified that the overcurrent relays are set too low. This resulted in relay chattering and relay hang up. The relays had to be cleared manually. The Preoperational Test Deficiency Report (Reference 2) identified that if the relay hung up for an inactive pump that was needed for auto-start, the pump could not be started until the relay was manually cleared.
ECN L5320 was issued in response to Deficiency Report PT-566 and Design Change Request SQ-DCR-741. The latter is a generic DCR requesting issuance of ECNs based on Preop Test Deficiencies (pts). The generic DCR was issued to allow timely resolution of pts. The Unreviewed Safety Question Determination Form accompanying ECN 5320 indicated that no Unreviewed Safety Question was introduced by the change.
Team review of field documentation showed the following:
The Office of Engineering is carrying the status of ECN L 5320 as
" complete." This is interpreted as complete insofar as remaining work by the Office of Engineering is concerned, but not necessarily field inplementation. The Office of Engineering changes the status of ECN to closed upon written notification by the Office of Nuclear Power that the field work and testing is complete.
Work Plan WP 8937, associated with the ECN L5320 has been cancelled.
Temporary Alteration Control Form 80-734-67 was issued and approved for emergency action on 10/4/80.
BASIS: Current TVA policy in effect since July 1985 requires that a Design Change Request be issued within 60 days of a Temporary Alteration Control Form if the temporary alteration is not returned to normal. The present condition is a situation where a change has been made without the change process associated with ECN preparation and Work Plan development, including testing. ANSI N45.2.ll (Reference 4) requires a process of design analysis and verification (e.g., test) which appear to be circumvented by the modifications implemented using only a Temporary Alteration Control Form.
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- D2.3-1 (Deficiency) Long Tern Unincorporated ECNs (continued)
REFERENCES
- 1. ECN L5320. Replace Essential Raw Cooling Water Pump Overcurrent Relays, 10/23/80.
- 2. Preoperational Test Deficiency Report PT-566.
- 3. Temporary Alteration Control Form 80-734-67.
- 4. ANSI N45.2.11-1974, " Quality Assurance Requirements for the Design of Nuclear Power Plants."
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SQN RESPONSE TO D2.3-1 (DEFICIENCY) LONG TERM UNINCORPORATED ECNs AND D5.3-1 (DEFICIENCY) TEMPORARY ALTERATIONS USING TACFs I. Cause The deficiencies occurred due to the following reasons:
- a. Before July 1985, design change requests (DCRs) were not required to be issued for TACFs in effect for more than 60 days. This resulted in some TACFs remaining open without DCRs being issued.
- b. The scheduling of work onsite is done usins a priority system, and due to higher priority tasks, some TACFs ano FCNs have remained open for extended periods of time.
II. Extent to Which the Condition Could or Does Exist in Our Design Approximately 158 TACFs are still open at Sequoyah Nuclear Plant.
III. Action to Correct Existing Condition The following actions have been taken to correct the deficiencies:
- a. Once a month a computer printout is issued to site managers identifying open TACFs and the associated ECNs.
- b. Every six months the Plant Operations Review Committee reviews the status of all open TACFs and associated ECNs.
- c. A safety evaluation has been performed for all open TACFs.
IV. Action Required to Prevent Recurrence The following actions will be taken to prevent recurrence:
- a. The Division of Nuclear Engineering will perform a safety evaluation on all safety-related TACFs before implementation. The appropriate procedures to implement this review will be revised by September 1, 1986.
- b. Sequoyah has been involved in a long-term program that places and maintains management emphasis on controlling the use of temporary alterations and reducing the total number of temporary alterations in effect. For example, during the period January 1, 1985 through January 31, 1986, the number of temporary alterations existing in the plant decreased from 297 to 158. Since January 1982, the number has decreased from approximately 990 to the present level.
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Of the open temporary alterations on January 31, 1986, approximately 84 date before 1984. The plant currently has an objective to close out this particular group by the end of the next refueling outage on unit 1 (INPO commitment). The overall status of each temporary alteration is tracked and reviewed by plant management on a periodic basis to ensure progress toward this goal.
Each open or existing temporary alteration has been reviewed previously to determine that the alteration was allowable within the technical specifications and that an unreviewed safety question (USQ) does not exist. As part of the current ongoing program to establish the adequacy of the present plant configuration for restart, plant management has directed that each open temporary alteration be reassessed to ensure that plant safety is not degraded due to the existence of the alterations. This evaluation is in progress and will be completed before plant restart.
Y. Other Relevant Information or Comments None acknowledged for these deficiencies.
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l D3.1-1 (Deficiency) Exhauster Installation l DESCRIPTION: TVA ECN L5267 added quick exhausters to the controls for eight valves installed in Sequoyah Nuclear Plant (SQN). The USQD indicates that: "Masoneilan is evaluating the installed configuration of the exhauster on the control valve and will be sending documentation of the seismic qualification later (with the possible but not expected addition of new support requirements for air tubing)." However, the TVA Electrical Engineering Branch could not obtain a seismic qualification document. The exhausters may therefore have been installed without a valid USQD. TVA confirmed that a total of eight exhausters have been installed to TVA in-house alternate support spacing criteria. The team reviewed Detail J120 (Reference 1), which provides a schematic of the quick exhauster and associated 1/2-inch tubing. Tubing spans for twa of the eight installations (control valves 2-LCV-3-148 and -171) appear to exceed the maximum allowable TVA support spacing for spans with concentrated weights (Ref erence 2) . The team noted that none of the eight installed exhausters appear to have been evaluated with respect to a governing Masoneilan seismic qualification document. Moreover, two of the eight exhausters do not even appear to be installed in accordance with the TVA alternate support spacing criteria.
i BASIS: Section 4.1.2.8 of TVA Engineering Procedure EN DES-EP 4.52 (Reference 3) requires, in part, that a USQD be prepared for each L-ECN or TVA-approved design change against a nuclear power plant with an issued operating license before physical work is authorized. Section 4.5 specifies that completion of design work for an L-ECN involves, in part, a completeness review of the associated USQD to be certified by meno. TVA Engineering Procedure EN DES-EP 2.03 (Reference 4), Section 3.4, notes that
- failure to adhere to requirements specifically identified in the USQD evaluation nullifies the USQD evaluation. Finally, two exhausters were not installed in accordance with the TVA span allowables tabulated in Reference 2.
REFERENCES:
- l. TVA Drawing No. 47W600-120, Mechanical Instruments and Controls, Revision ll, dated April 22, 1980.
- 2. TVA Table No. 47A053-15B, Mechanical Seismic Support Process Pipe 2-Inch Diameter & Less, Revision 0, dated September 15, 1977
- 3. TVA Engineering Procedure EN DES-EP 4.52, Engineering Change Notices (ECNs) Af ter Licensing-Handling, Revision 1, dated April 24, 1984.
- 4. TVA Engineering Procedure EN DES-EP 2.03, Unreviewed Safety Question Determination (USQD) - Handling and Preparation, Revision 6, dated April 24,1984.
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- TVA's Response D3.1-1 (Deficiency) Exhauster Installation I. g6 Egg Seismic qualification documentation for ECN L5267 was deficient because of inadequate design change control. The specific quick exhauster modifications were not analyzed and approved by engineers responsible for component seismic qualification and thus the appropriate documentation was not obtained.
In addition, as previously discussed in the response to. deficiency
. 2.1-1, a contributing factor to this condition involved weaknesses in the LECN/USQD review process.
II. EETENT TO WHICH THE CONDITION COULD OR DOES EKIST l The specific condition existed for quick exhauster installations on ECN L5267. The generic condition could exist for any ECN involving component seismic qualification if adequate design change control was not exercised. The exact extent of the generic condition has not been established, but approximately 300 ECNs involving component seismic qualification have been issued since fuel load.
III. ACTION TO CORRECT EXISTING CONDITION Seismic adequacy of the existing quick exhauster installations has been adequately documented. No hardware deficiencies were found by seismic analysis.
A representative sample of 60 ECNs involving electrical and mechanical component seismic qualification af ter fuel load has been selected and evaluated. The representative sample was evaluated for seismic adequacy of the existing installations. No seismically inadequate installations were found. Thus a 95-percent assurance of
! seismically adequate installations has been proven.
In addition, critical safety system ECNs (issued since SQN operating license was obtained) are being evaluated as part of the Design Baseline Verification Program. This work is controlled by Sequoyah Engineering Procedures SQEP-ll and SQEP-12 and it will be completed before restart.
Regarding actions to correct the LECN/USQD process, see TVA's response to deficiency 2.1-1.
IV. ACTION REQUIRED TO PREVENT RECURRENCE ( ARPR)
The applicable design interf ace control document is CEB-DI-121.03 entitled " Seismic Design, Review, and Control." CEB-DI-121.03 has been revised to clarify situations requiring seismic qualification review / approval and to ensure that qualified engineers in the civil engineering discipline perform this work.
FP00;D311 SQEP onsite l
'TVA's Response D3.1-1 (Daficiency) Exhnuster Installction (Continutd)
IV. ACTION REQUIRED TO PREVENT RECURRENCE (Continued)
The Sequoyah Engineering Project Manual has been updated to ensure implementation of CEB-DI-121.03. Training classes have been held at SQN on this topic.
The ARPR for SCR SQNEEB8521 ensures that current design changes to vendor-supplied equipment are properly documented including seismic qualification considerations. This is accomplished by producing drawings of subcomponent modifications of vendor-supplied equipment. Those drawings are then reviewed / approved for seismic design adequacy.
Regarding actions associated with the LECN/USQD process, see TVA's response to deficiency 2.1-1.
V. OTHER RELEVANT INFORMATION OR COMMENTS The representative sample, described above, included Gilbert /
Commonwealth audit issues 4, 9,10,14, and 19 and seven ECNs identified from unresolved item U6.3-2.
The seismic adequacy of some solenoid valve installations in control air systems is the subject. of SCR SQNCEB8629. Those installations have electrical condulet attachments which caused seismic adequacy of the installations to be questionable. Seismic qualification tests have been conducted for the questionable installations. Corrective action for this SCR includes modification of the installations which did not pass the qualification tests. This will be complete prior to restart of the units.
The tubing overspan concern described in this deficiency is .
considered an isolated instance for control air systems installed by alternate seismic support spacing criteria. The probable cause for this instance is that the quick exhauster valves (weight = 0.6 pounds) were considered insignificant concentrated j
weights. This apparent judgement led to a documentation deficiency, but not a hardware deficiency. Therefore, there is no basis for a general concern regarding overspans in alternate supported control air tubing. The corrective actions described in this response will provide adequate assurance of the seismic adequacy of such installations.
i FP00;D311 SQEP onsite
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DS.2-2 (Deficiency) USQD Requirement
- DESCRIPTION: TVA ECN L5500, in part, adds extension operators and covers to units 1 and 2 valves 67-507A installed in the essential raw cooling water system. The ECN calls for a seismic analysis to seismically re-qualify the existing valves. The USQD for the ECN specifies that: "a seismic analysis will be done to show that the new valve stem extensions do not invalidate the existing seismic qualifications for the valve, piping or associated components". However, the TVA Civil Engineering Branch was not able to obtain new seismic qualification dccumentation for the valves.
BASIS: Section 4.1.2.8 of TVA engineering procedure EN DES-EP 4.52 (Reference 1) requires, in part, that a USQD be prepared for L-ECN or TVA-approved design change against a nuclear power plant with an issued operating license before physical work is authorized. Section 4.5 specifies that completion of design work for an L-ECN involves, in part, a completeness review of'the associated USQD to be certified by memo. TVA engineering procedure EN DES-EP 2.03 (Reference 2), Section 3.4, notes that failure to adhere to requirements specifically identified in the USQD evaluation nullifies the USQD evaluation.
REFERENCES
- 1. TVA Engineering Procedure EN DES-EP 4.52, Engineering Change Notices (ECNs) After Licensing-Handling, Revision 1, dated April 24, 1984.
- 2. TVA Engineering Procedure EN DES-EP 2.03, Unreviewed Safety Question Determination-Handling and Preparation, Revision 6, dated April 24, 1984.
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. o D3.2-2 (Deficiency) USQD Requirement TVA's Response D3.2-2 I. CAUSE Seismic qualification documentation for ECN L5500 was deficient because of inadequate design change control. The specific valve extension operator modifications were not analyzed and approved by engineers responsible for component seismic qualification, and thus the appropriate documentation was not obtained.
In addition, as previously discussed in the response to deficiency 2.1-1, a contributing factor to this condition involved weaknesses in the LECN/USQD review process.
II. EKTENT TO WHICH THIS CONDITION COULD OR DOES EXIST No seismic qualification documentation has been located for any of the remote valve stem operators at Sequoyah. This condition also affects the seismic qualification of the approximately 300 valves to which the remote operators are attached. Pipe stresses and support loads in the vicinity of the valve operators are also affected.
The generic concern for inadequate design change control for modifications involving component seismic qualification is addressed in TVA's response to D3.1-1.
III. ACTION TO CORRECT EXISTING PROBLEM SCR SQNCEB8621 has been written to address the lack of seismic qualification documentation for valve extension operators at Sequoyah. The existing installations will be evaluated to ensure that safety functions are not compromised during a seismic event.
Adequate seismic qualification documentation will be provided for the remote valve stem operators, attached valves, and rigorously analyzed piping and supports prior to restart of the units. Alternately-supported piping will be evaluated for the additional concentrated weight effects after restart. See TVA's response to D.3.3-4.
l Regarding action to correct the LECN/USQD process, see TVA's response l to deficiency D2.1-1.
t l IV. ACTION TO PREVENT RECURRENCE l
l The seismic design interface document CEB-DI-121.03 has been revised I
to clarify situations requiring seismic analysis and approval and to ensure that qualified engineers perform this work. In addition, the Sequoyah Engineering Project Manual has been updated to implement I this revised document and associated training classes were conducted
- at SQEP.
FP00 ;D322 SQEP onsite 4
. . D3.2-2 (Deficiency) USQD Requirement (Continued)
IV. ACTION TO PREVENT RECURRENCE (Continued)
The Sequoyah Rigorous Piping Analysis Handbook has been revised to include remote valve actuator modeling requirements.
Regarding actions associated with the LECN/USQD process, see TVA's d
response to deficiency D2.1-1.
V. OTHER RELEVANT INFORMATION OR COMMENTS The corrective action for SCR SQNMEB8512 has already ensured that adequate relative motion capacity is provided for all Sequoyah remote valve operation installations. Since relative movement capability (particularly thermal movement) is the more significant safety :
concern, it is probable that the existing installations are seismically adequate and SCR SQNCEB8621 is expected to be a documentation problem only.
