IR 05000315/1988028
| ML17328A020 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/28/1989 |
| From: | Dainelson D, James Gavula, Jeffrey Jacobson, Liu W, Westburg R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17328A018 | List: |
| References | |
| 50-315-88-28, 50-316-88-32, IEB-79-14, NUDOCS 8905050273 | |
| Download: ML17328A020 (25) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-315/88028(DRS);
50-316/88032(DRS)
Docket Nos. 50-315; 50-316 Licensee:
Indiana Michigan Power Company 1 Riverside Plaza Columbus, OH 43216 License Nos.
D.C.
Cook Nuclear Power Station, Units 1 and
Inspection At:
D.C.
Cook Site, Bridgman, Michigan American Electric Power Service Corporation, Columbus, Ohio Inspection Conducted:
December 12, 1988, through March 16, 1989 Inspectors:
J.
M. Jacobs J.
A. Gavula Cr'. PM Q.-G.- M R. A. Mestburg W.
C. Liu ((/(
Approved By:
D. H. Danielson, Chief Materials and Processes Section Date Date W (AS/60 Date Z8 8'ate
~/~F 8'ate Ins ection Summar Ins ection on December 12, 1988 throu h March 16, 1989 (Re orts No. 50-315/88028(DRS);
50-3 8 0
DR Areas Ins ected:
Announced special team inspection of engineering activities supporting esign changes and modifications (37700);
licensee action on previous inspection findings (92702); review of licensee event reports (92700);
and review of facility modifications (37701).
The primary focus of the inspection was to assess the overall performance of both the corporate office and site engineering efforts.
Results:
Three apparent violations were identified; multiple examples of inainequate design control - Paragraphs 3.b. (1), 3.b. (6), 3.b(ld), 3.b. (15),
3.b. (16), 3.b. (19), 3.b. (20); lack of procedures
- Paragraph 3.b. (3);
inadequate records of activities affecting quality - Paragraph EXECUTIVE SUMMARY INSPECTION REPORT NO. 50-315/88028; 50-316/88032 D.C.
COOK NUCLEAR POWER STATION, UNITS
AND 2 During the period from December 12, 1988, to March 16, 1989, a four man team of NRC inspectors conducted an Engineering Assessment Inspection at the D.C.
Cook site and American Electric Power Service Corporation offices.
The,intent of this inspection was to evaluate the general areas of design control and design implementation by reviewing the documentation, analyses and as-built configurations for specific modifications.
The engineering are'as reviewed during this inspection were mechanical and electrical components, welding, mechanical systems, and civil/structural.
Licensee Stren ths The engineering staff was technically competent with good depth of experience.
The craft personnel at the plant were well qualified and highly experienced.
Recent reorganization of various engineering disciplines into one nuclear engineering division has the potential to improve the performance of the engineering function.
Licensee Weaknesses The quality of the documentation for the Request for Change packages is considered inadequate in many cases.
The design process utilizes undocumented and unsubstantiated engineering judgements.
The design verification and checking efforts have failed to disclose inappropriate design methodologies and calculational errors.
Design interfaces between engineering at AEPSC and the site are not well defined.
The effectiveness of the controls for the overall design process is questionabl TABLE OF CONTENTS 1.
Persons Contacted 2.
Licensee Action on Previousl Identified Items a.
Violation 316/88012-01 (Closed)
b.
Violation 315/88003-01; 316/88004-01 (Closed)
3.
Review of Plant Modification Pro ram a.
, General b.
Review of Request for Change (RFC)
Packages (1)
RFC No.
(2)
RFC No.
(3)
RFC No.
(4)
RFC No.
(5)
RFC No.
(6)
RFC No.
(7)
RFC No.
(8)
RFC Ho.
(9)
RFC No.
(10)
RFC Ho.
(ll) RFC No.
(12)
RFC No.
(13)
RFC No.
(14)
RFC No.
(15)
RFC Ho.
(16)
RFC No.
(17)
RFC Ho.
(18)
RFC No.
(19)
RFC Ho.
(20)
RFC No.
12-2718 (Violation 315/88028-01A; 316/88032-01A)
12-4054 12-2889 (Violation 315/88028-02; 316/88032-02)
12-4039 12-1795 01-4091 (Violation 315/88028-01B; 316/88032-01B)
12-2999 12-2/43 02-2924 12-2992 01-4087/02-2981 12-2862 12-2213 02-2892 (Violation 315/88028-01CSD; 316/88032-01C8D)
12-2665 (Violation 315/88028-01E; 316/88032-01E)
02-2926 (Violation 315/88028-01F; 316/88032-01F)
(Unreso1 ved Item 315/88028-03; 316/88032-03)
01-2979 01-2977 (Unresolved Item 315/88028-04; 316/88032-04)
02-2907 (Violation 315/88028-01G; 316/88032-01G)
02-2908 (Violation 315/88028-01H; 316/88032-01H)
4.
Desi n Basis Review (Open Item 315/88028-05; 316/88032-05)
5.
Evaluation of En ineerin Documentation (Violation 315/88028-06; 316/88032-06)
6.
