ML20207L792
| ML20207L792 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/31/1986 |
| From: | Domer J TENNESSEE VALLEY AUTHORITY |
| To: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 8701120348 | |
| Download: ML20207L792 (23) | |
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p V- .f TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 37401 SN 157B Lookout Place m Jr..t,' 7 pq;pg DEC 311980 U.S. Nuclear Regulatory Conunission Region II ATTN: Dr.'J. Nelson Grace, Regional Administrator j 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
Dear Dr. Grace:
SEQUOYAH NUCLEAR PLANT (SQN) UNITS 1 AND 2 - NRC-0IE REGION II INSPECTION REPORT 50-327/86-27 AND 50-328/86 REVISED RESPONSE TO DEFICIENCIES Enclosure 1 provides our revised response for deficiencies D2.1 's D3.2-3, D3.3-1, D4.3-1, D4.3-3, and D3.3-3. In addition, revised comple". ion dates are provided for items D2.3-1, D5.3-1, and US.3-4. Enclosure 2 provides a ,
listing of the revised commitments associated with this submittal. I As requested in the Design Baseline Verification Program inspection exit meeting on December 4, 1986 at SQN, TVA will submit additional information on items D3.3-5 and U6.3-1. This submittal will be made by January 30, 1987.
If you have any questions, please call 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, l TENNESSEE VALLEY AUTHORITY
. d. A
)J.A.Domer,AssistantDirector Nuclear Safety and Licensing Enclosures cc: .See page 2 i
i 8701120348 861231 l PDR ADOCK 05000327 G PDR
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Dr. J. Nelson Grace DEC 311996 cc (Enclosures):
Mr. James Taylor, Director Office of Inspection and Enforcement 3
U.S. Nuclear Regulatory Comn.lssion Washington, D.C. 20555 Mr. C. G. Zech 4 Director, TVA Projects U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 I
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ENCLOSURE 1 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 Funmanite 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 proccdure requires drilling into points near the center of the bonnet flange, to inject Furmanite into the clearance annull 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 af ter 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).
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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'TVA RESPONSE ITEM: D2.1-1 (Deficiency) - FURMANITE LEAKING VALVE BONNET I. 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 factor to this condition involved weaknesses in the LECN/USQD process in that special design considerations identified in USQDs were not adequately reviewed and implemented 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.
t II. EXTENT TO WHICH THIS CONDITION COULD OCCUR In general, the Furmanite process has existed in the SQN maintenance program in three phases sines 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 components 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.
l 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
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in regard to seismic qualification. A stress analysis was performed for the valve modified by'ECN L6317 and the valve is qualified.
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TVA RESPONSE ITEM: D2.1-1 (Deficiency) - FURMANITE LEAKING VALVE BONNET (Continued)
III. ACTION TO CORRECT EXISTING PROBLEM (Continued)
For the 85 components where the Furmanite process was implemented under MRs (Phase I) or USQDs (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 item 1 above to determine if the component is still in place or has been replaced.
- 3. Perform an evaluation on all components that were not replaced, to determine component failure consequences.
- 4. For all components listed in item 3 above that are determined to be necessary for plant safe shutdown, perform a detailed stress analysis on the component to qualify the component.
The actions described in steps 1 through 4 are complete.
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 rereview of the USQD at the time of LECN closure to ensure that any special requirements and/or design consideratiens 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 associated with the unit 2 system boundaries involved in the prerestart plant walkdowns to ensure that design criteria have been satisfied. The review will be completed before restart. This LECU review ensures that the technical / design requirements of a modification including USQD special requirements have been accomplished by comparison to design criteria requirements. The design criteria requirements have been based on the commitment / requirement data sheets which document TVA's design basis commitments.
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.- : l TVA RESPONSE ITEM: D2.1-1 (Deficiency) - FURMANITE LEAKING VALVE BONNET (Continued)
Safety-related components are not repaired using the Furmanite process on an MR at this time. Additionally, USQDs addressing the Furmanite process are now only issued by 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 have been instructed in the requirement for a stress analysis evaluation when a component is to be repaired by the Furmanite injection process. These personnel were instructed by August 11, 1986.
V. OTHER RELEVANT INFORMATION OR COMMENTS Design Study Request (DSR) No. SO47 (Final Draft 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 March 2,1987.
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TVA RESPONSE ITEM: D2.3-1 (Deficiency) - LONG TERM UNINCORPORATED ECNs and DS.3-1 (Deficiency) - TEMPORARY ALTERATIONS USING TACFs The revised completion date for this deficiency is Januar y 16, 1987. .