The more significant concerns identified by SQNCEB8621 will be fully addressed before restart as described above.
The significance of the concern about alternately-supported pipe stress and support loads is low due to the inherent conservatism in the alternate piping support criteria for seismic loads. Therefore, that concern will be addressed af ter restart as part of the Phase 2 alternate analysis program (see TVA's response to D3.3-4).
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D3.2-3 (Deficiency) Piping Flow Diagram DESCRIPTION: TVA ECN L5737 added a check valve to the primary water piping to the CVCS spent resin header. The primary water piping is TVA Class G while the CVCS piping is pressure boundary piping qualified to TVA Class D.
The check valve was added to the flow diagram (reference 1) but the change of Class from D to G is not shown on the revised flow diagram.
BASIS: TVA Design Criteria No. SQN-DC-V-3.0 (Reference 2) Section 3.5, requires, in part, that piping drawings be clearly marked to indicate the TVA pipe classification of all piping reresented, and that all interface boundaries of higher and lower class piping be pinpointed exactly.
i REFERENCES
- 2. TVA Design Criteria No. SQN-DC-Y-3.0, The Classification of Piping, Pumps Yalves and Vessels, Revision 1, dated June 28, 1985.
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k j FP00 ;D323 SQEP onsite
t TVA Resnonse D3.2-3 Piping Flow Disgram I .' CAUSE The cause can be attributed to unclear piping breaks shown on flow and piping drawings on the CVCS system that were in existence prior to issuance of ECN L5737. The checker accepted as correct the class -
boundary break shown as Class G on the flow diagram for the CVCS pipe
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at the point at which the Class G primary waterline ties in. The .
CVCS piping at this point should be shown as Class D as per the piping drawings, bills of material, and the seismic analysis on .
record. , ,
II. EKTENT TO WHICH THE CONDITION COULD OR DOES EXIST A review of the CVCS flow diagrams and the physical piping drawings has identified several other places on the affected drawings where class breaks are missing, incorrect, or ambiguous.
III. ACTION TO CORRECT EXISTING CONDITION SCR SQNMEB8614 RO was initiated to address this deficiency and has -
been revised to address the generic problems of class breaks on all ~
mechanical systems.
IV. ACTION TO PREVENT RECURRENCE All mechanical flow diagrams and drawings will be reviewed and revised as required to clearly indicate where all class breaks occur.
This will be completed prior to unit 2 restart.
- V. OTHER RELEVANT INFORMATION OR COMMENTS The deficiency as described in the audit report was a documentation only deficiency. Although the flow diagram indicated that the CVCS piping was Class G in the area of the tie-in (ECN L5737), the piping drawings show all Class D materials being used, and the piping analysis was done to Class D requirements. Other class break -
discrepancies identified in revision 1 of SCR SQNMEB8614 will be evaluated to determine if any piping has been installed to a piping class which is lower than required. This review will be completed <
prior to unit 2 restart. -
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FP00;D323 SQEP onsite
D3.2-4 (Deficiency) Sample Connection Support DESCRIPTION: TVA ECN L6462 authorized the installation of a support for a 3/4 inch sample connection coming off the downstream piping of the CVCS
- - - letdown heat exchanger. The support, designed to seismic Category I criteria, is being added to reduce vibration in the 3/4 inch sampling line
- connected to the 3 inch process line. The support was designed to a minimum frequency of 50 hertz. The details for the clamped connection of the support to the 3 inch process line are derived from a standard support detail sheet (Reference 1) and are attached to the ECN. The support detail is based on TVA Civil Engineering Branch Report No. 77-42 (Reference 2),
which originally specified two heavy Bergen-Patterson clamps with four 7/8 inch diameter bolts torqued to 100 ft-lbs. However, the standard support detail now specifies Basic Engineers heavy duty pipe clamp BE-122, which uses 3/4 inch bolts instead of 7/8 inch bolts for the 3 inch clamps. In addition, split washers or lock nuts were not specified for the installation in order to maintain the bolt design torque under vibratory loads. The team
_ notes that the support detail specified in the TVA Civil Engineering Branch report is based on static and not dynamic loads, s
BASIS: The basis for this deficiency is the use of smaller diameter bolts than called for in the standard support detail. In addition, the 100 ft-lbs originally specified for the 7/8 inch bolts and specified in the ECN to be used for the 3/4 inch bolts may overstress the 3/4 inch bolts.
REFERENCES
- 1. TVA SQN Drawing No. 47A406-2-4, Mechanical-Unit 2/ Category Support for Support Detail 2-4, Revision 1, dated August 2, 1985.
- 2. TVA Civil Engineering Branch Report No. 77-42, Static Pipe Support
' Tests and Development-Sequoyah Nuclear Plants 1 & 2, dated October 25, 1977.
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D3*2-4 (Deficiency) Sample Csnnsctica Suppsrt TVAs Response D3.2-4 I. SAMEE Refer to Section V.
II. EXTENT TO WHICH THIS CONDITION COULD OR DOES EXIST Refer to Section V.
III. ACTION TO CORRECT EXISTING PROBLEM Refer to Section V.
IV. ACTION TO PREVENT RECURRENCE Refer to Section V.
V. OTHER RELEVANT INFORMATION OR COMMpTS CEB Report 77-42 was written to document static pipe support tests on 4" main run pipe and heavy clamp Bergen-Patterson (BP) 298. The report is a reference document and not intended to be a basis for support details. The bolt designations contained in the report are for 4" main run pipe only. Bolt sizes for 3" main run pipe must be taken from the BP catalog. ECN L6462 designates 3/4" bolts for the 3" main run pipe support which is in agreement with the BP catalog.
ECN L6462 authorized installation of a support using ASTM A-193 or equal high-strength clamp bolts. A torque of 100 f t-lbs will not overstress the high-strength bolts. Also, it is TVA's position that split-lock washers or locknuts are not necessary because the 100 f t-lbs torque will maintain bolt preload under vibratory conditions.
The support detail is based on static and not dynamic loads because the support minimum natural frequency of 50 hertz is greater than the l
25-hertz minimum natural frequency requirement for seismic rigid response. The 50-hertz requirement is for vibration restraint purposes.
The original support detail specified heavy clamps--B-P type 298.
ECN L6462 authorized the installation of Basic Engineer type 122 heavy-duty pipe clamp or equal. Based on a review of catalog data,
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the clamp dimensions are equivalent and the load ratings are within 1 percent of each other. Thus, the clampo are equivalent and the support detail is acceptable.
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- D3.3-1 (Deficiency) Pipe Support Friction Design DESCRIPTION: Section 7.19 of the TVA pipe support design manual (Reference
- 1) indicates that friction forces are not considered for pipe support design at Sequoyah Nuclear Plant. The team noted that forces due to friction for thermal displacements greater than 1/16 inch are generally taken into account when computing pipe support reactions due to dead and thermal loads. Moreover, USAS B31.1-1967, the piping code of record for TVA safety class B, C and D piping at Sequoyah Nuclear Plant, requires consideration of frictional forces due to piping thermal expansion.
A pipe support which cannot accommodate piping thermal movement without resistance is subject to friction force in addition to bearing force. The magnitude of the applied friction force is about one-third of the piping dead load and operating thermal force bearing on the support. The direction of the applied friction force is the same as the direction of the piping thermal movement and is perpendicular to the direction of the pipe support bearing force. Friction force is potentially significant for a pipe support which is weaker in resisting friction force than in resisting bearing force. In order to qualify a pipe support subject to friction and bearing force by analysis, the stresses due to friction and bearing force must be separately computed, and combined in accordance with AISC code requirements.
BASIS: The basis for this deficiency is TVA's failure to analyze pipe supports for friction forces due to thermal displacements, as required by piping code B31.1. Section 120.2.3, Anchors or Guides, requires that:
"where anchors or guides are provided to restrain, direct, or absorb piping movements, their design shall take into account the forces and moments at I
these elements caused by internal pressure and thermal expansion". Section
! 121.2.1, Anchors and Guides, paragraph (e), specifies that: " Brackets shall l
be designed to withstand forces and moments induced by sliding friction in I
addition to other loads."
l REFERENCES
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. TFA's ResDonse D3.3-1 Pipa Support Friction
'I . EAEEE It was TVA's opinion during the original design of SQN pipe supports that the effect of friction loads due to temperature on pipe supports was insignificant and thus could be neglected. Thus, TVA's SQN pipe support designs did not consider friction loads due to thermal expansion.
II. EKTENT TO WHICH THIS CONDITION COULD OCCUR The effects of friction loads due to temperature are not considered in any of TVA's pipe support designs at SQN.
III. ACTION TO CORRECT EKISTING PROBLEM TVA has conducted a study for WBN unit 2 pipe support designs which demonstrates that friction loads do not generally govern the designs.
Out of a total of approximately 2,000 supports, 4 supports were identified as being governed by friction loads. This is considered to be an acceptable percentage for neglecting friction and provides adequate assurance that friction loads do not significantly affect pipe support designs. Since the WBN support designs are similar to the SQN designs, it is f elt the WBN study is valid to demonstrate the adequacy of the SQN supports to neglect friction. SQN has been an operating plant, and supports have not shown excessive stresses due to friction created by thermal gradient. For this reason and because of the results of the WBN study, this is not a restart item.
However, TVA is going to conduct an evaluation of the SQN support designs for the effects of friction loads as discussed in Section V below. Evaluation will start on October 1,1986, and will be completed on November 28, 1986.
IV. ACTION TO PREVENT RECURRENCE TVA has issued a design criterion for pipe supports that will require the consideration of friction loads due to temperature for the design of all new pipe supports and for consideration in the design of supports that are being modified due to load changes, changes in configuration, etc.
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TVA's Response D3.3-1 Pipe Support Friction (Continued)
V. OTHER RELEVANT INFORMATION OR COMMENTS The following areas represent the major systems which are subjected to thermal loads and movements:
- 2. Essential Raw Cooling Water
- 3. Chemical and Volume Control
- 4. Component Cooling Water
- 5. Mainsteam and Feedwater
- 6. Reactor Heat Removal
- 7. Fire Protection
- 8. Spent Fuel Cooling Water
- 9. Raw Cooling Water
- 10. Extraction Steam
- 11. Safety Injection
- 13. Makeup and Purification An in-depth worst-case biased evaluation will be made on each system.
The isometric drawings will provide a reference from which the supports may be selected.
The supports selected will be evaluated in two ways as follows:
- 1. Field investigations will be performed by site QA and design representatives during walkdowns.
- 2. An evaluation of the supports will be performed which includes the forces caused by friction to ensure that the weld stresses, plate bending stresses, and anchorage pullout loads are within the design basis.
The above information, in conjunction with the information provided from WBN friction investigation, will be used to substantiate that SQN pipe supports are adequate as designed.
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- D3.3-2 (Deficiency) Valve Accelerations l l
DESCRIPTION: FSAR Section 3.9.2.5.2 stipulates that: " seismic valve !
l accelerations are maintained below 2g vertical, and 3g horizontal for the SSE condition". However, the teams found that valves are being qualified on a case by case basis is acceleration levels which exceed the FSAR commitment. As an example, the team reviewed piping analysis N2-78-6A (Reference 1), which reanalyzes the 4 inch spent fuel pit cooling line from containment penetration X-83 to a three-way rigid restraint. The system was analyzed for static, normal thermal, design basis accident thermal and pressure, design basis accident inertia, operating basis earthquake seismic, and safe shutdown earthquake seismic conditions. The piping subsystem contains one 4 inch diaphragm valve and two 3/4 inch globe valves (Reference 2). The 4 inch Grinnell valve was qualified to computed accelerations of 2.66g vertical and 7.67g horizontal; the 3/4 inch Hancock valve was qualified to computed accelerations of 1.19g vertical and 19.58g horizontal, and the 3/4 inch Kerotest valve was qualified to computed accelerations of 2.92g vertical and 7.33g horizontal. TVA technical personnel have indicated that Appenoix F of the Sequoyah Nuclear Plant quality assurance manual (Reference 3) formed the technical basis for the procurement of valves prior to 1975. Section 6.1.1 of that document requires qualification of valves and other components supported by piping systems to 3.0g horizontal acceleration and 2.0g vertical acceleration in accordance with the FSAR ccamitment. FSAR Table 3.9.2-3 also notes that: " pumps and valves are supported to assure each component is not seismically loaded in excess of the "g" loading specified in the design specification".
BASIS: Some valves are being qualified to computed acceleration levels which exceed the FSAR commitments and the valve procurement qualification criteria.
REFERENCES
- 2. TVA SQN Drawing No. 47K454-58, Reactor Building Unit 1/ System i N2-78-6A/ Isometric of Static, Thermal, and Dynamic Analysis of Spent l Fuel Pit Cooling Piping, Revision 1, dated June 19, 1980.
- 3. TVA Appendix F of SQN Quality Assurance Manual, Design Criteria for Qualification of Seismic Class I and Seismic Class II Mechanical and Electrical Equipment, Revision 1, dated February 10, 1972.
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TVA's Response D3.3-2 (Daficiency) Vcive Accelcrations I. ' CAUSE When a calculated valve acceleration exceeds the 3g(H) and 2g(V) lLnits to which the valve was originally qualified, an analysis or test is performed and documented to justify the increased acceleration allowable or the support system is modified to limit accelerations to allowable values. This process is controlled by the piping analysis handbook, but it is not reflected in the FSAR.
This condition can be attributed to a failure to update the FSAR, in accordance with applicable procedures, to reflect current programs.
This finding also applies to deficiencies D3.3-3 and D3.3-5.
II. EXTENT TO WHICH THIS CONDITION COULD OR DOES EXIST There are numerous cases where increased allowable accelerations have been approved by component seismic qualification engineers on a case-by-case basis. This is not a technical concern because sound engineering principles have been used in establishing these increased allowables.
III. ACTION TO CORRECT EXISTING PROBLEM The FSAR has been revised, in Amendment 3, to state that accelerations higher than specified in the design specification are approved by TVA on a case-by-case basis, as limited by the seismic capacity of the valves.
IV. ACTION TO PREVENT RECURRENCE Additional procedural training (including preparation of FSAR material) has been provided to DNE personnel as part of the 1985 Office of Engineering Procedures (OEP) training program (subsequently revised as Nuclear Engineering Procedures). Also, the DNE Division of Responsibility (DOR) for FSAR Sections has been revised and reissued to update organizational responsibilities for FSAR input.