Licensee Event Re orts LER 315/89 - 002 (Open)
7.
"Unresolved Items
~,
8.
~0en Items 9.
Exit Interview
DETAILS
'ersons Contacted American Electric Power Service Cor oration (AEPSC)
Indiana Hic i an Power Com an IHPCo
+D.
+J.
+J.
+H.
B.
+T.
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- L
+P.
+F.
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+*K.
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+*Am
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- B
- J
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- p D.
US Nuclear Re ulator Commission (NRC)
Williams, Jr., Senior Executive Yice President, Engineering and Construction Harkowsky, Senior Yice President and Chief Engineer R. Stroyk, Senior Yice President, Engineering and Design P. Alexich, Vice President, Nuclear Operations H. Bennett, Assistant Vice President, Civil Engineering D. Argenta, Assistant Vice President, Nuclear Engineering G. Smith, Jr., Plant Hanager Rutkowski, Asst. Plant Hanager
~
S. Gibson, Asst. Plant Hanager A. Barrett, Director, guality Assurance S. YanPelt, Jr.,
Group Hanager, Design Division L. Williams, Hanager, Hechanical Design Section H. Steinhart, Hanager, Nuclear Plant Engineering Division J. Brewer, Nuclear Safety and Licensing J. Toth, Nuclear Safety and Licensing P. Lauzau, Nuclear Safety and Licensing Kwiatkowski, Design Division, Staff eya Dey, Design Division Staff H. VanGinhoven, Site Design Supervisor A. Svensson, Licensing Coordinator H. Kauffman, Construction Hanager P.
Bei lman, I&C Superintendent R.
Sampson, S&A Superintendent J. Wyckoff, I&C/Hodification Control
'F. Krause, Site Compliance Coordinator
- B. L. Jorgenson, Senior Resident Inspector
+*J. A. Gavula, Reactor Inspector, Hechanical
- J. H. Jacobson, Reactor Inspector, Hetallurgical R. A. Westberg, Reactor Inspector, Electrical
+W. C. Liu, Reactor Inspector, Structural
- Denotes those attending the interim exit meeting at the D.C.
Cook Site on February 3, 1989.
+Denotes those attending the final exit meeting at the AEPSC Offices in Columbus, Ohio on Harch 16, 1989.
The inspectors also contacted other licensee and contractor personnel during the course of the inspectio.
Licensee Action on Previousl Identified Items (92702)
The following violations were reviewed to determine if the full extent of the noncompliances had been determined and whether follow-up actions to correct the present conditions as well as corrective actions to preclude recurrence of similar circumstances had been implemented. 't should be noted that the additional violations cited during this inspection for comparable design control deficiencies could not have been prevented by the corrective action taken for the previous violations, since the engineering work reviewed in this inspection was performed during the same timeframe as the work cited by the previous inspections.
a.
(Closed) Violation (316/88012-01):
Failure to execute adequate design contro in t at a t oroug design review was not performed with respect to the use of alternate weld material.
As discussed in NRC Inspection Report No. 50-316/88019, AEPSC Calculation No.
DCCPV12PW01N was performed to verify the acceptability of the alternate weld material.
See Paragraph 3.b.(15)of this inspection report for additional discussions on this issue.
With regard to the issue of assuring complete, well documented engineering reviews in the future, the licensee effected changes to the applicable AEPSC and D.C.
Cook Site procedures.
In w effort to assess the extent of deviations from the welding procedure, the licensee reviewed over 700 weld records.
This review demonstrated that, while the deviations were not design significant, the quality of documentation was poor.
As a follow-up action to strengthen the design control process, the licensee performed a review of closed Problem Reports that were dispositioned
"use as is."
The purpose of this review was to ensure that.a properly documented engineering evaluation existed to support the closure of the Problem Report.
Out of a total of 37 reports dispositioned
"use as is," five were judged to have insufficient documentation to support the evaluation and disposition.
The licensee corrected the documentation deficiencies and revised the corrective action procedure to ensure adequate documentation of reviews in the future.
The NRC inspector reviewed the above actions and found them acceptable.
This item is considered closed.
b.
(Closed) Violation (315/88003-01; 316/88004-01):
Failure to ensure that a equate design contro measures were app ied to fuse-breaker coordination design changes.
The NRC inspector reviewed NRC Inspection Report No. 315/88003; 316/88004; the licensee's April 29, 1988, response to the violation; the corrective actions implemented by RFC No.
12-2992, Supplement 1,
RFC No. 01-4087 and RFC Ho. 02-2981; and the results of the IHPELL coordination study.
The results of this review were satisfactory.
This item is considered close.
Review of Plant Modification Pro ram (At Site and in Office) (37701)
a.
General The licensee's modification program for the D. C.
Cook plant is described in Instruction No. Pt1I-5040,
"Design Change Control Program."
Design changes are initiated by a "Request for Change"'(RFC)
as detailed in Procedure No.
PHP 5040 MOD.004, "Request for Changes."
The NRC inspector reviewed these documents and found them acceptable.
In an effort to'ssess the extent and significance of a licensee weakness previously identified by the NRC in the area of design control, the NRC inspection team reviewed a sample of documented modifications to the D.