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D3.2-3 (Deficiency) PIPINC 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 represented, and that all interface boundaries of higher and lower class piping be pinpointed exactly.
REFERENCES
- 2. TVA Design Criteria No. SQN-DC-V-3.0, The Classification of Piping, Pumps Valves and Vessels, Revision 1, dated June 28, 1985.
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TVA RESPONSE ITEM: D3.2-3 (Deficiency) - PIPING FLOW DIACRAM 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 before the issuance of ECW L5737. The checker accepted as correct the class boundary break shown as Class G on the flow diagram for the CVCS pipe 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. EXTENT 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. Review and revision of the flow diagrams and piping drawings is complete.
V. OTHER RELEVANT INFORMATION OR COMMENTS The deficiency as described in the audit report was a documentation only deficiency. Although the ficw 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 before unit 2 restart.
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D3.3-1 (Deficiency) PIPE SUPPORT FRICTION DESIGN l 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 l account when computing pipe support reactions due to dead and thermal l i loads. Moreover, USAS B31.1-1967, the piping code of record for TVA safety l class B, C and D piping at Sequoyah Nuclear Plant, requires consideration of frictional forces due to piping thermal expansion.
l 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 sama as the direction of the piping l
[ therma'l 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 these elements caused by internal pressure and thermal expansion". Section
( 121.2.1, Anchors and Guides, paragraph (c), specifies that: " Brackets shall i be designed to withstand forces and moments induced by sliding friction in addition to other load."
l f REFERENCES 1
f 1. TVA SQN Pipe Support Design Manual, Vol. 3, Revision 0, dated April 22, 1983.
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- TVA RESPONSE ITEM: D3.3-1 (Deficiency) - PIPE SUPPORT FRICTION DESIGN I. CAUSE 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 were negligible and thus could be neglected. Thus, TVA's SQN pipe support designs did not consider friction loads due to temperature.
II. EXTENT 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.
At Watts Bar Nuclear Plant (WBN) unit 1 and at Browns Ferry Nuclear Plant (BFN), friction loads were not considered in the past but presently are considered in designs; while at Bellefonte Nuclear Plant, friction loads are considered in all designs.
III. ACTION TO CORRECT EXISTING 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, four 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 believed 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. Due to this and 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 of this report. Evaluation will start on January 2, 1987, and will be completed on June 1, 1987.
IV. ACTION TO PREVENT RECURRENCE TVA has issued a design criteria for pipe supports that will require the consideration of friction loads due to temperature for the design of all new pipe suppotts and for consideration in the design of supports that are being modified due to load changes, changes in configuration, etc.
This criteria was issued on June 23, 1986.
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TVA RESPONSE ITEM: D3.3-1 (Deficiency) - PIPE SUPPORT FRICTION DESIGN (Continued)
V. OTHER RELEVANT INFORMATION OR COMMENTS The following areas represent the major systems which are subject to thermal loads and movements:
- 2. Essential Raw Cooling Water
- 3. Chemical and Volume Control
- 4. Component Cooling Water S. Main Steam and Feedwater
- 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 wold stresses, plate-bending stresses, and anchorage pullout loads are within the design basis.
The above information, in conjunction with the information provided from WBU friction investigation, will be used to substantiate that SQU pipa supports are adequate as designed.
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. 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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TVA RESPONSE ITEM: D3.3-3 (Deficiency) - VALVE FUNDAMENTAL FREQUENCY The following provides clarification of our original response to D3.3-3.
TVA models valves performing a safety-related function at SQN as rigid or nonrigid depending on the fundamental frequency of the valve's extended structure. Under TVA's design, valves are rigid when the extended structure has a fundamental frequency of 25 hertz or more (FSAR Table 3.9.2-1 and 3.9.2-3 as revised by Amer.dment No. 3). If the extended structure of a valve has a fundamental frequency of less than 25 hertz, the extended structure is modeled as a flexible cantilever in the piping model. When modifications are made to existing valves which affect the frequency and dynamic response of the valve and piping system, they are evaluated on a case-by-case basis. Before 1972, the procurement seismic requirements of valve specifications were not as explicit as after that date. Even though the same type of valves were procured before 1972, there are some cases where documentation of valve frequency cannot be verified. Any deviation is handled under TVA's program for correcting conditions adverse to quslity. The design of safety-related piping systems (including valves) at SQN is maintained through proper application of the governing criteria aad Rigorous Analysis Handbook.