Also, TVA is proceeding with a program to establish upper tier design
! criteria documents that reflect the applicable plant design bases and l
commitments (including those in the FSAR). Any discrepancies identified that require a revision to the FSAR will be documented and dispositioned accordingly. The design criteria documents associated with the unit 2 system walkdown effort will be completed prior to startup. FSAR revisions resulting from this effort will be submitted at the next appropriate 10CFR50.71(e) update.
1 I
In addition, Significant Condition Report (SCR) SCRGENNEB8602 has been issued to document a generic question regarding the control of FSAR information. Further actions associated with this SCR are not related to SQN restart in that the Design Basis Verification effort described above provides adequate assurance that the plant design basis described in the FSAR is consistent with TVA design input documents.
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, , TVA's Response D3.3-2 (Deficiency) Valve Accelerations (Continued)
Y. OTHER RELEVANT INFORMATION OR COMMENTS Documented analyses exist which justify approval of the increased valve accelerations.
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1 D3.3-3 (Deficiency) Valve Fundamental Frequency DESCRIPTION: FSAR Tables 3.9.2-1 and 3.9.2-3 indicate a minimum fundamental frequency for pumps and valves of 33 hertz. However, Section 3.1.3 of the specification used to procure mechanical and electrical equipment for Sequoyah Nuclear Plant prior to 1975 (Reference 1) notes that equipment exhibiting a fundamental frequency of 25 hertz can be considered rigid, and can be qualified by static rather than dynamic analysis. This is a less conservative requirement than the FSAR commitment.
BASIS: The technical specification used to procure pumps and valves for Sequoyah Nuclear Plant prior to 1975 specifies a minimum fundamental frequency which is unconservative with respect to the FSAR commitment.
REFERENCES
- 1. TVA Appendix F of the Sequoyah Nuclear Plant Quality Assurance Manual, Design Criteria for Qualification of Seismic Class I and Seismic Class II Mechanical and Electrical Equipment, Revision 1, dated February 10, 1972.
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c , TVA's Response D3.3-3 (Deficiency) Valve Fundamental Frequency I. CAUSE The FSAR commitment was intended to signify that valves were purchased to be seismically rigid. The definition of seismic rigidity was changed from 25 to 33 hertz in TVA valve procurement specifications in 1972. This process was not explained in the FSAR.
The cause of this condition is discussed in the response to deficiency 3.3-2.
II. EXTENT TO WHICH THIS CONDITION COULD OR DOES EXIST The exact extent of this condition is not known at this time.
Theoretically, all valves purchased prior to 1972 could have fundamental frequencies as low as 25 hertz. However, in reality this would be a rare occurrence since most basic valve designs, without seismic considerations, ensure a fundamental frequency greater than 33 hertz.
III. ACTION TO CORRECT EXISTING PROBLEM FSAR Tables 3.9 2-1 and 3.9.2-3 were revised in Amendment 3 to indicate a minimum fundamental frequency for valves of 25 hertz.
Subsequent to this revision, however, a question was raised concerning the applicability of the revised valve frequency to other components. TVA will submit a letter of clarification to NRC on this matter by September 2,1986. See TVA's response to D3.3-5.
IV. ACTION TO PREVENT RECURRENCE Regarding actions to address inconsistencies in the FSAR against design inputs, see TVA's response to deficiency D3.3-2.
V. OTHER RELEVANT INFORMATION OR COMMENTS
(
Although the pre-1972 frequency requirement of 25 hertz was inconsistent with the FSAR value of 33 hertz, this is not considered to be a significant technical concern. An informal review of the seismic response spectra for Sequoyah has shown that the difference in response at 25 and 33 hertz is not significant. In addition.
TVA's other conservative analytical assumptions including low damping and nozzle load allowable have ensured a conservative design interf ace between fluid system equipment / components and attached piping.
The seismic design criteria for Sequoyah equipment and components will be updated as a part of the Design Baseline Verification Program. This action will also help to minimize recurrence of concerns such as this one.
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,, .D3.3-4 (Deficiency) Alternate Pipe Support Criteria DESCRIPTION: FSAR Sections 3.9.2.5 and 3.9.2.6 define the extent of piping to be explicitly analyzed and piping which will be supported to generic qualification criteria for Sequoyah Nuclear Plent. TVA Cless B, C, D, G. K and M process piping and instrument lines that do not require complete analysis may be field routed. Piping 3/8 inch through 1 inch and subject to a maximum temperature of 650 degrees F may be field routed. Piping 1-1/4 inch through 4 inch and subject to a maximum temperature of 200 degrees F may also be field routed. Mechanical seismic supports for process pipe 2 inch and less are detailed on a series of standard detail sheets (Reference 1). General definitions, requirements and guidelines are tabulated on sheet 1A. Note 11 indicates that: "The above guidelines do not consider thermal expansion or anchor movements". In light of this note, the team asked for documentation to confirm that field routed pipe subject to thermal loads was evaluated to calculate.the additional support reactions due to thermal loads. The TVA Civil Engineering Branch was not able to obtain documentation which could confirm the systematic evaluation of thermal loads on supports.
Moreover, TVA Civil Engineering Branch produced two nonconformance reports written in 1982 which indicated in-house concern with the field routing program at Sequoyah Nuclear Plant. Nonconformance report SQNSWP8215, dated September 21, 1982, notes, under Description of Condition, that:
A joint Civil Engineering Branch-SWP review on SQN alternate analysis has shown generic technical and documentational deficiencies on the analysis criteria used (Civil Engineering Branch-74-2, Civil Engineering Branch-80-5, Civil Engineering Branch-76-5, and Civil Engineering Branch-75-9) and their application on SQN. These deficiencies are in the following areas:
- 1. Incomplete documentation of analysis assumptions and criteria exceptions.
- 2. Improper documentation of the report criteria.
- 3. Inadequate compliance with applicable analysis criteria during implementation of alternate piping analysis; particularly for determination of support locations and design loads.
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-* 03.3-4 (Deficiency) Alternate Pipe Support Criteria (continued)
Nonconformance report SQNSWP8222, dated December 21, 1982, notes, under Description of Condition, that:
EN DES has failed to develop a procedurally controlled system to ensure that all piping, defined in the Sequoyah Design Criteria SQN-DC-V-13.7
" Alternate Piping Analysis and Support Criteria for Category I Piping Systems" (last revision 11/8/74), has been supported according to the appropriately specified criteria. As a result, there exists the possibility that piping segments may not have been analyzed nor their dependent pipe supports released to CONST.
TVA corrective action to address these nonconformance reports should have been prompt and included systematic identification and evaluation of pipe supports for field routed piping subject to thermal loads.
TVA Civil Engineering Branch indicate that efforts are currently underway to address these nonconformance reports.
BASIS: The basis for this deficiency is TVA's failure to systematically address pipe support thermal loads for field routed pipe. TVA's failure to address nonconformance reports SQNSWP8215 and -8222 in timely fashion represents inadequate corrective action. Criterion XVI, Corrective Action, of 10CFR50, Appendix B requires in part that measares be established to assure that conditions adverse to quality such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
REFERENCES
- 1. TVA Mechanical Seismic Support, Process Pipe 2" Diameter and Less, Sequoyah Nuclear Plant, Drawing Series 47A053. Rev. 5. February 14, 1981.
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O e TVA's Resoonse D3.3-4 (Deficiency) Alternate Pipe Support Criteria !
I. S&ggg The lack of documentation of the thermal evaluation of field-routed pipe is recognized. The condition has been assessed and is primarily attributed to a lack of clear assignment of responsibility regarding application of alternate analysis criteria, resulting in inadequate implementation and in vague instructions being provided to the field.
NCRs SQNSWP8215 and SQNSWP8222 addressed this and other potential discrepancies regarding alternate analysis. Based on the review
- of the condition as documented in NCR SQNSWP8215, and in i
J. B. Thomison's memorandum to R. O. Barnett and J. C. Standif er dated October 5,1982, the condition was judged to be one that was not an immediate safety concern. Also, the condition as discussed in NCR SQNSWP8222 was judged to be a documentation matter only and not an immediate safety concern. Although none of the discrepancies
, were thought to represent an immediate safety concern, several were recommended for additional-study and review. Because there was no immediate safety concern identified, the review was not expedited.
An alternate analysis review program is presently in progress.
II. EXTENT TO WHICH THIS CONDITION COULD OR DOES EXIST Some Category I (TVA Classes B, C, D) and most Category I(L) type (TVA Classes C, K, and M) process piping (2 inch and less in size) and instrument lines are generally field-routed. The TVA design
- requirements during plant construction required field isometrics to j be submitted to engineering for thermal evaluation on a case-by-case l basis. Discussions with cognizant personnel had indicated that for the majority of cases the criteria was complied with, but documentation of this activity is not always adequate to confirm Laplementation.
1 It is important to note that certain lines are generally not candidates for alternate analysis or field-routing as listed below:
- 1. TVA Class A lines (primarily reactor coolant pressure boundary l piping).
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- 2. High energy piping larger than 1-inch diameter unless it can be
! determined that there is not a potential for unacceptable pipe rupture interactions.
- 3. TVA Class B and C lines 6-inch diameter and larger (except
! alternate analysis may be used on small sections of low energy 7
lines 6 inch and larger).
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- 4. Piping systems requiring analysis consideration of design basis i
accident loads.
i l This excludes many of the high temperature lines from consideration for alternate analysis or field-routing.
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. , TVA's Response D3.3-4 (Deficiency) Alternate Pipe Support Criteria (Continued)
III. ACTION TO CORRECT EXISTING PROBLEM An alterrate analysis review program described in revised SQN Nuclear Performance Plan Volume 2 has been established and is being Laplemented on alternately analyzed TVA Class B, C, D, and some Class M piping systems and supports. The program being implemented does specifically address thermal expansion flexibility.
Category I(L)-type piping which performs a pressure retention or position retention secondary safety function (including TVA Classes G K, and some Class M) is being evaluated under another program to assess the seismic integrity of piping system per discussion in NUREG 1061.
IV. ACTION TO PREVENT RECURRENCE To ensure appropriate analysis and future compliance, a clarification of the lines of responsibility among Engineering, Construction, and Operations is addressed in Volume I of the Corporate Performance Plan. Engineering requirements for alternately analyzed field-routed pipe and for the design and evaluation of their respective supports will be specified on typical drawings and/or applicable procedures.
In conjunction with the alternate analysis review program, boundaries of rigorously and alternately analyzed piping will be specified on the physical drawings.
V. OTHER RELEVANT INFORMATION OR COMMENTS Thermal loading is a secondary self-relieving loading condition.
Because many of the high energy lines are normally excluded from alternate analysis, it follows that many high temperature lines are also excluded. The possibility of such f ailures occurring does not represent a significant risk to plant safety and is not considered to be a short-term safety concern. Consequently, resolution of this issue is not a restart item.
The timeliness of corrective action resulting from identified concerns and deficiencies is the focus of other organizational programmatic changes being planned and implemented at SQN.
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D3.3-5 (Deficiency) Pump Fundamental Frequency DESCRIPTION: The team reviewed piping analysis N2-3-3A, -4A, -5A (Reference
- 1) for the auxiliary feedwater system from penetrations X-40A and X-40B to the discharge of auxiliary feedwater turbine pump 1A-A and motor pumps lA-A and IB-B. The Upset nozzle forces due to pressure, dead weight, thermal and operating basis earthquake loads for the discharge nozzle of turbine driven auxiliary feedwater pump 1A-S are tabulated on page 25L of the analysis along with the maximum allowable nozzle forces and moments provided by Ingersoll-Rand, the pump vendor, on page 25C of the analysis. As noted on page 1058 of the calculation, the nozzle loads for the turbine driven auxiliary feedwater puup were qualified by similarity to the Watts Bar pump based on a review of its stress report (Reference 2).
The free vibration analysis performed for the pump in the stress report yields fundamental frequencies of 13 and 16 hertz in the lateral directions i and 20 hertz in the vertical direction. Sequoyah Nuclear Plant FSAR Tables 3.9.2-1 and 3.9.2-3 commit to pump fundamental frequencies equal to or greater than 33 hertz. The team reviewed the procurement specification for the turbine driven and motor driven pumps (Reference 3). Section 5, Seismic Requirements, of the procurement document does not specify a minimum fundamental pump frequency. The pump nozzles were modeled as rigid anchors in the piping analysis. There appear to be no requirements to model j flexible equipment in piping analysis.
i The team is concerned that piping subsystems which contain flexible equiptnent could be subjected to amplified pipe stresses, pipe support reactions and equipment nozzle loads during a seismic event. The team does not consider piping subsystems at Sequoyah Nuclear Plant which contain flexible equipment to be adequately qualified by analysis.
BASIS: The procurement document did not specify a minimum pump fundamental frequency of 33 hertz in accordance with the FSAR commitment. As a consequence, piping and supports downstream of the turbine driven auxiliary feedwater pumps are subject to possible amplification of pipe stresses and support reactions during a seismic event.
REFERENCES l 1. TVA SQN Piping Analysis N2-3-3A, 4A, SA, Revision 5, dated September ,
27, 1985.
- 2. Mcdonald Engineering Report No. ME-161, Seismic-Stress Analysis of Auxiliary Feedwater Pumps /TVA Specification 1547/ Watts Bar Nuclear
- Plant Units 1 and 2/ Manufactured by Ingersoll-Rand Company, dated i October, 1974.
- 3. TVA Specification 9955 for Steam-Turbine-Driven and
. Electric-Motor-Driven Auxiliary Feed Pumping Units for Sequoyah Nuclear j Plant Units 1 and 2, Contract Date July 9, 1971.
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, , TVA's Response D3.3-5 (Deficiency) Pump Fundamental Frequency I. Q& gig The fundamental frequency of pump casings is normally greater than
, 33 hertz because of pump operational requirements. This was the basis of the FSAR Amendment 2 statement. The design criteria for Sequoyah pumps and floor-mounted equipment did not have a fundamental frequency requirement. The cause of this condition is discussed in the response to deficiency D3.3-2.
II. EXTENT TO WHICH THIS CONDITION COULD OR DOES EXIST The extent of this condition has not been established at this time.
Because there was no criteria requirement for pump fundamental frequency greater than 33 hertz, the condition could theoretically exist for any floor-mounted pump. However, as stated above, pump casings are typically greater than 33 hertz. This is the case for the specific auxiliary feedwater pumps addressed by this deficiency.