C'.
Cook facility.
This review concentrated on the correctness and thoroughness of engineering design input from both the Columbus and Site engineering organizations, through implementation of that design at the plant.
Engineering efforts reviewed by the NRC inspection team included the mechanical, electrical, and welding disciplines.
b.
Review of=RFC Packa es (En ineerin
)
(1)
RFC No.
12-2718 This RFC was issued to change out the body to bonnet studs from carbon steel to stainless steel on selected valves.
Revision
of this RFC states that "in lieu of AEP design calculations, the respective valve manufacturers would be asked to approve the material change."
The valve manufacturers approved the use of ASTN A194, Grade 6 nuts (proof load 90 KSI).
Subsequent to this approval, the licensee elected to utilize Grade 8F nuts (proof load 48 KSI).
No documented technical basis for suitability of the lower strength material existed.
Based on NRC comments, additional evaluations were performed by the licensee that indicated the lower strength" material was acceptable.
However, failure to execute adequate design control in that, a technical review, including basis, was not performed,nor documented, is an example of a violation of 10 CFR 50, Appendix B, Criterion III (315/88028-01A; 316/88032-01A).
(2)
RFC No.
12-4054 This RFC was developed to modify an equalizing line on the inlet isolation valve for the RHR system.
This modification would permit valve seat leakage testing.
The engineering effort for this RFC was judged to be acceptable by the NRC inspector with the exception that weld size was not specified by design or documented by inspection.
See Paragraph 3.b.(6)
of this inspection report for additional discussion.
(3)
RFC No.
12-2889
This modification consisted of removal of the equalizing lines on the RHR heat exchanger valves, and the addition of flexible hosing and support for instrument lines.
The following design basis documents were reviewed for compliance with licensee commitments and NRC requirements:
Calculation No. DC-D-l-RH-R-0061, Revision 2, March 10, 1988.
This calculation had the following deficiencies:
The 3 inch pipe clamp was not evaluated for adequacy relative to the application of the applied torsional moment.
The source of the dimensional design information for the 3 inch pipe clamp was not referenced.
.
Subsequent evaluations'performed by the licensee in response to NRC questions demonstrated the acceptability of the above pipe clamp.
'Calculation No: DC-D-l-SH-C-534, Revision 1, September 28, (4)
This calculation used a simplified span-by-span analysis method for qualification of the small bore piping modified by this change.
The methodology used in this calculation did-not correspond to fundamental engineering principles with regard to moment combinations.
In addition, the applicability of straight beam formulas to ell shaped configurations was not readily apparent.
Discussions with the lead site design engineer revealed that inherent conservatisms in the method used would compensate for the noted'naccuracies.
Also, the method used was said to be consistent with techniques used at other utilities.
Based on these discussions, the NRC inspector concluded that there was no safety significance associated with this specific instance.
However, the design activities associated with analysis of small bore piping were not prescribed by documented procedures.
This is an example of a violation of 10 CFR 50, Appendix B, Criterion Y (315/88028-02; 316/88032-02).
In addition to the above deficiency, the weld size for the socket welded fittings was not specified.
See Paragraph 3.b.(6) of this inspection report for additional discussion.
RFC No.
12-4039 This modification extended the piping from some ESW drain valves to a point accessible for sample taking.
The NRC inspector found the design work acceptable except that the weld size for the socket welded fitting was not specified.
See Paragraph 3.b.(6)
of this inspection report for additional discussion.
The
procedure for sample taking was also reviewed and found to be acceptably revised.
Field verification of support locations and piping configu'rations by the NRC inspector disclosed no deficiencies.
(5)
RFC No.
12-1795 This RFC was issued for the installation of sample points for the Boron Injection Tank (BIT).
The NRC inspector found the engineering effort to be acceptable with the exception of insufficient welding detail.
See Paragraph 3.b.(6) of this inspection report for additional discussion.
(6)
RFC No. 01-4091 This RFC was developed for the installation of low point drains on the ESW system for control room air handling units.
Though this RFC was not complete at the time of this inspection, several concerns were noted by the NRC inspector.
(a)
The "safety classification" per the site evaluation states,
"involves a Class I system; does not involve a Class II system."
The Safety Evaluation performed by the NSSL group states exactly the opposite.
The NRC inspector understood the evaluation, as performed by NS&L, to be incorrect.
(b)
With regard to seismic qualification, NS8L states that,
"evaluation concludes that seismic analysis is not affected due to the small weight."
The NRC inspector noted that no documented evaluation was available.
This lack of a documented evaluation should be compared with the simple hand calculations contained in RFC 12-2999
.which are considered to be a good example of a minimal evaluation (see Paragraph 3.b.(7) of this inspection report).
(c)
This RFC requires a
1" diameter branch connection to a 3" diameter header.
The engineering effort performed at AEPSC Columbus, however, did not specify the fitting type or connection detail for this branch.
Site engineering then proceeded to specify the fitting type but did not specify the weld reinforcement size.
Furthermore, both Columbus and site engineering failed to perform the calculations required by the ANSI B31. 1 Power Piping Code to determine whether or not piping reinforcement was required.