TVA's original response to this deficiency part III, " Action to Correct Existing problem," referred to the applicability of revised valve frequency to other components. These "other components" are the floor-mounted components discussed in the Deficiency item 3.3-5. Information on this item is provided in TVA's original and upcoming supplemental response te item 3.3-5.
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L 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 interface 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: Secticn 4-l.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
'j 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 2.30 (reference 8) Section 3.4, notes that failure to adhere to
! requirements specifically identified in the USQD evaluation nullifies the USQD j evaluation.
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REFERENCES:
- 2. TVA Calculation 41N730-1, Concrete Steam Generator and Pressurizer Enclosure-Reinf., Rev. 1, 6/29/79 5 3. TVA Calculation 41N730-1, Reactor Building Pressurizer Compartment, Final
- Design, Rev. 2, 8/9/82
- 6. TVA Calculation 10N320, Diesel Generator Building Superstructure and Slabs, Rev. 1, 10/27/80
- 7. TVA Engineering Pcocedure EN DES-EP 4.52 Engineering Change Notices (ECNs) After Licensing-Handling, Revision 1, dated April 24, 1984
{ 8. TVA Engineering Procedure EN DES-EP 2.30, Unreviewed Safety Question 1 Determination-Handling and Preparation, Revision 6, dated April 24, 1984 l 1
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TVA RESPONSE ITEM: D4.3-1 (Deficiency) - EVALUATION OF STRUCTURES FOR REINFORCING BAR CUTS r I. CAUSE 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 inclusion 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 factor to this condition involved weaknesses in the LECN/USQD review process.
II. EXTENT TO WHICH THIS CONDITION COULD OR DOES EXIST 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 TO PREVENT RECURRENCE Policy Memorandum PM 86-04 (DNE) issued April 25, 1986 states that engineering judgment 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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TVA RESPONSE ITEM: D4.3-1 (Deficiency) - EVALUATION OF STRUCTURES FOR REINFORCING BAR CUTS (Continued)
V. OTHER RELEVANT INFORMAIION OR COMMENTS The reinforcing bar cuts were approved, reviewed, and verified by engineering judgment by engineers familiar with the design and TVA design procedures. Any previous reinforcing bar cuts (after 1982) 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. Therefore, this issue does not pose a safety problem for reinforcing bar cuts after 1982.
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).
Reinforcing bar cuts made before 1982 are being addressed through the Employee Concerns Program.
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d 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 s?eam 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 L3034 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 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 failure to adhere to requirements specifically identified in the USQD evaluation nullifies the USQD evaluation.
REFERENCES:
- 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 ITEM: D4.3-3 (Deficiency) - STEAM GENERATOR ACCESS PLATFORM DESIGN I. CAUSE A specific policy regarding documentation of technical justification for engineering judgment did not exist. In addition, as previously discussed in the response to deficiency D2.1-1, a contributing factor to this condition involved weaknesses in the LECN/USQD review process.
II. EXTENT TO WHICH THIS 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 A documentation search has been made for both units to document all known attachments to the steam generator lower support (SGLS). The bulk of attachments found are on the SGLS leg closest to the reactor coolant pumps (members 17 and 18, FSAR figure 5.2.1-5). Some of the systems being supported as attachments on the SCLS are reactor coolant (System 68), steam generator blowdown (15), safety injection (63), and chemical and volume control (62). Approximately 30 supports for lines varying in size from one-inch diameter to three-inch diameter with varioue support configurations distributing forces and moments to the SGLS exist on this particular support leg. According to Westinghouse Report No. SD-119, Volume II-Supports System, this support leg is stressed to a maximum of 52 percent of maximum permissible stress for the upset condition and 23 percent of maximum permissible stress for the faulted condition. (Also see FSAR table 5.2.1-7). Depending on the magnitude of the accumulated upset loads, this condition could be the most critical for this particular leg.
The design calculations which qualifies the SCLS were done entirely by Westinghouse Electric Corporation. The only calculations which TVA has produced for the SGLS were done in 1977 to account for similar attachments to the same support leg. In this analysis the assumptions were conservative and the maximum increase in stress was approximately six ksi. This would have little effect on the interaction formula.
However, the loads applied were not in total agreement with the loads found in the documentation review. This creates uncertainty as to the actual stress level in the SCLS members. Since there are no updated calculations that reflect the present configuration of the SCLS, TVA has asked Westinghouse to perform this calculation. Also, a Problem Identification Report (pIR SQNCEB8661) has been initiated to documrit this concern.