The AFW pumps are skid-mounted along with the drive motor or turbine.
The calculated lower fundamental frequencies (13 hz and 16 hz) were based upon a conservative analysis of the skid base plate flexibility. Thus the deficient condition would most probably exist on skid-mounted equipment assemblies such as the motor-driven AFW l pump assembly.
i The impellar drive shaft column sections of deep draft pumps are usually nonrigid, but the pump casings containing the attached piping nozzles are normally rigid.
Floor-mounted tanks and heat exchangers were procured with seismic rigidity requirements. Soil-mounted tanks were analyzed as structures and excluded from rigidity requirements.
Some floor-mounted HVAC equipment such as air handling units were qualified as assemblies, and there was not a requirement for seismic rigidity of such assemblies.
III. ACTION TO CORRECT EXISTING PROBLEM The FSAR will be revised at the next appropriate 10CFR50.71(e)
I update to clarify the interface considerations for pumps and other floor-mounted equipment with attached piping as follows:
i A safe design at floor-mounted equipment / piping interfaces is ensured by use of conservative nozzle load allowables or flexible metal hoses attached to the equipment nozzles. Floor-mounted equipment is i of ten rigid (f > 25 hertz), but this is not a requirement for seismic qualification.
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. . TVA's Resnonse D3.3-5 (Deficiency) Pump Fnadamental Frequency (Continued)
IV. ACTION TO PREVENT RECUBBENCE The Sequoyah Rigorous Analysis Handbook has been revised to give j additional instructions for handling floor-mounted equipment / piping interfaces. Those instructions include a requirement to evaluate the possible effects of nonrigid floor-mounted equipment seismic displacements before obtaining approval of increased nozzle load allowables for the equipment.
Regarding actions to address inconsistencies in the FSAR against design inputs, see TVA's response to deficiency D3.3-2.
V. OTHER RELEVANT INFORMATION OR COMMENTS The Sequoyah AFW pump / piping interface is a safe and conservative i
design. Low nozzle load allowables were satisfied by rigidly supporting the piping in the vicinity of the pump nozzles. Thermal expansion effects were conservatively analyzed assuming the pipe supports and pump nozzles to be rigid, and thermal expansion of the pump was included. Low damping (1 percent of critical) was used in
. the seismic analysis of the attached piping for SSE conditions and the combined effects of thermal expansion and seismic loads were included when satisfying nozzle load allowables. Finally, the calculated free displacement at the AFW pump nozzles due to SSE loading were less than 0.01 inch--a value which is insignificant relative to thermal expansion effects. Because this is a representative situation for existing floor-mounted safety-related j equipment with rigid attached piping, there is no need for further review of Sequoyah piping analyses relative to this concern.
In addition, earthquake experience, as dercribed in NUREG 1061
- Volume 2, indicates that TVA's analytical approach is adequate and i that additional conservatism is not warranted. All Sequoyah floor-mounted equipment is well anchored to avoid excess relative i displacements during a seismic event.
I Flexible metal hoses have been used at floor-mounted equipment nozzles when the combination of low nozzle load allowables and equipment nozzle displacements made a rigid attached piping solution impractical.
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D4.3-1 (Deficiency) Evaluation of Structures for Reinforcing Bar Cuts DESCRIPTION: ECN L6495 (Reference 1) replaced the angle valve (1-68-446A) with a straight through gate valve. In addition, the condensate pot had to be lowered, necessitating the installation of a new sleeve through the pressurizer cavity wall. Due to the installation of the sleeve, two horizontal and two vertical reinforcing bars were cut inside the pressurizer cavity wall. These were shown on drawing 41N730-1 (reference 2) related to Workplan 11847.
The USQD for ECN L6495 states that Civil Engineering Branch would evaluate the installation of the new sleeve to ensure that the structural integrity of the wall is not degraded. Contrary to this statement, the original calculations for this wall (reference 3) were not revised to evaluate the effects of cutting these reinforcing bars.
ECN L5202 (Reference 4) deals with the interf ace of conduit, cabling and piping between the existing diesel generator building, the powerhouse and the additional diesel generator building. Due to interference of certain reinforcing bars. Field Change Request (FCR) 1476 was issued to cut them as shown on drawing 10N321-2 (reference 5). Although the location of the reinforcing bar cuts are shown on this drawing, a review of calculation 10N320 (reference 6) showed that the effects of cutting reinforcing bars on the structural adequacy of the slab were not evaluated.
BASIS: Section 4.1.2.8 of TVA Engineering Procedure EN DES-EP 4.52 (reference 7) requires, in part, that the USQD be prepared for each LECN or TVA-approved design change against a nuclear power plant with an issued operation license before physical work is authorized. Section 4.5 specifies that completion of design work for an LECN involves, in part, a completeness review of the associated USQD to be certified by memorandum.
TVA Engineering Procedure EN DES-EP 2.03 (reference 8), Section 3.4, actes that failure to adhere to requirements specifically identified in the USQD evaluation nullifies the USQD evaluation.
REFERENCES
- 2. TVA Calculation 41N730-1 Concrete Steam Generator and Pressurizer Enclosure-Reinf., Rey, 1,6/29/79
Final Design, Rev. 2, 8/9/82
Sheet 2. Rev. 2, 7/22/83
- 6. TVA Calculation 10N320, Diesel Generator Building Superstructure and Slabs, Rev. 1, 10/27/80
- 7. TVA Engineering Procedure EN DES-EP 4.52, Engineering Change Notices (ECNs) Af ter Licensing-Handling, Revision 1, dated April 24, 1984
- 8. TVA Engineering Procedure EN DES-EP 2.03, Unreviewed Safety Question Determination-Handling and Preparation, Revision 6, dated April 24,1984 FP00;D431 SQEP Onsite
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TVA's Response D4.3-1 (Deficiency) Evaluation of Structures for Reinforcing Bar Cuts I. saggg The TVA design control program in effect at the time (approximately 1982 to present) utilized Field Change Request (FCR) procedures for minor changes when the changes were documented for inclusions in future analyses. However, use of FCR procedures did not require the preparation of engineering calculations for these minor changes.
In addition, as previously discussed in the response to deficiency D2.1-1, a contributing f actor to this condition involved weaknesses in the LECN/USQD review process.
II. EKTENT TO WHICH THE CONDITION COULD OR DOES EXIST IN OUR DESIGN Rebar cuts could occur in any reinforced concrete structure at the site but were approved by the FCR procedure. The drawings were revised indicating location of reinforcing bar cuts and calculations were performed where necessary. In areas of low stress, the reinforcing bar cuts were approved by inspection. The majority of the cuts were approved by inspection.
III. ACTION TO CORRECT EXISTING CONDITION All effected drawings have been revised to depict the as-built conditions including the reinforcing bar cuts. The engineers' evaluation of the reinforcing cuts were recorded on the FCR log sheet as approval that did not require design calculations.
This evaluation confirmed the acceptability of the condition.
Regarding actions to correct the LECN/USQD process, see TVA's response to deficiency 2.1-1.
IV. ACTION REQUIRED TO PREVENT RECURRENCE Policy Memorandum PM86-04 (DNE) issued April 25, 1986, states that engineering judgement must be accompanied by technical justification. This policy will be reflected in future revisions to DNE procedures. This is an ongoing DNE activity not associated with SQN restart.
Regarding actions associated with the LECN/USQD process, see TVA's response to deficiency 2.1-1.
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V. OTHER RELEVANT INFORMATION OR COMMENTS The reinforcing bar cuts were approved, reviewed, and verified by engineering judgement by engineers f amiliar with the design and TVA design procedures. Any previous reinforcing bar cuts which would have effected the approval of these cuts are documented on the design drawing. By documenting the rebar cuts on the design drawing, the ability to assess possible effects from accumulated rebar cuts is provided. The approval noted that design calculations would not be revised. The drawings were then updated to reflect the degraded condition. Future analyses would include any "as-constructed" configurations on the design drawings.
l Therefore this issue does not pose a safety problem.
The FCR log sheets note that the changes were approved and reviewed specifically without design calculations. The revisions were minor in nature, and a formal review of the calculations concluded that the changes were acceptable. The drawings were revised to record that the original design had been degraded. This methodology is no longer valid since the issuance of Policy Memorandum PM86-04 (DNE).
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D4.3-3 (Deficiency) Steam Generator Access Platform Design DESCRIPTION: ECN L5034 (reference 1) added platforms for access to the steam generator ports. These permanent platforms were built on the lower steam generator girders. Although calculations were performed for the design of these platforms, the effect of the platforms on the lower steam generator girders was not evaluated.
The USQD for ECN L5034 states that the additional loads are transmitted to the lower steam generator supports and do not exceed the design basis load. A review of the calculations for the platforms (reference 2) did not show any consideration related to the structural i
adequacy of the girders for the additional loads from the platforms.
BASIS: Section 4.1.2.8 of TVA Engineering Procedure EN DES-EP 4.52 (reference 3) requires, in part, that a USQD be prepared for each LECN or TVA-approved design change against a nuclear power plant with an issued operating license before physical work is authorized. Section 4.5 specifies that completeness review of the associated USQD is to be certified by memorandum. TVA Engineering Procedure EN DES-EP 2.03 (reference 4), Section 3.4, notes that f ailure to adhere to requirements
- specifically identified in the USQD evaluation nullifies the USQD evaluation.
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l REFERENCES i 1. TVA ECN L5034,11/18/81
- 2. TVA Calculation 48N908-1, Reactor Building Steam Generator Access Platform, Rev. 2 , 2 /20/ 86 .
- 4. TVA Engineering Procedure EN DES-EP 2.03, Unreviewed Safety Question Determination-Handling and Preparation, Revision 6, dated 4/24/ 86 .
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TVA Response D4.3-3 Steam Generator Access Platform Design I. ,CE gg A specific policy regarding documentation of technical justification for engineering judgement did not exist. In addition, as previously discuseed in the response to deficiency D2.1-1, a contributing f actor to this condition involved weaknesses in the LECN/USQD review process.
II. EXTENT TO WHICH THE CONDITION COULD OR DOES EXIST This deficiency exists for four platforms in each unit attached to the lower steam generator supports.
III. ACTION TO CORRECT EXISTING CONDITION Engineering will revise the calculations to confirm the acceptability of all accumulated loads on the lower steam generator supports. Because of the relatively small impact of the platform loads on the lower steam generator supports, this item is not considered a restart item. The completion of this activity is scheduled for September 1,1986, for unit 2.
Regarding actions to correct the LECN/USQD process, see TVA's response to deficiency D2.1-1.
IV. ACTION TO PREVENT RECURRENCE The Policy Memorandum PM86-04 (DNE) dated April 25, 1986, has been issued which states that use of engineering judgement must be accompanied with technical justificatien. This policy will be reflected in future revisions to DNE procedures. In addition, instructions were issued to specify requirements for considering the effects of attachments for future plant changes.
Regarding actions associated with the LECN/USQD process, see TVA's response to deficiency D2.1-1.
l V. OTHER RELEVANT INFORMATION OR COMMENTS l Previous calculations for reactor building steam generator platform with pipe loads were performed to justify the attachment of various supports to the lower steam generator supports. These calculations showed that the attachments added only slight stress j increases. During the design event for the lower steam generator
( support, the platform will be free of live load and the effect of the dead load of the platform on the heavy steel sections will be minimal.
The platforms were designed to minimize their effect on the lower steam generator support.
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, ,D4.3-4 (Deficiency) Cable Tray Support Response Spectra DESCRIPTION: P.ssponse spectra curves for 1 percent damping given a North-South earthquake were used in the design of the cable tray supports for elevation 714.0' (Reference 1). The response spectra curves for the same amount damping have higher accelerations in the East-West earthquake (reference 2). Although the present design criteria for cable tray supports allows the use of a less conservative 2 percent response spectra curve, the existing design may be unconservative (reference 3) when the competing effects of the accelerations are considered in combination with the damping values. That is, the North-South earthquake with 1 percent damping maybe less conservative than the East-West earthquake (which has higher accelerations) using 2 percent damping.
BASIS: TVA Design Criteria No. SQN-DC-V-1.3.4 (reference 3), Section 4.0, requires that seismic loads be computed and added to the static loads. The design criteria also states that these seismic loads are determined from the peak accelerations of floor spectra. In this particular deficiency, the TVA engineer failed to use the maximum peak
_ acceleration of the floor spectra for the two horizontal earthquakes.
REFERENCES
- 1. TVA Calculation 48N1330, 48N1332, 48N1333, Auxiliary Building Cable Tray Supports El. 714.0 ', Rev 2, 2/6/80.
- 2. TVA Dynamic Earthquake Analysis of the Auxiliary Control Building and Response Spectra for Attached Equipment, Rev 1.
- 3. TVA Design Criteria for Category I Cable Tray Support Systems, SQN-DC-V-1.3.4, Rev 0, 8/20/75.
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TVA Resnonse D4.3-4 (Deficiency) Cable Tray Support Response Spectra I. E&ggg Revision 1 to the CEB Report 80-20 contained a figure with the same designation as the original report but provided different response spectra.
II. EKTENT TO WHICN THE CONDITION COULD OR DOES EXIST IN OUR DESIGN The proper response spectra was used in the design of the cable trays supports. Therefore, a deficient condition does not exist.
III. ACTION TO CORRECT EXISTING CONDITION The existing calculations will be revised to emphasize that the referenced response spectra was from the original issue (Revision 0) of CEB Report 80-20.
IV. ACTION REQUIRED TO PREVENT RECURRENCE None V. OTHER RELEVANT INFORMATION OR COMMENTS The design calculations which contained the identified conditions referenced Figure C11 of CEB-Report 80-20. Revision 0, which contains the response spectra for 1-percent damping for the East-West earthquake.
However, the NRC inspector was given a copy of CEB-Report 80-20 Revision 1. Figure Cll of that revision of the report contained not the response spectra for 1-percent damping for the East-West earthquake, but the response spectra for 1-percent damping for the North-South earthquake at elevation 714. The design calculations were actually performed using the correct response spectra, but in the revised issue of the CEB report, the Eart-West response spectra was contained on a Figure Dil rather than Cll.
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D4.3-5 (Deficiency) Loads on Cable Tray Supports DESCRIPTION: The TVA Design Criteria for Category I Cable Tray Support Systems (Reference 1) states that for an 18-inch tray, the loads on cable tray supports should be 75 pounds per linear foot for the top tray and 45 pounds per linear foot for the additional trays.