Due to the size of the branch with respect to the header, this calculation was required before the connection detail could be specified.
Failure to specify fillet weld size (also applies to Paragraph 3.b(2) above), failure to perform required piping connection reinforcement calculations and failure to specify weld reinforcement size (applies to this Paragraph)
are additional examples of a violation of
(7)
CFR 50, Appendix B, Criterion III (315/88028-01B; 316/88032-01B).
RFC No.
12-2999 This RFC was issued for the replacement of reactor incore instrumentation thimble tubes.
Due to an increase in wall thickness, an evaluation for seismic qualification was required.
Simple han'd calculations demonstrated that the weight increase, when considering the total mass used in determining structural loadings, was insignificant (0.5'A).
This is considered a good example of using engineering judgement with minimal supporting documentation.
(8)
RFC No. '12-2743 (9)
(10)
(11)
This RFC replaced the 12" diameter strainer housings at the suction of the containment spray pump with 12",
schedule 40 stainless steel pipe.
Evaluation concluded that due to the significant decrease in weight, the existing seismic supports were adequate.
A hydrostatic test was performed as required by Code and found acceptable by the NRC inspector.
RFC No. 02-2924 This modification was required due to control rod guide tube fastener fai lures.
The design work and safety evaluation for changing from a four bolt design to a three bolt design was performed by Mestinghouse.
The engineering effort was reviewed by the NRC inspector and found to be acceptable.
RFC No. 12-2992, Su lement
This RFC installed fuses that coordinate properly with the Local Shutdown and Indication (LSI) panel breakers.
The NRC inspector reviewed this package to verify that the corrective actions taken for NRC Violation No. 315/88003-01; 316/88004-01 were completed.
The results of this review were acceptable.
RFC No. 01-4087/02-2981 Delete balance of plant (BOP) loads from 250 VDC Essential Service (ESS) distribution cabinets.
The NRC inspector reviewed these packages to verify that the corrective actions taken for NRC Violation No. 315/88003-01; 316/88004-01 were completed.
The results of this review were acceptabl (12)
(13)
(14)
RFC Ho.
12-2862 Change 4KY safety-related motor phase (PJC)
instantaneous relay setting to higher settings to account for additional margin during motor starting.
The NRC inspector's review of the calculations supporting this design change produced one concern.
Review of the PJC settings indicated that several of the motors had high actual pick-up values (current of PJC setting plus excitation current)
and the secondary excitation curves for the current transformers (CTs) indicated that the CTs were saturated (in the flat part of the curve).
The NRC inspector questioned the ability of the PJC to protect the motor when the CT was saturated.
The preparer of the design calculation was consulted and he demonstrated that these higher settings still remain effective for the expected high-grade fault conditions.
The NRC inspector had no further questions.
RFC No.
12-2213 Upgrade performance of current circuits.
The NRC inspector's review of. this design change indicated that this RFC reduced the burden on the saturated CTs associated with the 4KY safety-related motors documented in RFC-12-2862, above.
The HRC inspector found this RFC acceptable.
RFC Ho. 02-2892 This was a relatively minor modification which modified two existing pipe supports and added a third support to correct
"overspan" conditions and a safety interface concern.
This modification addressed Condition Report (C/R) No.
12-11-84-2330.
The supports were designed and installed in 1985.
The following design basis documents were reviewed for compliance with licensee commitments and NRC requirements:
Calculation No. DC-D-12-CS-R936, Revision 1,
Octo er This calculation had the following deficiencies:
The vertical deflection check for Item "D" did not ac'count for the rotation of the vertical member "E".
Although the horizontal deflection check did consider the affect from the applied moment, this consideration was not carried over into the vertical deflection check.
The deflection was underestimated by a factor
of two.
The documented basis for the acceptable deflection was inadequate.
A mathematical error in the deflection calculation in the "X" direction was not noted nor corrected by the checker.
The above deficiencies are additional examples of a violation of 10 CFR 50, Appendix B, Criterion III, in that design verification and checking activities were inadequately performed (315/88028-01C; 316/88032-01C).
As a result of questions asked by the NRC inspector pertaining to the above issue, an additional calculation was performed by AEPSC personnel on February 2, 1989.
The intent of this calculation was to "clarify" the displacement calculation for this support.
Although this document was not necessarily an official calculation, it was checked by AEPSC personnel.
This revised calculation was supposed to address the rotational effects on the calculated displacement.
A review of this "clarification" calculation disclosed that the vertical displacement of the horizontal member due to the moment applied to the vertical member was still not properly accounted for.
The calculation refined the displacement, calculation by determining the axia3 compression of the vertical member.
Although this latter detail was performed correctly, the analysis inaccurately determined the moment displacement which is over 100 times more significant.
This moment displacement is the largest single displacement component and was inappropriately calculated using a linear correlation with the horizontal deflection.
Based on this informati'on, the adequacy of design verification appears to be deficient.
Calculation No. DC-D-1-CS-R936, Revision 0, October
,
85.
During this inspection, an additional Condition Report (C/R) No. 2-12-88-1785, was issued pertaining to this same support.