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TVA RESPONSE ITEM: D4.3-3 (Deficiency) - STEAM GENERATOR ACCESS PLATFORM DESIGN (Continued)
III. ACTION TO CORRECT EXISTING CONDITION (Continued)
In addition to the problem of having documented design drawings unaccounted for in the design basis of the SGLS, there are supports and attachments to the SGLS which are not documented on engineering drawings. Based on a cursory walkdown, it is believed that this i
condition is minimal, but it does exist. This requires that a formal l walkdown be used to document such attachments. This is potentially significant since member 26 is already stressed to 100 percent of the a allowable for the faulted load condition. (Reference FSAR figures 5.2.1-6 and 5.2.1-71 In order to determine the current loads on the SGLS, the following steps a're being taken:
- 1. Document all attachments to the SCLS:
- a. By review of available documentation (i.e., design drawings)
- b. By conducting a formal walkdown to verify the documentation and identify any additional attachments.
j 2. Record all attachments with associated loads (i.e., applied forces and moments by loading condition) for inclusion in the reanalysis.
- 3. Westinghouse will revise the structural analysis of reactor coolant loop / supports system report No. SD-119 to include the effects of the attachments.
- 4. Based on results of the reanalysis, initiate an update to the FSAR.
- 5. Ensure that administrative controls are procedurally in place to limit any attachments in the future to the SGLS.
Currently, the documentation search and walkdowns have been accomplished. Based on the above information, completion of this effort will be determined af ter we have reviewed the revised structural analysis from Westinghouse.
Regarding actions to correct the LECN/USQD process, see TVA's response to deficiency D2.1-1.
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TVA RESPONSE ITEM: D4.3-3 (Deficiency) - STEAM GENERATOR ACCESS PLATFORM DESI_GN (Continued)
IV. ACTION REQUIRED TO PREVENT RECURRENCE The Policy Memorandum PM86-04 (DNE) issued April 25, 1986 states that the use of engineering judgment must be accomplished by technical justification. This policy will be reficcted in future revisions to DNE procedures. In addition, instructions were issued to specify requirements for considing 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.
V. OTHER RELEVANT INFORMATION OR COMMENTS 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 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.
The detailed procedure for preparation of diesel generator loading calculations was issued via memorandum dated December 12, 1986. As stated before, this document is applicable to all TVA nuclear plants.
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ENCLOSURE 2
" LIST OF REVISED COMMITMENTS"
( Item # Section Commitment As Currently presented by Revised Response D2.1-1 III Valve modified by ECNL6317 underwent stress analysis and l was qualified.
D2.1-1 IV Review of LECNs associated with-unit 2 system boundaries involved in prerestart plant walkdowns to ensure that design criteria have been satisfied. This review will be completed before restart.
I D2.1-1 IV NEB personnel responsible for the issuance of USQD were instructed in the requirement for Furmanite injection repaired components to undergo stress analysis ,
evaluation. This instruction was completed August 11, 1986.
D2.1-1 V Design Study Request (DSR) No. SO47 results will be incorporated into a general construction specification by March 2, 1987.
D9.2-3 IV Review and revision of the flow diagrams and piping drawings are complete.
D3.3-1 III TVA is going to conduct an evaluation of the SQN support design for the effects of friction loads as discussed in Section V below. Evaluation will start on January 2, 1987, and will be completed on June 1, 1987.
D4.3-3 III TVA to obtain calculations from Westinghouse to reflect present configuration of Steam Generator Lower Supports (SGLS).
Completion of the following actions will be determined af ter receipt and review of structural analysis from -
Westinghouse.
- 1. Document all attachments to the SGLS:
- a. By review of available documentation (i.e.,
design drawings)
- b. By conducting a formal walkdown to verify the documentation and identify any additional attachments.
______________ _ ____ _ /
- 2. Record all attachments with associated loads (i.e.,
applied forces and moments'by loading condition) for inclusion in the reanalysis.
- 3. Westinghouse will revise the structural analysis of reactor coolant loop / supports system report No. SD-119 to include the effects of the attachments.
- 4. Based on results of the reanalysis, initiate an update to the FSAR.
- 5. Ensure the administrative controls are procedurally in place to limit any attachments in the future to the SCLS.
D.2.3-1 -IV The Division of Nuclear Engineering will perform a l l
& DS.3'l-safety evaluation on safety-related TACFs before
{
implementation. The appropriate procedures to implement -
j this review will be revised by January 16, 1987.
US.3-4 IV A detailed procedure for preparation of diesel generator loading calculation will be developed as part of the electrical calculation program. This documt.nt was issued by memorandum dated December 12, 1986, and is applicable to all TVA nuclear plants.
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