TVA calculations for the cable tray supports MK 26B, MK 42 MK 18A, and MK 18B show that for the top trays only 45 pounds per linear foot were taken as the loading in the support design (Reference 2). These represent about 10 percent of the cable tray support calculations reviewed by the team. The rest of the support calculations adhered to the loading requirements of the design criteria. Since a loading lower than required by the criteria was used in the design, the as-built cable tray supports might be overloaded.
EASIS: TVA Design Criteria SQN-DC-V-1.3.4 (Ref erence 1), Section 4.0, requires that for an 18-inch tray the static maximum loading of the top tray in a tier should be 75 pounds per linear foot.
REFERENCES:
- 1. TVA Design Criteria for Category I Cable Tray Support Systems, SQN-DC-V-1.3.4, Rev 0, 8/20/75.
- 2. TVA Calculation 48N1330, 34, 35, 74, Auxiliary Building Cable Tray Support Below el 734.0', 2/2/79 SQEP Onsite D435
TYA. Reanonse D4.3-5 (Deficiency) Loads on Cable Tray Supports I. S& Egg A specific policy regarding documentation of technical justification for engineering judgement did not exist.
II. EXTENT TO WHICH THE CONDITION COULD OR DOES EKIST Since the e,ase which considered 75 pounds per linear foot in the top tray was not considered to be a controlling load case for design of any cable tray support component, this case was frequently not documented in the calculations for cable tray supports. While theoretically, this condition could apply to any top cable tray design, a cursory review by TVA did not identify any other discrepancies of this nature.
III. ACTION TO CORRECT EXISTING CONDITION SCR SQNCEB8622 covers several potential design deficiencies for the design of the cable trays supports. One of the items being investigated is the static load case for the top tray. Calculations for this item will be completed by August 1,1986, for restart considerations.
IV. ACTION REQUIRED TO PREVENT RECURRENCE In the future, engineering judgement will be accompanied by technical justification per TVA Policy Memorandum PM86-04 (DNE) which was issued April 25, 1986. This policy will be reflected in future revision to DNE procedures.
The intent of the Design Criteria SQN-DC-V-1.3.4 was to consider only the static load of 45 pounds per linear foot in the top tray l when combined with the dynamic load case. The design criteria will be revised to clarify the design requirements regarding the 75 pounds per linear foot static load in the top tr'ay.
V. OTHER INFORMATION OR COMMENTS The 75 pounds per linear foot static design load was intended to be applied as a construction load case only. It is expected that the adequacy of all cable tray supports will not be affected by this condition. The corrective action for SCR SQNCEB8622 will address the issue of the design load case which considers a load of 75 pounds per linear foot in the top tray.
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. D4.3-6 (Deficiency) Torsional Shear Stress Effects on Weld Design DESCRIPTION: Certain cable tray supports will be effected by torsional shear stresses during an earthquake due to their asymmetrical geometry.
TVA drawing 48N1334 (reference 1) shows that cable tray supports MK4 through MK4G are loaded on one side of the support. This configuration will lead to the twisting of the vertical structural member, inducing torsional stresses into the veld between this member and the embedded plate.
Team review of TVA calculation (reference 2) revealed that the additional stresses due to torsion on the welds were not considered in the cable tray support design.
BASIS: An incomplete analysis was performed for the design of the welds by not considering the torsional shear stresses. Such consideration is required by the AISC Specification that TVA invokes in Section 3.0 of Design Criteria SQN-DC-V-1.3.4 (reference 3).
REFERENCES
- 1. TVA Drawing 48N1334, Miscellaneous Steel Cable Tray Supports El. 714.0 '-Sh 6. Rev 15, 4/13/77.
- 2. TVA Calculation 48N1332, 48N1333, Auxiliary Building cable Tray Supports El. 714.0 ', Rev 2, 2/6/ 80.
- 3. TVA Design Criteria for Category I Cable Tray Support Systems.
SQN-DC-V-1.3.4, Rev. 0 8/20/75.
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. TVA Resnonse D4.3-6 (Dsficisucy) Tarsienal Shacr Strees Effscts en Wald Dssign
. . I. S&gli A specific policy regarding documentation of technical justification of engineering judgement did not exist.
II. EXTENT TO WHICH THE CONDITION COULD OR DOES EXIST Category I cable tray supports which will be subjected to torsion exist throughout the plant.
III. ACTION TO CORRECT EKISTING CONDITION SCR SQNCEB8622 covers several potential design deficiencies. One of the items being investigated is the effect of the torsion induced into the welded joint at the base plate. Calculations for this item will be completed by August 1,1986, for startup considerations.
IV. ACTION TO.PREYENT_ RECURRENCE The Design Criteria SQN-DC-V-1.3.4 does invoke AISC specifications which consider the effect of torsional shear stresses. The use of further evaluations by engineering judgement require technical justification per Policy Memorandum PM86-04 (DNE) issued April 25, 1986. This policy will be reflected in future revisions to DNE procedures. This is an ongoing DNE activity not associated with Sequoyah restart.
V. OTHER RELEVANT INFORMATION OR COMMENTS The resolution of SCR SQNCEB8622 utilizes engineering principles to determine a sample of " worst-case" supports. Two " worst-case" supports with a configuration similar to the supports MK 4 through MK 4G shown of 48N1334 were selected in the initial sample. Based on the preliminary evaluation of one of these samples. the total ,
impact of the addition of torsional shear on the weld at the base plate was less than 1 percent. Technically, the torsional shear on the weld at the base of the steel tube should be included, but it has been shown that the effect would be minimal.
The Category I cable tray supports which are subjected to torsional shear stresses are expected to be confirmed as structurally adequate pending completion of the analysis.
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U4.3-7 (Unresolved Item) Cable Tray Support Base Plate Analysis DESCRIPTION: TVA drawing 48N1333 (reference 1) shows that a surf ace base plate with threaded bolt anchors was used for certain cable tray supports. TVA calculation (reference 2) shows that the design of the base plate and the anchor bolts used the rigid base plate analysis.
TVA design standard (reference 3) requires that plate flexibility be
- considered to determine the anchor tensile loads.
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BASIS: Although the design of this particular base plate was performed
~
before the issuance of the design standard, there is a possibility that cable tray support base plates designed recently might still be using the rigid base plate approach. A nonconformance report (reference 4) written on the base plate design for pipe supports states that the requirements of the design standard have not been followed since the issuance of the standard.
REFERENCES
- 1. TVA Drawing 48N1333, Miscellaneous Steel Cable Tray Supports E1.714.0 - Sh 5, 7/16/75.
- 2. TVA Calculation 48N1330, Auxiliary Building Cable Tray Supports Below E1.734.0 ', 2/23/79.
- 3. TVA Design Standard DS-C1.7.1, General Anchorage to Concrete, Rev 3 11/16/84.
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. . TVA Response U4.3-7 (Unresolved Item) Cable Tray Support Base Plate Analysis ;
I. CAUSE Design criteria and design standards in effect at the time of most SQN support base plates did not specifically address the requirements for applicability of rigid analysis methods to base plate design.
II. EXTENT TO WHICH THE CdNDITION COULD OR DOES EXIST Rigid base plate analysis methods were used for the original design of cable tray supports at SQN. However, many base plates meet current rigidity requirements because of the thickness of the base plate or the use of stiffeners.
III. ACTION TO CORRECT EXISTING CONDITIONS The adequacy of anchorage design with consideration of base plate flexibility and as-built configuration is being investigated as one element of a broad evaluation of cable tray support design (SCR SQNCEB8622). The scheduled completion date for calculations which address restart considerations for this condition is August 1,1986.
IV. ACTION TO PREVENT RECURRENCE TVA Civil Design Standard DS-CI.7.1 (which has superseded the previous criteria related to base plate design) covers the analysis of the base plate and the determination of the anchor loads. This standard requires use of flexible base plate analysis methods if specific rigidity requirements are not met.
V. OTHER RELEVANT INFORMATION OR COMMENTS i As corrective action for a previous nonconformance (NCR SQNCEB8404),
a sampling program was performed on pipe supports some of which used rigid plate analysis for nonrigid plates. The sampling program showed that all pipe supports had an adequate expansion anchor factor of safety. The same conclusions would be expected for cable tray supports since the same design methods were used.
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D5.3-1 (Deficiency) Temporary Alterations using TACFs DESCRIPTION: During the review of the ECNs, the team noted that some ECNs were generated as followup documentation for the TACFs. Therefore, the team reviewed the Sequoyah plant administrative procedure for temporary plant modifications using Temporary Alteration Control Forms (TACF)
(Reference-1). The team reviewed selected TACFs issued during a period from 1980 through 1985 (Reference 2). !
TACF procedure Section 3.4 gives the plant manager an option to require the shift technical adviser to perform an independent verification of the controlled copies of the drawings. These drawings, located in the control room, are marked up for the "as-built" configuration by the individual completing work under the TACF. Failure to require an independent verification of "as-built" drawings also applies to the drawing changes made as a result of ECNs. We noted that G/Cs review and the TVA "three system" review both identified drawing errors in the main control room drawings.
The team believes that such errors would be minimized if the control room "as-ballt" drawings received independent verification when changes were entered.
The team reviewed eight TACFs (Reference 2) and noted the following:
TACF-84-81 Issued and installed on 7/2/84 but found no ECN issued.
TACF-80-625 Issued and installed on 7/31/80, PORC review for USQD was done six months later on 2/23/81. Signature block date on the USQD was missing.
TACF-82-214 Issued and installed on 10/6/82 and the ECN was issued two years later in December 1984. ECN is still open.
TACF-82-242 Issued and installed on 9/22/82. ECN 6217 was issued two years later in December 1984. ECN is still open.
TACF-84-81 Issued and installed on 7/2/84, ECN is not yet issued.
TACF-85-2009 Issued and installed on 7/8/85, no further related documents were found.
All of the above TACFs were found open. The team reviewed the computer printout list of the TACFs. TVA informed us that this list is published in the third quarter of each month. The team reviewed this list and noted that the list does not indicate any requirements for issuance of the followup documents required to close the TACFs. The team concluded that TVA does not have an effective tracking system for open TACFs which require an ECN or other followup documents to complete closure. The above referenced computer list does not provide any dates by which TACFs must be closed and/or followup documents issued. The team found that of TACFs could remain open for an indefinite period.
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DS.3-1 (Deficiency) Temporary Alterations using TACFs (continued)
The team observed that TACFs have been used routinely for permanent changes in the plant. This approach is apparently used when the plant believes that OE cannot support a change in a timely fashion. The TACFs do not require review and/or approval from the Office of Engineering. The team feels that in the absence of a proper tracking system, many TACFs used for the permanent changes will remain open and unanalyzed by the responsible design organization for a long time, resulting in the possible degradation of the safety systems. In addition, the team is concerned that the Office of Nuclear Power may not have sufficienct controls in place to be considered a responsible design organization for original engineering.
The team also noted that the TACF procedure Section 5.2 directs the use of a TACF for a long term change and a Maintenance Request (MR) or the Plant Instructions for a short term change. However, the procedure does not define the time duration for a short term or long term change.
BASIS: ANSI-45-2-11 (Reference 3) Section 6 " Design Verification" requires that the original design of the safety systems or components, must receive an independent verification. Section 8. " Design Change Control" of ANSI-45.2.11 states that " Changes shall be justified and subjected to design control measures commensurate with those applied to the original design."
REFERENCES
- 1. TVA Procedure #AI-9-Rev. 20 for Temporary Modifications using the TACF Form.
- 2. TVA TACFs #80-0625, 81-0458, 82-0214, 82-0242, 84-0081, 84-0107, 84-0115, and 85-2009.
- 3. American National Standards Institution ANSI-N45-2.11.
, 4. TVA's Topical Report #TVA-TR-75-1A Rev. 8.
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i SQN RESPONSE TO D2.3-1 (DEFICIENCY) LONG TERN UNINCORPORATED ECNs AND D5.3-1 (DEFICIENCY) TEMPORARY ALTERATIONS USING TACFs ;
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I. Cause The deficiencies occurred due to the following reasons:
- a. Before July 1985, design change requests (DCRs) were not required to be issued for TACFs in effect for more than 60 days. This resulted in some TACFs remaining open without DCRs being issued.
- b. The scheduling of work onsite is done using a priority system, and due to higher priority tasks, some TACFs and ECNs have remained open for extended periods of time.
II. Extent to Which the Condition Could or Does Exist in Our Design Approximately 158 TACFs are still open at Sequoyah Nuclear Plant.
III. Action to Correct Existing Condition
- The following actions have been taken to correct the deficiencies:
- a. Once a month a computer printout is issued to site managers identifying open TACFs and the associated ECNs.
- b. Every six months the Plant Operations Review Committee reviews the status of all open TACFs and associated ECNs.
- c. A safety evaluation has been performed for all open TACFs.
IV. Action Required to Prevent Recurrence The following actions will be taken to prevent recurrence:
- a. The Division of Nuclear Engineering will perform a safety evaluation on all safety-related TACFs before implementation. The appropriate procedures to implement this review will be revised by September 1, 1986.
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- b. Sequoyah has been involved in a long-term program that places and maintains management emphasis on controlling the use of temporary alterations and reducing the total number of temporary alterations in effect. For example, during the period January 1, 1985 through 4
January 31, 1986, the number of temporary alterations existing in the plant decreased from 297 to 158. Since January 1982, the number has decreased from approximately 990 to the present level.
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Of the open temporary alterations on January 31, 1986, approximately 84 date before 1984. The plant currently has an objective to close out this particular group by the end of the next refueling outage on unit 1 (INPO commitment). The overall status of each temporary alteration is tracked and reviewed by plant management on a periodic basis to ensure progress toward this goal.
Each open or existing temporary alteration has been reviewed previously to determine that the alteration was allowable within the technical specifications and that an unreviewed safety question (USQ) does not exist. As part of the current ongoing program to establish the adequacy of the present plant configuration for restart, plant management has directed that each open temporary alteration be reassessed to ensure that plant safety is not degraded due to the existence of the alterations. This evaluation is in progress and will be completed before plant restart.
V. Other Relevant Information 3r Comments None acknowledged for these deficiencies.
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. . US.3-2 (Unresolved Ites) Sizing Calculations DESCRIPTION: The team reviewed the operation of the steam driven AFW system during a loss of ac power. The steam throttle valve and the vent ,
fan for this system operate continuously during operation of the system and i are supplied from the 125V DC station battery system. The team reviewed '
the battery sizing calculations to verify that the battery system's capacity is adequate to meet the system demand.