It was determined that Support No.
12-ACS-R936 was installed on the wrong line.
As stated in the condition report:
"The hanger sketch for hanger No.
12-ACS-R936 provided by the lead engineer was confusing and the installer could not readily determine whether the hanger was to be on line No.
12-CS-26-10 or 12-CS-26-11.
Physical restrictions led the installers to assume that the hanger to be modified was on 12-CS-26-11."
The safety significance associated with this subsequent condition report was considered minimal by the NRC
inspector.
However, failure to correctly translate the design bases into drawings is an additional example of a violation of 10 CFR 50, Appendix B, Criterion III (315/88028-01D; 316/88032-01D).
(15)
RFC No. DC-12-2665 This was an extensive modification which added a 4 inch cross-tie piping between the Unit 1 and Unit 2 CVCS systems.
The initial work was started in late 1983 and the project was completed in mid 1987.
The following design basis documents were reviewed for compliance with licensee commitments and NRC requirements.
Calculation No. DC-D-02-CS/SI-2107 This calculation had the following deficiencies:
On Sheet 8 of 8 "Summary of Loads," the interfacing calculation was revised.
The "X" moment for seismic should have been noted as being increased to 910 ft-lbs instead of 878 ft-lbs.
At Node 130, the stress intensification factor (SIF)
was erroneously entered as 1.0.
The correct SIF should have accounted for a fillet welded attachment and had a 1.3 SIF assigned to this node.
The pipe has lugs attached to it 15 inches above the elbow which is below Node 130.
These lugs were not accounted for in the analysis nor were they evaluated in the support calculation.
Calculation Ho. DC-D-12-CS/SI-2108 This calculation had the following deficiencies:
On "Summary of Loads" sheet, data for Node 2 (R-910)
was superseded by another analysis and this was not noted in the calculation.
The "Z" Moment for seismic load at R928 should have been 1228 ft-lbs, not 1218 ft-lbs.
Drawing No. DC-D-12-CS/SI-2108 showing the finite element piping model was revised to account for a new piping configuration.
No notation was made that the original configuration had been superseded.
The location of Support No.
12-ACS-L912 was entered as 2'-6" from the adjacent elbow in the piping stress analysis.
The support drawing that was issued to the field specified the support location as 6'-4" from
the adjacent elbow.
The locational error of 3'-10" was not caught during a checking process for the drawing.
The licensee's evaluation of these discrepancies concluded that there was no associated safety significance.
Calculation No. DC-D-'12-CS-L916, Revision 1,
prl p
o This calculation had the following deficiencies:
The vertical and horizontal load directions were interchanged in the calculation.
The 1/2 inch bars restraining the horizontal loads were not evaluated.
The 3/16 inch fillet welds attaching the above bars were not evaluated.
Sources for allowable loads for bolts were not referenced.
The bending moment resulting from application of the horizontal load was not included in the evaluation of the structural angle, the base plate or the anchor bolt.
Calculation No.
12-ACS-L916, Revision 1,
Decem er This calculation had the following deficiency:
No formula or reference to a formula was given for calculating the deflection due to the moment applied at the end of the cantilevered support.
Calculation No.
DCCPV12PWOlN, Revision 0, April 12, 1988.
j This calculation had the following deficiencies:
The source of "maximum allowable stress in material" was incorrectly stated.
The formula for longitudinal pressure stress was incorrectly stated.
The calculation concluded that the wall thickness at the weld was acceptable based on a weld crown of 0.063 inches.
The basis for this value was never established in the calculation.
Calculation No.
DCCPV12PWOlN, Revision 2, July 1, 1988.
This revised calculation was performed in response to the NRC concerns listed above.
Several references or portions'f references were provided with this calculation.
In reviewing these references, several discrepancies were noted by the NRC inspector.
For Reference 3, Calculation No. DC-D-Ol-CS/SI-2106, the stress value used to show Code compliance with the design basis earthquake condition was not the maximum stress value combination and resulted in understating this value by at least 10%.
For Reference 5,
AEPSC Specification DCCPV104(CS, the design pressure for the CVCS Crosstie is 75 psig lower than the Centrifugal Charging Pump Discharge pressure.
The licensee's evaluation of the above deficiencies concluded that they were not safety significant.
However, the deficiencies are additional examples of a violtaion of 10 CFR 50, Appendix B, Criterion III in that the design bases were not correctly translated into drawings and design verification and checking activities were not performed adequately (315/88028-01E; 316/88028-01E
).
( 16)
RFC No. DC-02-2926 This modification to replace the Chemical and Volume Control System containment isolation check valve started in 1986 and was completed in early 1987.
As a result of the relocation of the valve, several pipe support modifications were required.
The following design basis documents were reviewed for compliance with licensee commitments and NRC requirements:
Calculation No. DC-D-R590, Revision 1, January 25, 1988.
This calculation had the following deficiency:
The intent of the analysis was to qualify a 1/2" x 2" bent plate to replace existing U-bolts for six 3" piping supports.
The side load for the U-bolts exceeded the manufacturer's rated load.
The analyst chose to analyze the plate using a computer program specifically intended for piping analysis.
As a result, the analyst had to approximate the rectangular shape with a circular section.
The analyst accomplished this by equating the section moduli instead of the moments of inertia for the two sections.
By doing this, the analyst failed to realize that the stiffness I
matrix for the finite element is based on the moment of inertia and not the section modulus.
This discrepancy was not evaluated during design verification activities for this calculation.
Based on the NRC inspector's concern, additional analyses were performed by the licensee.
An additional analysis using the piping computer code indicated that the original analysis had underestimated the moments by approximately 14/.
A separate analysis using a computer code with prismatic beam elements was also performed.
The results from this analysis gave an indication that the original
'nalysis using the pipe elements was conservative by approximately 3'/.
Although no additional verification was performed by the NRC, the conservative nature of the
. original analysis may be attributable to the flexibility characteristic associated with the piping element.
However, the degree of conservatism associated with the piping element was not definable.
It should be noted that a review of the prismatic beam analysis showed that the load combination used in the original analysis did not give conservative moment results.
The highest bending moment and shear load occur during the application of the lateral force only and will result in a 7% increase in applied moment.
This will apparently not affect the qualification of the bent plate.
However, the design verification activity also failed to disclose this deficiency.
The above deficiencies are additional examples of a violation of 10 CFR 50, Appendix B, Criterion III, in that design verification activities were not performed adequately (315/88028-01F; 316/88032-01F
).
By comparing the original analysis with the analysis performed for this RFC, a significant difference is noted in the support loads.
The new loads are in some cases one-third of the original-loads.
On this basis, if the existing U-bolts were overstressed with the new loads, then they would have been three times as bad with the old loads.
The overall adequacy of the original supports is now questionable as well as the assumption that if the new loads are less than the old loads that the support is currently adequate.
Pending additional reviews by the licensee, this issue will be considered.
an Unresolved Item (315/88028-03; 316/88032-03).,
(17)
RFC No. 01-2979 This modification addressed pipe support discrepancies identified in Problem Report PR-87-696.
During the review
of the associated documentation, the NRC inspector had two concerns.
First, the documentation associated with the problem report did not discuss the actual discrepancy which eventually was corrected by this RFC.
The discrepancy was discovered after doing additional investigation into the problem report.
However, a new problem report or a revision to the existing problem report was not generated.
By not properly documenting the discrepancy, trending evaluations will not be accurate.
Second, the discrepancy identified in the problem report should. have been found and addressed during the work associated with IE Bulletin 79-14.
The IE Bulletin 79-14 documentation provided to the NRC
" inspector did not show that this discrepancy was previously noted.
This situation implicates the adequacy of the IE Bulletin 79-14 program and is considered part of Unresolved Item 315/88028-03; 316/88032-03 as discussed in Paragraph 3.b.( 16) above.
(18)
RFC No. 01-2977 This modification addressed the inadequately supported and overstressed small bore piping identified in-Problem Report PR-87-552.
The RFC required the installation of four new pipe supports and the modification of two existing pipe supports in order to eliminate the identified seismic inadequacies.
These inadequacies were found on two of the four reactor coolant pump seal leak off lines.
The following documentation was reviewed for compliance with licensee commitments and NRC requirements:
Calculation No.
1987.
Calculation No.
1987.
Calcu'lation No.
1987.
Calculation No.
1987.
Calculation No.
1987.
DC-D-Ol-CS-2175, Revision 0, August 14, DC-D-Ol-CS-R948, Revision 0, August 14, DC-D-01-CS-R949, Revision 0, August 14, DC-D-Ol-CS-R950, Revision 0, August 14, DC-D-01-CS-R951,, Revision 0, August 14, The NRC inspector had no adverse technical comments relative to the above calculations.
However, in reviewing the cause of the problem, the NRC inspector had additional concerns.
The description of the cause listed in the problem report stated in part, ".
an error was apparently made during the design of the.
Unit 1 systems in question."
This design error not only caused the FSAR stress limit to be exceeded, but also caused the operability limit of the piping to be exceeded as well.
Even though the system was declared inoperable, the licensee's safety evaluation concluded that cooling water to the reactor coolant pump seals coul'd have been
. maintained and would not have led to an unsafe condition.
However, this situation in conjunction with similar situations discussed in Paragraphs 3.b.(19)
and 3.b.(20),
causes the adequacy. of the original small bore design and implementation efforts to be questioned.
Pending additional reviews by the licensee -into the above situation, this is considered an Unresolved Item (315/88028-04; 316/88032-04).
See Paragraph 3.b.( 19), 3.b.(20),
and
for additional discussion.
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RFC No. 02-2907 This modification required the addition of four small bore pipe supports to eliminate a stress problem'n the glycol piping system.
The overstress problem was discovered during a seismic review of the associated piping.
Apparently, no problem report was initiated as a result of this discovery.
The following documentation was reviewed for compliance with licensee commitments and NRC requirements:-
Calculation No.
1986.
Calculation No.
1986.
Calculation No.
1986.
Calculation No.
1986.
Calculation No.
1986.
Calculation No.
1986 Calculation No.
1986.