The team determined that TVA does not have proper calculations for the sizing of the station batteries. The existing calculations (reference 3) do not address the correction factors for the operating ambient temperature and for aging. The calculations were performed before initial operation of the plant, and have never been reviewed or revised although, the loading profile of the de system has undergone changes.
In the absence of analysis and/or calculations, the team could not verify that the installed equipment has adequate capacity to meet the design ,
demands. Although the battery calculation was performed before the issuance of IEEE-485 (reference 2), it is necessary to use temperature correction and aging f actors for assessment of the battery's performance.
Changes to the loading must be evaluated to prove that the battery system will have a sufficient capacity to meet the design commitments per FSAR Section 8.3.2.1.
The team examined the system to determine if a similar problem exists with the sizing calculations for the battery charger and the 120V vital ac inverter. TVA informed us that sizing calculations for these components do not exist. These calculations were performed before procurement of these components but were not documented.
BASIS: TVA has committed to implement the guidance of ANSI N45.2.11.
Section 3 of this standard states that:
3.1 General Applicable design inputs, such as design bases, regulatory requirements, codes and standards, shall be identified, documented, and their selection reviewed and approved. Changes from specified design inputs, including the reasons for the changes, shall be identified, approved, documented and controlled.
The design input shall be specified on a timely basis and to the level of detail necessary to permit the design activity to be carried out in a correct manner and to provide a consistent basis for making design decisions, accomplishing design verification measures, and evaluating design changes.
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- - US.3-2 (Unresolved Item) Sizing Calculations (Continued) 3.2 Requirements The design input shall include but is not limited to the following, where applicable:
- 1. Basic functions of each structure, system, and component.
- 2. Performance requirements such as capacity, rating and system output.
- 3. Codes, standards, and regulatory requirements including the applicable issue and/or addenda.
- 4. Design conditions such as pressure, temperature, fluid, chemistry and voltage.
- 5. Environmental conditions anticipated during storage, construction, and operation such as pressure temperature, humidity, corrosiveness, site elevacion, vind direction, nuclear radiation, electromagnetic radiation and duration of exposure. . . .
REFERENCES
- 1. Sequoyah Unit 1 FSAR Section 8.3.2.1.1.
- 2. IEEE-Standard 485-1978-IEEE Recommended Practice for Sizing Large Lead Storage Batteries for Generating Stations and Substations
- 3. Battery contract No. 73C8-83800 Calculation for Battery Sizing l
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TVA's Resoonse US.3-2 (Unresolved Item) Sizing Calculations I. CAUSE The cause of this condition is attributed to lack of adherence to ,
design control procedures that required these calculations to be updated as design changes occurred.
II. EXTENT TO WHICH THIS CONDITION COULD OR DOES EXIST Performance of electrical calculations in general has been identified as a problem and documented in SCR SQNEEB8527. Basically, some existing electrical calculations have not been maintained and new calculations have not been performed for some design changes. This includes the sizing calculations for the vital batteries, chargers and inverters.
III. ACTION TO CORRECT EKISTING PROBLEM Sizing for the vital batteries, chargers and inverters has been reviewed as part of the ongoing upgrade of the electrical calculations. These components have been determined to have adequate capacity for the existing loads. This will be formally documented by August 27, 1986.
IV. ACTION TO PREVENT RECURRENCE An overall electrical calculations program has been established for SQN. This program's primary objective is to ensure that calculations of this type are developed and maintained as design changes occur.
In addition, it will ensure that design requirement changes are evaluated with respect to electrical calculations. This program was described in TVA's February 27, 1986 submittal to NRC.
V. OTHER RELEVANT INFORMATION OR COMMENTS None i
FP00;U532 l SQEP onsite
U5.3-3 (Unresolved Item) Motor-Operated Valve Thermal Overload Trip Setting DESCRIPTION: The team reviewed elementary diagrams (reference 4) for motor-operated valves and noted that the thermal overload trip for the ESF motor-operated valves is not bypassed by an accident signal. The team noticed that the overload trip settings of these thermal overload relays were set by the TVA construction staff, in accordance with the Procedure SNP-INSP-INSTR No.17 (reference 1). This procedure directs the technician to set the relays based on a range of 16-30 seconds of locked rotor current. The team found that some motor-operated valves take up to 60 seconds to complete their travel under the degraded voltage conditions;
' therefore, the arbitrary setting of 16-30 seconds may result in a trip by the overloads during valve travel. The team found that the setting duration of 16-30 was transmitted to the No. 315-LB-K(2) dated May 8,1974 (reference 2). However, the Office of Engineering did not perform analyses on a case-by-case basis to verify that a spurious trip of the thermal overload during the travel will not prevent the valve from completing its intended safety function.
BASIS: Incomplete travel of the motor-operated valves may defeat the engineered safety system's purpose of safe shutdown by preventing the safety systems to initiate or complete the required safety functions on demand.
REFERENCES:
- 1. SNP-INSP. INSTR NO.17 - Overload Relay Heater Inspection
- 2. Memo-315-LB-K(2), 5/8/74 - Selecting and Testing of MCC Overload Elements, Stations
- 3. 45N779 Sheets 1 through 16 - Wiring diagrams, 480V Shutdown Auxiliary Power Schematic Diagrams l
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-l TVA Response U5.3-3 (Unrassived Item) Mstor-0parated Valva Thsrmal Overload
', , Trip Sstting
- I. CAUSE Engineering determined that installation of a Class 1E MOV !
overload (OL) heater with the capability to withstand the valve's l locked rotor current for a minimum of 16 seconds and a maximum of )
30 seconds would be sufficient. Taking this approach for any valve l that has a travel time greater than 30 seconds appeared to create the potential for a spurious trip of the overload element.
II. EXTENT TO WHICH THE CONDITION COULD OR DOES EXIST See response in Section V.
III. ACTION TO CORRECT EXISTING CONDITION Operating experience has shown that the TVA established criteria is appropriate; therefore, there is no required action to correct the existing condition.
IV. ACTION REQUIRED TO PREVENT RECURRENCE Although operating experience has verified the adequacy of the established criteria, a policy memorandum (to be issued by August 15, 1986) will replace the engineering guidance noted by the NRC as t
- reference 2 of this deficiency. This policy memorandum will be used i by the TVA design organizations to establish criteria for determining overload trip settings of the thermal overload relays. The testing procedures will then be revised accordingly.
Y. OTHER RELEVANT INFORMATION OR COMMENTS TVA concurs that a case-by-case analysis has not been performed for
, the selection of OL heaters; however, the adequacy of TVA's approach is to be confirmed by the following:
- TVA has several test programs, both past and ongoing, that provide confidence that overload timing has not been a problem. First, the valves passed the system preoperational tests proving system operability. Second, valves are periodically timed in Surveillance
- Instruction (SI) 166. Third, in response to IE Bulletin 85-03 TVA 1
has set up a MOVATS testing program under Maintenance Instruction (MI) 10.43 which can monitor valve thrust and motor current. Another analysis required by IE Bulletin 85-03 involves recording the required thrust to position a valve under full differential pressure conditions. This will be a special test to further prove operability. Fourth, testing in accordance with SI 251 ensures the overloads trip at the proper point for locked rotor current. These i test programs have not found the 16-30 second overloads to be a j problem.
FP00;U533 SQEP onsite
4
. . US.3.4 (Unresolved Item) Diesel Generator Loading Calculations l DESCRIPTION: The team reviewed the diesel generator load analysis (reference 1) and 6.9kV one-line drawings (reference 2) and noticed the following items:
The diesel generator loading analysis was carried out using a 540 hp load lumped on the 25 second step of the sequencer for the AFW pump motor.
However, in reality the load gradually increases from 486 hy to 540 hp in 7 seconds. This 7-second ramp overlaps the 30-second step, which is the critical step for diesel loading. TVA did not perform an analysis to examine the effects of this situation on the voltage and frequency response and recovery limits to verify that the response is within the values given in Regulatory Guide 1.9 (reference 3).
The diesel generator loading analysis assumes that all the pressurizer heaters are turned "off" by the accident signal; however, the loading table !
correctly shows heaters which are energized. The 6.9kV bus one-line drawing (reference 2) has a draf ting error in note No. 6 in which tripping of the pressurizer heaters was omitted. These are considered documentation items in that the calculation used the correct configuration.
One assumption of the analysis states that the transformer nameplate rating was used for the load analysis; however, the loading table indicates that the actual connected loads ratings were used. The load table shows that the transformer load on the 6.9kV bus consists of two 1500kVA and one 300kVA transformers. However, design drawings (reference 2) show that there are three 1500kVA and one 300 kVA transformers. The team found that TVA did not analyze the effect of the third 1500kVA transformer, which remains connected to the 6.9kV bus during the zero block loading along with the other two 1500kVA and one 300kVA transformers. This will effect the frequency and voltage recovery of the diesel generator in the two-second interval between closing of the diesel breaker and application of the first block load.
BASIS: TVA has committed to implement guidance of Regulatory Guide 1.9 (reference 3). Section C-4 of this guide states that "The diesel generator unit design should be such that at no time during the loading sequence should the frequency and voltage decrease to less than 95 percent of nominal and 75 percent of nominal respectively (a larger decrease in I
voltage and frequency may be justified for a diesel generator unit that carries only one large connected load). Frequency should be restored to within 2 percent of nominal and voltage should be restored to within 10 percent of nominal within 60 percent of each load sequence time interval." ANSI-N45.2.11 (reference 4) Section 4, stipulates use of correct design inputs for the design analysis.
l FP00 ;U534 SQEP onsite
. , - - - - - - -,_ , . ,c.--.--__--,..--._,....--,.-...,e.,,,,,__ ,n ,. ._,._-,n ,_ ,.---,._,,,.._--...-,n,
t U5.3.4 '(Unresolved Item) Diesel Generator Loading Calculations (Continued)
REFERENCES:
- 1. Diesel Generator Load Analysis (B25 860204 300)
- 2. TVA Drawing No. 45N-724-1,-2,-3 6.9kV One-line Diagram TVA Drawing No. 45N-765 Sheets 1 through 18 and 6.9kV Shutdown Auxiliary Power Schematic Diagram
- 3. U.S. NRC-RG-1.9, Rev 2, - Selection, Design, and Qualification of Diesel-Generator Units used as Standby (Onsite) Electric Power Systems at Nuclear Power Plants
- 4. ANSI-N45.2.11,1975 - Quality Assurance Requirements for the Design of Nuclear Power Plants FP00;U534 SQEP onsite
TVA's Response US.3-4 (Unresolvsd Item) Dissal Ganzrator Locding
', , Calculetions
- 1. Q Lig The cause of this condition is attributed to errors in the preparation of diesel generator analysis and a drafting error on the 6.9 kV single-line diagrams, drawing numbers 45N724-1,-2 -3.-4.
II. uTrm TO WHICH THIS CONDITION COULD OR DOES EXIST The extent of this condition is being covered in the electrical calculations program described in TVA's February 27, 1986 submittal to the NRC.
III. ACTION TO CORRECT EXISTING PROBLEM The diesel generator analysis has been revised and issued on May 16, 1986, to correct these items. The draf ting errors on the 6.9 kV single-line diagrams will be corrected by August 30, 1986.
IV. ACTION REQUIRED TO PREVENT RECURRENCE A detailed procedure for preparation of diesel generator loading calculation will be developed as part of the electrical calculation program. This document will be issued by November 28,1986, and will be applicable to all TVA nuclear plants.
V. OTHER RELEVANT INFORMATION OR COMMENTJ NRC's review identified a discrepancy in an assumption with respect to the use of transformer nameplate rating versus the actual connected loads. For the smaller 6.9 kV to 480 V ac transformers, the transformer nameplate data was used; however, for the larger transformers (e.g. , 6.9 kV to 480 V ac shutdown board transformers),
actual load data was used. This information was clarified in the May 16,1986 revision to the diesel generator analysis.
DNE concurs that there were some errors in the diesel generator analysis; however, the reanalysis concluded that these errors did not affect the results. In addition, there are some drafting errors on the 6.9 kV single-line diagrams, and we concur with NRC that these are documentation items since the calculation used the correct configuration.
i FP00 ;U534 SQEP onsite i
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US.3-5 (Unresolved Item) Loss of Control Power Annunciation DESCRIPTION: The team reviewed TVA drawings (References 1 and 2) for the l 6.9 kw feeder breaker control circuit for the AFW pump motor and noticed '
that the breaker control circuit does not have a provision to detect and annunuciate the loss of control power. In the event of loss of control of 1 power, the circuit breaker will not be able to close when required. This will prevent automatic operation of the AFW pump, a required function J important to the safety of the plant. TVA informed the team that the control room operators monitor the breaker status indicator lights. The "off" status of these lights (neither open nor closed indication) can be taken as an indication of the loss of control power. TVA further informed the team that at the end of each shift, a documented record is prepared by the operator for those lights which have changed their status (from "0N" to "0FF" or from "0FF to "0N"). The team acknowledged these comments; however, noted that it is possible that a change in status of these lights could go unnoticed by the operators for some time.
Regulatory Guide 1.47 states that, "A practical indicating system covering a wide range of commonly expected conditions, however could be designed if it included provisions for automatic indication of each bypass or deliberately induced inoperable condition that meets all three of the following guidelines.
- 1. The bypass or inoperable condition affects a system that is designed to perform automatically a function that is important to the safety of the public,
- 2. The bypass will be utilized by plant personnel or the inoperable condition can reasonably be expected to occur more frequently than once per year, and
- 3. The bypass or inoperable condition is expected to occur uhen the affected system is normally required to be operable.
The team feels that AFW system meets the three conditions stated above. The AFW system is important to the safety of the public; plant operators use removal of control power to maintain equipment; and the inoperable condition is expected to occur when the AFW system is required such as during accident conditions. In addition, it is possible that loss of control power may occur due to blown fuses, short circuits, and open circuits.
RG 1.47 further states that " Bypass indication should aid the operator in l recognizing the effects on plant safety of seemingly unrelated or i insignificant events. Therefore, the indication of bypass conditions should i be at the system level, whether or not it is also at the component or channel level. For example, in a design which utilizes a de power system to i
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e5.3-5 (Unresolved Item) Loss of Control Power Annunciation (continued) control circuit breakers, de-energizing during maintenance should result in an indication for each safety system whose operation is dependent on that power system that the safety system is inoperable." The team feels that the above guidance also applies when de-energizing of de control power occurs automatically due to system fault.