DC-D-02-AR-R-0001, Revision 0, January 3, DC-D-02-AR-R-0002, Revision 0, January 3, DC-D-02-AR-R-0003, Revision 0, January 30, DC-D-02-AR-R-0004, Revision 0, January 30, DC-D-02-GR-L-1054, Revision DC-D-02-R-2141, Revision 0, 0, February 10, February 27, DC-D-02-R-2147,'evision 0, February'7,
"3, Deficiencies identified with the above documents are as follows:
For support R-001, the X-Moment was calculated inaccurately and resulted in'
10'A nonconservative error.
U-bolt capacities were not mentioned in support ca 1 cul at ions.'4
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U-bolt materials were not specified on the drawings and instead were noted as being determined by the field.
The sources of the valve weights were not specified.
Piping stress summa ies did not reference a specific computer output as the source of the information.
For piping R-2141, the pressure stress was incorrectly stated for the DBE load case.
Although none of the above deficiencies were considered safety significant, they are considered additional examples of a violation of 10 CFR 50, Appendix B, Criterion III, in that design verification and checking activities were not
.
adequately performed (315/88028-01G; 316/88032-01G).
As a result of the initial overstressed condition, the NRC inspector had additional concerns regarding the adequacy of the original small bore design and implementation efforts.
This aspect is considered part of Unresolved Item 315/88028-04; 316/88032-04 as discussed in Paragraph 3.b.( 18).
RFC No. 02-2908
/
This modification is associated with the steam generator replacement project.
The NRC reviews during this inspection were strictly limited to instrument piping supports and isometric drawings.
No calculational reviews were performed and very limited time was spent.
Initially a problem was identified by the field concerning the clearances in the U-bolts used to support the steam generator instrumentation piping.
Inconsistencies in the installed clearances were noted by operational personnel and subsequent evaluations concluded that the installed clearances exceeded analytically justifiable tolerances.
The U-bolt clearances had not been specified on the support drawings.
This is an additional example of a violation of 10 CFR 50, Appendix B, Criterion III in that the design bases were not translated into the
. design drawings (316/88028-01H; 316/88032-01H).
In addition to the above deficiency, the NRC inspector had other concerns regarding the adequacy of the initial small bore design and implementation efforts.
During the initial evaluations of the as-built instrumentation line configuration by the mechanical design group, additional supports were required in order to meet stress criteria.
These evaluations were performed using computer analyses and as such were much more accurate than the presupposed conservative method used during original construction.
Although the initial conservatisms were reduced with this
more accurate approach, additional supports were required.
This situation further questions the adequacy of the original small bore efforts.
This aspect is considered part of Unresolved Item 315/88024-04; 316/88032-04 as discussed in Paragraph 3.b.( 18) above.,
4.
Desi n Basis Review (37700)
Safety System Functional Inspections (SSFIs)
and Safety Systems Outage Modification Inspections (SSONIs)
conducted by the NRC have demonstrated that many licensees have not maintained an adequate design basis.
Recent initiatives taken by AEPSC to perform their own SSFIs also identified, as a generic weakness, the lack of'
well documented plant design basis.
The purpose of this review was to assess the progr'ess made in correcting this concern.
The NRC inspector interviewed key personnel responsible for establishing design basis documentation requirements.
These interviews indicated that the licensee had formed a task force and held (Ad Hoc) committee meetings to address the design basis concern.
The task force had proposed the following actions:
a.
Establish the design basis based on the
CFR 50.2 definition.
b.
Identify the documents that contain the design basis.
c.
Define the program and where and how to use it (roadmap).-
d.
Investigate the effect of how the design basis interacts with the RFC program.
e.
Revise Procedure No.
GP 3. 1 "Design Changes,"
to indicate how the cognizant/lead engineers treat design basis material and how to document design basis material.
f.
Update present system descriptions and create new system descriptions, as required.
The NRC inspector reviewed the changes to Procedure No.
GP 3. 1, the system description foi the Auxiliary Feedwater System No.
SD DCC-HP109, Revision 6, and the Electrical Generation Section policy for structured design basis review.
The NRC inspector concluded that significant progress had been made relative to establishing the D.C.
Cook design basis documentation.
As a result of the concerns identified in Paragraphs 3.b.( 18), 3.b.(19),
and 3.b.(20)
above, the design basis documentation associated with the
"Alternate Piping Analysis Criteria" was reviewed by the NRC inspector.
This original criteria, dated September 1971, was used to design and install most of the small bore piping at the plant.
In reviewing this criteria, the NRC inspector had several concerns regarding its assumptions and applicability. 'lthough there were specific statements requiring the support of concentrated loads,
there appeared to be a certain amount of flexibilityleft to the design engineer.
The comments for "Special Situation" as well as the "Criteria Checklist" gave specific factors to be applied for supports near unsupported concentrated loads but the effects on pipe stress are left to "engineering judgement."
In addition, there is a lack of direction pertaining to eccentrically applied loads (i.e.,
loads due to valves with massive operators).
This aspect may be partially responsible for the current modifications needed for Copes-Vulcan Valves, although it is recognized that the original discrepant design input is the predominant cause.