BASIS: TVA has committed to implement the guidance of Regulatory Guide 1.47, Bypass and Inoperable Status Indication (Reference 3). Loss of control power for the breaker control circuit will prevent the auxiliary feedwater pump from being able to respond to system demand, yet this condition is not indicated as an inoperability at the system level.
REFERENCES
- 1. TVA Drawing #45N-724-1,2,3 & 4-6.9tv One Line Diagram.
- 2. TVA Drawing #45N-765 SH.1 through 18,6.9kV Shutdown Aux Power Schematic Diagram.
- 3. US NRC RG 1.47 - Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems.
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- TVA Response US.3-5 (Unresolved Item) Loss of Control Power Annuciation l I. CAUSE In a letter from L. M. Mills to E. Adensam dated September 13, 1982, TVA !
informed the Commission of its plans for removal of the Status Monitoring System (SMS) at Sequoyah Nuclear Plant (SQN). The SMS provided automatic indication of bypass or deliberately induced inoperable conditions per the requirements of Regulatory Guide 1.47 Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems. The subject letter stated that the proposed Technical Support Center (TSC) design will implement the function previously provided by the SMS.
II. EITENT TO WHICH THIS CONDITION COULD OR DOES EXIST Automatic indication of bypass or deliberately induced inoperable conditions per Regulatory Guide 1.47 requirements has not been implemented on the TSC system.
III. ACTION TO CORRECT EXISTING CONDITION TVA intends to implement the Regulatory Guide 1.47 function. A scoping document is being prepared to assist in the evaluation of whether to proceed with implementation on the TSC system or to proceed with an alternate design. Interim methods to provide operator information are being reviewed.
IV. ACTION REQUIRED TO PREVENT RECURRENCE The design concept for the implementation of the Regulatory Guide 1.47 requirements will be discussed with NRC-NRR in the very near future and prior to any implementation to ensure that an acceptable approach is developed and that interim measures are acceptable.
V. OTHER RELEVANT INFORMATION OR CONMENTS l
None.
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D6.1-1 (Deficiency) AFW Pump Discharge Pressure Switch Ratings DESCRIPTION: Auxiliary feedwater pump discharge pressure switches 1-PS-3-148. -156, -164, and -171 provide a safety-related interlock for positioning of bypass control valves. Gilbert / Commonwealth (G/C) reviewed two Engineering Change Notices (ECNs) where existing pressure switches were replaced with environmentally qualified devices (References 1 through 4).
The team noted that.G/C had not compared the technical requirements for the replacement instruments with the original procurement instrument data sheet to assure that design basis requirements remained satisfied. G/C stated that such a comparison was not in their assigned scope of review. The instrument data sheet is used to specify technical requirements for procurement of the pressure switches from equipment vendors. Consequently, the team performed this design basis comparison and determined that an intermediate replacement had also been made for these pressure switches.
Results of this comparison are provided below:
Technical Original Interim Current Characteristic (Ref. 5) (Ref.6) (Ref. 7)
Proof Pressure 4500 psi *2000 psig or 2000 psig or 150% des.pr. 150% des.pr Maximum Pressure 1200 psig 1650 psig *1085 psig Process Connection 0.5 inch 0.25 inch 0.25 inch Contact Rating 0.4 ampere 0.5 ampere 0.5 ampere Contact Voltage 140 VDC *125 VDC 140 VDC Contact Action Close-decr. *0 pen-decr. Open-decr.
Trip Setpoint 500 psig *485.3 psig *400 psig Adjustment Range 285 to 660 psi *5 to 200 psig 45 to 550 psig Manufacturer Custom Comp. Asco Static-0-Ring The team found no indication that those changes denoted by an asterisk (*)
l to the original design basis for the interim or current replacements had
! been technically documented. TVA stated that existing plant documentation was not revised when these modifications were initiated; rather, a new instrument data sheet was prepared in each instance.
For the interim modification, the voltage specification of 125 volts de was in error since it did not accommodate a battery recharging condition. The trip setpoint change to 485.3 psig was not supported by a calculation and implied an unrealistic setpoint accuracy for this instrument.
The current modification has a design basis impact for maximum pressure and trip setpoint characteristics. The 1085 psig maximum design pressure did not provide for additional margin above the maximum system operating pressure, and the trip setpoint change to 400 psig was not supported by an appropriate c'alculation.
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t The team noted that Gilbert / Commonwealth's (G/C's) review had not identified that the TVA instrument data sheets and Static-0-Ring vendor drawing were not labelled as a safety-related for the end user. A minor catalog number transposition error between the vendor drawing (5N6-B45-NX-CIA-JJTTX6) and the TVA instrument data sheet (5N6-B45NX-CIA-JJTTX6) was noted by the team.
BASIS: For the design modifications involving both interim and current replacement pressure switches, a number of changes were made in the design basis without a documented engineering justification when the modification was prepared, approved and Laplemented. The team did not find evidence that the reduction in proof pressure and changes in maximum operating pressure values were satisfactory from a system perspective as required by IEEE Standard 279-1971 Section 3(7). Setpoint changes for these safety-related instruments were not supported by calculations as required by Sections 3(4), 3(5), and 3(9) of IEEE Standard 179-1971. One change in the direct current voltage rating of the switch contacts did not conform with IEEE Standard 279-1971 Section 3(7).
Instrument data sheets and vendor drawings were not labelled as safety-related, even though other TVA drawings have been marked in accordance with IEEE Standard 494-1974. This aspect was not identified in the G/C review.
REFERENCES
- 1. ECN-L-5823, AFW Pump Disch. Press. Sw. Replacement, Rev. O 10/5/83.
- 2. ECN-L-5883, AFW Pump Disch. Press. Sw. Replacement, Rev. O ,10/20/ 83.
- 3. Gilbert / Commonwealth Technical Issue Data Sheet 5. Rev. O,1/24/86.
- 4. Gilbert / Commonwealth Technical Issue Data Sheet 16, Rev. O,1/24/86.
- 5. TVA Instrument Data Sheet Specification 1596, Rev. O, 6/11/75.
- 6. TVA Instrument Data Sheet PR-W-3098 (Watts bar), Rev. 2, 8/3/82.
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I FP00 ;D611 SQEP onsite - July 20, 1986
D6.1-1 (Daficiency) AFW Pump Dicchargs Prassure Switch Ratings TVA's Response I. {Ay,1g Refer to Section V.
II. EKTENT TO WHICH THE CONDITION COULD OR DOES EXIST Refer to Section V.
III. ACTION TO CORRECT EXISTING CONDITION Refer to Section V.
IV. ACTION REQUIRED TO PREVENT RECURRENCE Refer to Section V.
V. OTHER RELEVANT INFORMATION OR COMMENTS The auditor addressed his questions to a person who was not knowledgable about the subject calculations.
Documentation does exist for differences in instrument data sheets.
The data sheet differences were caused by a design change to the system which involved replacing a pressure control valve with a cavitating venturi. This system change was documented by a feasibility study.
Also, a setpoint calculation was performed which documented the change in setpoints. The proof pressure on the original data sheet was not the minimum acceptable, but was the actual proof pressure of the switch being used. The later data sheets listed the minimum acceptable.
The interim data sheet was not written for procurement but was for transfer from another project. Therefore, 125V DC nominal voltage was listed rather than 140V DC maximum voltage.
The adjustment range on the interim data sheet was a mistake since it does not incorporate the setpoint.
l The maximum pressure on the current data sheet is incorrect. The current pressure switch can withstand 2000 psig and can withstand 1500 psig without affecting the setpoint. The maximum system pressure at the pressure switch is less than 1500 psig.
l FP00;D611 SQEP onsite - July 28, 1986 L
D6.1-2 (Deficiency) Feedwater Bypass Control Valve Solenoid Replacement DESCRIPTION: Replacement solenoid valves 1-FSV-3-35A.-48A,-90A, and -103A were installed to improve the response time of the main feedwater bypass control valves (Reference 1). Similar replacement solenoid for Sequoyah Unit 2 were designated as non-quality assurance material (reference 2). A
- subsequent unreviewed safety question determination for this modification stated that Class 1E solenoid valves were provided; however, this requirement was not satisfied (Reference 3).
Gilbert / Commonwealth's review of this modification identified documentation inconsistencies in the safety-related versus non-safety-related designation for these replacement solenoid valves, and recommended that the solenoid valve and its electrical circuits be made safety-related to provide redundancy for main feedwater isolation from the steam generators (Reference 4 through 6). During the Gilbert / Commonwealth plant walkdown, the non-Category I seismic mounting of the replacement solenoid valve was identified as a deficiency.
! The team held a number of discussions with TVA personnel regarding the feedwater isolation safety function required of these solenoid valves.
IVA's reasons for using non-Class 1E replacement solenoid valves were based on the valve's location in the non-Category I turbine building, the desire to avoid use of Class 1E cables in this building, and the fail-safe characteristics of the solenoid. However, this analysis failed to address the need to satisfy the isolation safety function requirement. TVA should have reccanized this safety function requirement when the solenoid valves were replaced, and should have upgraded the original non-Class 1E solenoid valves at that time, consistent with the USQD.
In response to the recent Gilbert / Commonwealth review of completed design modifications, TVA has indicated that a Class 1E solenoid qualified for service conditions that exclude 10CFR50.49 environmental considerations will be specified and that detailed solenoid mounting requirements will be developed to limit seismic responses.
l BASIS: A feedwater isolation safety function has been required of the solenoid valves associated with the feedwater bypass control valves.
Replacement solenoid valves did not meet the Class 1E requirements needed to ensure accomplishment of this safety function. TVA's reasons for providing non-Class 1E solenoid valves did not adequately address the need to satisfy the feedwater isolation safety function. The installation of non-Class 1E-solenoid valve violates the unreviewed safety question determination.
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'D6.1-2 (Deficiency) Feedwater Bypass Control Valve Solenoid Replacement l (continued)
REFERENCES
- 1. ECN-L-5717, EW Bypass Control Valve Solenoid Change, Rev. O, 5/14/80.
- 2. TVA Memorandum, SWP 801016 022, Transfer of Solenoid Valves from Watts Bar Nuclear Plant to Sequoyah Nuclear Plant, 10/15/80.
- 3. TVA Unreviewed Safety Question Determination for ECN-L-5717, SWP 830217 802, 2/17/83.
- 4. Gilbert / Commonwealth Technical Issue Data Sheet 7, Rev. O, 1/24/86.
- 5. Gilbert /Commomwealth Technical Issue Data Sheet 13, Rev. 1, 1/28/86.
- 6. Gilbert / Commonwealth Observation Sheet, Rev. O, 1/24/86.
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1 D6.1-2 (Deficiency) Feedwater Bypass Control Valve Solenoid Replacement TVA's Resoonse I. EAlf]
The non-1E solenoids were transferred from Watts Bar'and installed by a TACF. ECN L5717 was written to make this modification permanent.
ECN L5717 specified IE solenoids for this application. A detailed review of the documentation was not performed to ensure the solenoids were Class IE.
In addition, as previously discussed in the response to deficiency 2.1-1, a contributing factor to this condition involved weaknesses in the LECN/USQD review process.
II. EXTENT TO WHICH THE CONDITION COULD OR DOES EXIST TVA'is currently reviewing all TACFs in the Design Baseline Program to ensure design adequacy. This review will be completed before restart and extent of the condition will be known at that time.
III. ACTION TO CORRECT EXISTING CONDITION SCR SQNEEB8624 was written to track this modification. ECN L6692 is being processed to replace these solenoids with Class lE solenoids.
These solenoids will be replaced in both units before the units restart.
Regarding actions to correct the LECN/USQD process, see TVA's response to deficiency D2.1-1.
IV. ACTION REQUIRED TO PREVENT RECURRENCE As a result of the Gilbert / Commonwealth " Assessment of the Design Control Program for Sequoyah Nuclear Plant" dated October 1985 TVA instituted a modification criteria which establishes a design basis for each modification. This new modification criteria which is contained in the SQN Project Manual as an expansion to OEP-6 " Design Input" became effective January 15, 1986.
Regarding actions associated with the LECN/USQD process, see TVA's response to deficiency D2.1-1.
. V. OTHER RELEVANT INFORMATION OR COMMENTS ECN L6692 will replace the unqualified solenoid with a Class IE solenoid.
~ '5E"91_ . ___ _ _ _ _ _ _ . _ . . . . _ _ _ _ _
b6.1-3 (Deficiency) AFW Pump Suction Pressure Switch Setpoint Calculation DESCRIPTION: Automatic transfer of auxiliary feedwater pump suction from l the condensate storage tank to the essential raw cooling water system is l accomplished by safety-related instruments monitoring for auxiliary feedwater pump low suction pressure. Gilbert /Commomwealth could not determine whether setpoint values and time delay requirements were adequately reevaluated as required after system testing and stated that existing calculations did not take into consideration the Technical Specification limiting safety setting requirement.
Gilbert / Commonwealth recommended that a new calculation for these setpoints be performed, but did not identify that the existing calculation of record (Reference 5) should have been updated or superceded when the pressure switch modifications were made. The team noted that this calculation had '
not been referenced, updated, or superceded as a result of setpoint changes listed in a 1981 memorandum (Reference 6) and three subsequent change notices (References 1, 2, and 3).
In their review, Gilbert / Commonwealth did not state that this calculation had not been marked as a safety-related calculation, and that numeric changes made in input values were not carried through to calculational results.
BASIS: The adequacy and control of existing design basis documentation was not addressed in that the original setpoint calculation should have been referenced in subsequent TVA design documents and then either corrected or superceded. Such controls are required by ANSI N45.2.11 section 4.2, Design Analyses, and section 8. Design Change Control.
REFERENCES
- 1. ECN-L-5721, AFW Pump Suction Setpoints, Time Delays, Rev. O, 4/3/84,
- 2. ECN-L-6124, AFW Pump Suction Press. Sw. Setpoints. Rev. O, 4/25/84.
l 3. ECN-L-6254, AFW Pump Suction Press. Sw. Setpoints, Rev 0, 11/19/84.
- 4. Gilbert / Commonwealth Technical Issue Data Sheet 15, Rev. 1, 1/28/86.
- 5. Tva Calculation, SQN-Ca-DO53, AFW Setpoints, Rev. O, 4/6/79.
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TVA's Response D6.1-3 1*. geg1E This deficiency resulted from the failure to consider tiie original setpoint calculation in subsequent modifications.
II. EXTENT TO WHICH THIS CONDITION COULD OR DOES EXIST TVA is reviewing all setpoint calculations for adequacy. This study will be complete before restart. If conditions adverse to quality are discovered during this study, they will be documented.
III. ACTION TO CORRECT EXISTING PROBLEM A new calculation was performed on March 6,1986, documenting the set points and time delays for the automatic transfer of auxiliary feedwater pump suction from condensate storage tank to ERCW system.
The setpoint is now acceptable.
IV. ACTION TO PREVENT RECURRENCE As a result of the Gilbert / Commonwealth " Assessment of the Design Control Program for Sequoyah Nuclear Plant" dated October 1985 TVA instituted a acdification criteria which establishes a design basis for each modification. This new modification criteria which is contained in the SQN Project Manual as an expansion to OEP-6 " Design Input" became . effective January 15, 1986.
V. OTHER RELEVANT INFORMATION OR COMMENTS None l
FP00;D613 SQEP onsite I
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D6.2-1 (Deficiency) Reactor Coolant System Narrow Range Resistance
- Temperature Detector Qualifications Category Change DESCRIPTION: Reactor coolant system temperature detectors are used in the reactor protection system for the determination of the reactor coolant system average temperature which is used to compute reactor trip parameters such as Overpower and Overtemperature Delta T. These detectors were originally designated as TVA qualification category A (References 1 and 2),
but were changed to category C. Category A components are those that are subject to mitigate the consequences. Category C components are those that are subject to harsh environmental conditions of design basis accidents but are not required for mitigation of that accident and whose failure in any mode would not be detrimental to plant safety. The stated basis for this change was their use as back-up rather than primary trip signals as described in FSAR transient and accident analyses (Reference 3).
Westinghouse had provided a similar basis for the elimination of environmental and seismic qualification for ex-core neutron detectors in late 1983 (Reference 4).
The team did not agree with this change in qualification category. The instrument sensors connected to the reactor protection system must be environmentally qualified for their intended service conditions. During the inspection, the team was advised that the Office of Engineering had initiated a revision to the engineering change notice to restore these sensors to qualification category A.
BASIS: The change from qualification category A to C violated a requirement that reactor protection system sensors be qualified for their intended service conditions as stated by Section 4.4 of IEEE Std. 279-1971. All reactor trips should de designed to meet the requirements of IEEE Std. 279 in order to prevent a possible degradation of the reactor protection system (Reference 5).
REFERENCES:
- 2. TVA Unrevised Safety Question Determination for ECN-L-6449, B25 850918 509. Rev. 1, 9/18/85.
- 3. Tva Quality Information Release, B45 851231 268, 10CFR50.49 Category and Operating Times Calculation Change for Reactor Coolant System Resistance Temperature Detectors.
- 4. Westinghouse Letter, WAT-D-5709, NEB 830930 637, Seismic and Environmental Qualifications of Ex-Core Neutron Detectors, 9/22/83.
- 5. NUREG 0800, Branch Technical Position ICSB 26 Requirements for Reactor Protection System Anticipatory Trips, pg. 7A-18 Rev. 2, 7/81.
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D6.2-1 (Deficiency) Reactor Coolant System Narrow Range Resistance Temperature Detector Qualification Category Change TVA Response D6.2-1 I. Egg]
The cause for this noted discrepancy is the lack of thoroughly documented and accessible design basis for reactor trip functions.
This was compounded by incorrect information supplied by Westinghouse.
II. EXTENT TO WHICH CONDITION COULD OR DOES EXIST This condition could potentially exist for other reactor trip functions.
III. ACTION TO CORRECT EKISTING CONDITION For the reactor coolant system narrow-range resistance temperature detectors (RTDs), we had re-evaluated their category and determined that they were Category A for the inside containment main steam'line break event prior to issuance of the finding. (Reference Quality Information Release NEB 86041,10CFR50.49 Category and Operating Times.)
Accordingly, the subject RTDs are included in the 10CFR50.49 program.
IV. ACTION REQUIRED TO PREVENT RECURRENCE An updated list of required reactor trips has been requested from Westinghouse. Upon receipt, the trips will be reviewed against the existing category and operating times to ensure that the equipment associated with those required reactor trips is properly specified and qualified. This will be accomplished prior to restart.
V. OTHER RELEVANT INFORMATION OR COMMENTS TVA's qualification categories which are documented in the Category and Operating Times Calculations are for the purpose of establishing the scope of equipment to be included in the 10CFR50.49 qualification program. 10CFR50.49 states that equipment covered by the rule is that relied upon to remain functional during and following a design basis event to ensure (1) integrity of reactor coolant pressure boundary.
(2) capability to shut down the reactor and maintain it in a safe j shutdown condition, and (3) capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to 10CFR100 guidelines. The rule further limits its scope by excluding natural phenomena and external events and equipment in a mild environment. Thus, some reactor protection system equipment as specified by IEEE 279-1971 may not fall in the scope of the 10CFR50.49 qualification program since it may be located in a mild plant environment or may be required only for specific events which do j not produce harsh environments. Even though some RPS features may not require inclusion in the 10CFR50.49 program, they are required to meet the intent of IEEE-179-1971 and related standards and are qualified for their intended service.
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D6.3-1 (Deficiency) Specification of Hydrostatic Test to Demonstrate Instrument Pressure Boundary Integrity After Seismic Qualification Testing DESCRIPTION: Process instruments connected directly into safety class piping must conform with seismic category I requirements and maintain the pressure boundary integrity of safety class piping. The demonstration of system pressure boundary integrity is ordinarily achieved by separate hydrostatic pressure tests performed immediately before and after a seismic qualification test.
During th team's review of specific process instruments used at Sequoyah, it was determined that procedural guidance existed for the specification of hydrostatic test requirements. For example, TVA procedure OEP-09, which has been applicable to instrument procurement since June 1985, stated that tests and acceptance criteria for hydrostatic pressure tests may be included in procurement specifications where applicable (Reference 3). In addition, the
, Sequoyah Office of Engineering Projc..t Manual specifically required that component test requirements include a consideration of hydrostatic pressure
- t9sts (reference 4).
However, the team determined that TVA had not specified a design performance test for instruments purchased for recent plant modifications (References 1 and 2). For one procurement contract, the instrument vendor successfully demonstrated hydrostatic pressure integrity before and after the seismic qualification test (Reference 5). However, for a second procurement contract, the vendor did not perform a hydrostatic test after the seismic qualification test (Reference 6).
BASIS: TVA procedural requirements with spect to the specification of a hydrostatic pressure test after seismic gealific stion have not been satisfied. The pressure boundary integrity of one set of instruments connected to the reactor coolant system has not been demonstrated after the seismic qualification test.
I REFERENCES
- 1. ECN-L-6380, RCP Bypass Line dp Switch Replacement, Rev. O, 4/29/85
- 2. ECN-L-5620, AFC Turbine Discharge Pressure Transmitter, Rev. O, 3/14/83
- 3. TVA Procedure OdP-09, Attachment 9, General Content and Format l Requirements for Procurement Specifications, section 8.2.2
- 4. TVA OE Sequoyait Project MANUAL, lection II, Expansion to OEP-06, item 4.4, Test and Inspection Requirements, . 10/86.
- 5. Foxboro N-E11DM Differential Pressure transmitter Qualification Report,
- B70 851125 528, Rev. O, 1/28/86.
- 6. Static-O-Ring 103AS-bb803-NX-JJTTX6, differential Pressure Indicating Switch, Action Environmental Test "orp. Reports 18878-84N-1, Rev. 1, 8/30/84 and 18878-84N-3, Rev. 1, 9/25/84.
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'. . D6.3-1 (Daficiency) Sptcificatica cf Hydrostatic Test to Densnatreta
, , Instrument Pressure Boundary Integrity Af ter Seismic Qualification Testing TVA Response D6.3-1 I. QEg This item has been determined to not be a problem. See Section V.
II. EXTENT TO WHICH THE CONDITION COULD OR DOES EKIST Refer to Section V.
III. ACTION TO CORRECT EXISTING CONDITION Refer to Section V.
IV. ACTION REQUIRED TO PREVENT RECURRENCE Refer to Section V.
V. OTHER RELEVANT INFORMATION OR COMMENTS Safety-related, fluid system components must be shown to satisfy applicable industry codes and standards, NRC requirements and TVA
' precedures, criteria, and specifications.
This would include hydrostatic pressure testing for ANSI or ASME code components, and seismic testing which would typically demonstrate operability under imposed seismic loading. It is accepted practice in common use throughout the nuclear industry that these two requirements are considered totally separate--totally independent of each other.
There is no nuclear industry or NRC commitment that a hydrostatic pressure test be imposed on a safety-related fluid system component following its seismic qualification testing.
However, to provide additional assurance, the following actions are being accomplished:
A. Revised data sheet No. l AR1 has been issued to document the maximum design pressure as 2485 psig instead of 2400 psig.
B. As allowed by ANSI B31.1, we will pneumatically pressure test our units onsite to 3000 psig (1.2 x 2485 = 2982 PSIG). 3000 psig is the rated overrange pressure for the units.
C. The vendor has been requested to confirm in writing before restart that:
- 1. The units can withstand up to 8000 psig without leaking.
However, 3000 psig is the maximum pressure the unit should be subjected to without danger of change in operating characteristics, shif t in set point or damage to the device.
- 2. The original seismically qualified unit (prototype) will be tested pneumatically at 3000 psig.
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r U6.3-2 (Unresolved Item) Engineering Change Notice (ECN) Quality Assurance and Seismic Analysis Designations DESCRIPTION: During the preparation, review, and approval of an ECN, the application of quality assurance and seismic analysis requirements must be designated by yes or no entries on the form (ruferences 1 through 3).
The team reviewed eighty (80) individual ECKs for the 1980 through 1985 period, and noted an approximate 9 percent error rate and a 10 percent reversal rate for the designation of quality assurance and seismic analysis requirements. Several variations in these designations were noted by the team; namely, the application of cne requirement without the other, the application of neither requirement for safety-relateu equipment modifications, and the reversal of an_ initial designation for one or both of these requirements.
The team believes that the final designation of the following ECNs were in error by specifying the application of quality assurance without requiring seismic analysis. Each modification involved one or more Class lE components which are required to meet both the quality assurance requirements of 10CFR50 Appendix B and the seismic requirements of IEEE Std 344-1975. A "no" entry for seismic analysis on the ECN would not provide confirmation of seismic adequacy for these Class lE components:
ECN-L-5057, Reactor Coolant Pump UV and UF PPS Sensors.
ECN-L-5092, AFW Turbine Resistor Box Moved to Wall Mount.
ECN-L-5314, Pressure Switch Moved Outside Crane Wall.
ECN-L-533 9, AFW Flow Control Valve Replacement.
ECN-L-5490, AFW Speed Control Moved to Wall Mourt.
ECN-L-5717, AFW Control Valve Solenoid Replaced.
ECN-L-5758. Traveling Screen Bubbler dP Instrument Added.
The team noted that the following ECNs had a reversal of the initial determination for one or both of these requirements:
ECN-L-5057, Reactor Coolant Pump UV and UF PFS Sensors.
QA changed fron no to yes.
ECN-L-5620, AFW Turbine Pump Surveillance Point Added.
QA and seismic changed from no to yes.
ECN-L-5717, AFW Control Valve Sole-old Replaced, QA changed from no re'yer.
ECN-L-5726. Instrument Line Iai? st f ; t and Re-Routing, QA and seismic tha sti i tan no to yes.
2CN-L-5760, Venturi Flow Restrictora Added, QA and seismic changed from no to yes.
f ECN-L-5789, Main Feedwater Solenoid Valve Leakage.
QA and seismic changed from no to yes.
ECN-L-5884, AFW Flow Transmitter Changed, seismic changed from no to yes.
4 ECN-L-6109, Reactor Coolant Pump Oil Reservoir Level Monitor, seismic changed from no to yes r
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U6.3-2 (Unresolved Item) Engineering Change Notice (ECN) Quality Assurance o
- and Seismic Analysis Designations (Continued)
Since approximately 20 percent of the initial determinations for ECN's reviewed by the team were in error, the team's opinion is that individual engineers have had obvious difficulty in understanding how the written criterion was to be applied to a given design modification situation. This view appears to be supported by the additional ECNs identified by the team that remained in error following review and approval steps. The team's assessment is that while the criterion was technically correct, they lacked sufficient clarify necessary for a more uniform application.
BASIS: Criterion for making determinations regarding quality assurance and seismic analysis was provided in superseded and current TVA design change procedures (references 1 and 3). Section 4.3.1 of TVA Procedure OEP-09 states that nuclear safety-related work includes the specification of quality assurance requirements and applicable industry codes. The seven ECNs identified by the team where quality assurance aspects and applied without corresponding seismic analysis requirements did not conform with these TVA procedures or provide a justification for the omission of seismic analysis.
REFERENCES
- 2. TVA Procedure OEP-ll, Change Control, Rev. O, 4/26/85.
- 3. TVA Procedure OEP-09, Procurement, Rev. O, 4/26/86.
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L e l U6.3-2 (Unresolved Item) Engineering Change Notice (ECN) Quality Assurance and Seismic Analysis Designation TVA's Response I. g&ggg The TVA design control program lacked a clear and consistent definition of when QA and seismic requirements would be designated on the ECN cover sheet.
II. EXTENT TO WHICH THE CONDITION COULD OR DOES EXIST This condition could have existed until June 1985 when this information was no longer designated by a Yes or No entry on the cover sheet when the ECN was initiated, but was included in the USQD after the design was completed.
III. ACTION TO CORRECT EXISTING CONDITION TVA conducted a seismic review of the seven listed ECNs that were found marked QA required but seismic not required which was completed on June 23 , 1986. There were no problems discovered with the subject ECNs as a result of the review. These ECNs are included in the .
representative sample described in TVA's response to D3.1-1.
IV. ACTION REQUIRED TO PREVENT RECURRENCE As a result of Gilbert / Commonwealth (G/C) " Assessment of the Design Control Program for Sequoyah Nuclear Plant" dated October 1985. TVA instituted a modification criteria which establishes a design basis i
for each modification. This new modification criteria which is contained in the SQN Project Manual as an expansion to OEP-6 " Design Input" became effective January 15, 1986.
Also see TVA's response to D3.1-1 relative to seismic design change control.
V. OTHER RELEVANT INFORMATION OR COMMENTS NONE 6
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