However, since the center of gravity location was evidently never obtained by the licensee, there was an apparent lack of sensitivity originally associated with this aspect.
This is considered part of Unresolved Item 315/88028-04; 316/88032-04 discussed in Paragraph 3.b.( 18)
above.
Additional concerns related to the small bore piping program were also raised during field walkdowns by the NRC inspectors at the site.
In one case a 1/2 inch test line was questioned relative to the length of the cantilevered section containing the shut-off valves.
Subsequent evaluations by the licensee determined that the stresses were well within Code allowables.
However, there did not appear to be any straight forward guidance specified in the original small bore procedure regarding acceptable cantilever pipe lengths.
V Of more significance was the concern regarding a field fabricated lateral on the Post Accident Sampling Line.
This was 1 inch tubing with a lateral intersecting at 22'.
Documentation reviews by the licensee did not locate any evaluation of this design.
A subsequent analysis performed by the licensee concluded that for pressure loading, the connection would be adequate if the Code suggested corrosion allowance was neglected.
However, since the Code design formulas are restricted to laterals between 45'nd 90', additional requirements are specified regarding this specially designed component.
In addition to the pressure calculations performed, the Code requires that the design be verified by either experimental stress analysis, proof testing or extensive and successful performance experience.
Since this was not done, the Licensee has committed to repla'ce the connection in question with a standard fitting.
Based on the above discussion, this is an additional concern regarding the adequacy of the original small bore effort and is considered part of Unresolved Item 315/88028-04; 316/88032-04 discussed in Paragrah 3.b.( 18)
above.
During the reviews of RFC 12-4054, 12-,2889, 12-4039, 12-1795, and 01-4091 as previously discussed, it was noted that the fillet weld size for socket weld fittings was not specified during the design process.
Discussions with guality Control personnel revealed that the inspection procedures used editions of ASIDE or B31. 1 Codes which required fillet weld sizes to be 1.09 times the nominal pipe wall.
This is inconsistent with the size specified in the original code of construction which is 1.25 times the nominal pipe wall.
Although this is not considered a safety significant issue, the apparent discrepancy between what is specified in the FSAR and what is currently being used, needs to be resolved.
Pending the
~s
, ~
'.
7.
licensee's decision on how to resolve the apparent design basis discrepancy, this is considered an Open Item (315/88028-05; 316/88032-05).
Evaluation of En ineerin Documentation (37700)
In general, the quality of the mechanical RFC documentation packages is considered inadequate.
The packages are fragmented, lack appropriate references, (e.g.,
seismic calculation packages, NDE procedures),
and do not contain adequate documentation to support engineering judgements.
With respect to the inspection of piping fillet weld size, the NDE procedure requires a determination of wall thickness and a simple hand calculation.
Due to lack of documentation, it cannot be verified that this was done correctly.
This lack of documentation coupled with the lack of up front weld size specification makes it impossible to determine from records, the installed weld size.
I Failure to maintain sufficient records to furnish evidence of activities affecting quality is an example of a violation of 10 CFR 50, Appendix B, Criterion XVII (315/88028-06; 316/88032-06).
Licensee Event Re orts (LERs)
(92700)
(0 en)
LER (315/89-002):
Non-Ser vice Induced Deformation of Emergency Gore Coo ing System uction Line Seismic Restraint.
During routine ISI support inspections by the licensee, pipe support 2-GSI-L102 was noted as being damaged.
An evaluation of the 24-inch line indicated that the suction line stresses would have exceeded FSAR criteria during a design basis seismic event with this damaged support.
Additional walkdowns of two adjacent supports revealed that although ther e was no apparent damage, the as-installed configuration was not in accordance with design documents.
The licensee concluded that the damage,to the initial support was not service induced.
This was based on additional walkdown information and engineering judgement.
Pending the review of the bases for concluding that this was not service induced damage this LER will remain open.
The cause of the event, according to the LER, was attributed to startup activities in 1977.
If the damage was due to these activities, then it was not apparent to the NRC inspector how the IE Bulletin 79-14 walkdowns would not have disclosed the significant clearance discrepancy.
In addition, the other two supports which were noted as not being in accordance with plant design also challenge the adequacy of the IE Bulletin 79-14 program.
This aspect is considered part of the Unresolved Item (315/88028-03; 316/88032-03)
discussed in Paragraph 3.b.(16)
above.
Unresolved Items An unresolved item is a matter about which more information is required in order to ascertain whether it is an acceptable item, an open item, a
deviation or a violation.
Unresolved items disclosed during this inspection are discussed in Paragraphs 3.b.( 16), 3.b.( 17), 3.b.( 18),
3.b.(19), 3.b.(20),
4 and 6.
8.
~0ee Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the HRC or licensee or both.
Open items disclosed during this inspection are discussed in Paragraph 4.
9.
Exit Interview (30703)
The Region III inspectors met with the licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on March 16, 1989.
The inspectors summarized the purpose and findings of the inspection.
The licensee representatives acknowledged this informa'tion.
The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed during the inspection.
The licensee representatives did not'dentify any such documents/processes as proprietary.
19