ML20215F871
| ML20215F871 | |
| Person / Time | |
|---|---|
| Issue date: | 10/06/1986 |
| From: | Correia R, Jocelyn Craig NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20215F859 | List: |
| References | |
| REF-QA-99900400 NUDOCS 8610160403 | |
| Download: ML20215F871 (17) | |
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ORGANIZATION: BABC0CK & WILC0X LYNCHBURG, VIRGINIA INSPECTION INSPECTION REPORT OATES: 6/23-27/86 JN-SITE HOURS: 144 N0.: 99900400/86-01 CORRESPONDENCE ADDRESS:
Babcock & Wilcox, a. McDermott Company Nuclear Power Division ATTN: C. W. Pryor, Vice President and General Manager Post Office Box 1260 Lynchburg, Virginia 24506-0935 ORGANIZATIONAL CONTACT: T. Stevens, Manager, Quality Assurance TELEPHONE NUMBER: (8041385-3138 NUCLEAR INDUSTRY ACTIVITY: Design and engineering services for B&W plants requesting reanalysis and modifications to existing systems, components and structures.
~ .
ASSIGNED INSPECTOR: (c m R. P. Corrtia, Special Projects Inspection m- 2-dc Date Section(SPIS)
OTHER INSPECTOR (S): P. D. Milano, SPIS K. C. Leu, SPIS
' O. Ambrosek, Consultant APPROVED BY: s ' cu r /6dk.
'Da te
/ John W. Craig, Chief, SPIS; Vendor Program Branch INSPECTION BASES AND SCOPE:
A. BASES: 10 CFR Part 21,10 CFR Part 50, Appendix B.
B. SCOPE: (1) Review the status of previous inspection findings, T2TTnspect the design and engineering activities performed by B&W for plant modifications.
PLANT SITE APPLICABILITY: Arkansas 1(50-313), Belefonte 1 & 2 (50-438 &
439), Crystal River 3 (50-302), Davis-Besse 1 (50-346), Oconee 1, 2, & 3 (50-269, 270 & 287), Rancho Seco 1(50-312), Three Mile Island 1,(50-289),
Washington Nuclear 1(50-460),
t 8610160403 861003 PDR GA999 EMVDW 99900400 PDR
ORGANIZATION: BABC0CK & WILC0X LYNCHBURG, VIRGINIA REPORT INSPECTION RESULTS: PAGE 2 of 10 NO.: 99900400/86-01 A. VIOLATIONS:
None.
B. NONCONFORMANCES:
- 1. Contrary to Criterion III of 10 CFR 50, Appendix B and B&W Administrative Manual Procedure NPG-0402-01,Section VI, A.7 and A.1, the reported results of reactor power level and DNB power level in calculation No. 32-1158579-00 for PSC 17-83 could not be independently verified. (86-01-01)
- 2. Contrary to Criterion XVI of 10 CFR 50, Appendix B and B&W Administrative Manual Procedure NPG-1717-02,Section VIII, 1, calculation No. 32-1158579-00 (PSC 17-83) ' contained recognized code deficiencies, (i.e., inadequate feedwater flow response with best estiniate gains and , code instabilities during transient evaluations) which were not reported and resolved as required by
- the procedure. (86-01-02)
- 3. Contrary to B&W Administrative Manual Procedure, NPG-0903-03, Appendix 1, an overcooling transient is not identified as an analysis performed by computer code Digital Power Train (DPT).
(86-01-03)
- 4. Contrary to Criterion III of 10 CFR 50, Appendix B and B&W Administrative Manual Procedure NPG-0402-01,Section VI, A.1, calculation No. 32-1163870-00, "Radcal Gamma Thermometer Cable Restraint Stresses" was not demonstrated to be technically accurate and complete, and cid not contain clear and concise results.
(86-01-04)
C. UNRESOLVED ITEMS:
None.
D. STATUS OF PREVIOUS INSPECTION FINDINGS:
- 1. (0 pen) Unresolved Items: PSC 17-83, Ov.ercooling Events at Low Reactor Power A review of " Unresolved Items," Report No. 99900400/85-01 for Potential Safety Concern (PSC) 17-83 " Overcooling Events at low Reactor Power" was conducted. These steam generator secondary side transients could produce a relatively high peak power without a reactor trip when initiated from low power levels.
ORGANIZATION: BABC0CK & WILC0X LYNCHBURG, VIRGINIA REPORT INSPECTION RESULTS: PAGE 3 of 10 N0.: 99900400/86-01 The review was initiated with a package that contained the PSC close-out notice with the evaluation report at.d a draf t letter to the effected utilities as attachments. The evaluation report stated analyses with computer code Digital Power Train (DPT) showed that core power level did not exceed 70% during the transient. The source of this statement was not referenced. A statement was made that departure from nucleate boiling (DNB) is not expected to occur if the core power level does not exceed 80% full power. The source of this statement was not referenced.
The file for PSC 17-83 was reviewed. This file contained a copy of the letters to the utilities and a letter from R. W. Moore to D. Mars which referenced calculation No. 32-1158579-00. This letter also contained a comment which noted a personal conversation with R. A.
Kochendarfer, Fuels Engineering, the conclusion, of which, was that overfeeding transients should not result in DN8 if the core power level does not exceed 80% ful.1 power. This conclusion was not supported by adeouate documentation or references and was therefore -
not independently verifiable.
The operating manual for DPT was reviewed. The manual did not include an overcooling transient in its list of system analysis and scoping studies for anticipated transients as required by B&W NPG-0903-03, Appendix 1.
Discussions were held with an engineering supervisor concerning the applicability of the single point kinetics method of analysis to this type of transient. The supervisor could not provide documented direct applicability of DPT to overcooling transients, but did provide as an example, topical reports for the TRAP 2 ead CADDS computer codes which showed the single point kinetic method to be acceptable for an over-cooling event.
The NRC inspector reviewed calculation No. 32-1158579-00. The base model was identified, changes that had been made were recorded, and steady state analysis with modifications to demonstrate stability were provided. Results for the transient cases were included as-well-as a discussion of results as compared to the simulator results.
The graphic results were difficult to verify since the Y-axis was identified only with an acronym.
The calculation's preliminary results had instabilities which were identified as computer code instabilities. A modification was made to the model and the results obtained provided the 70% full reactor power limit when the transient ended. This was the value stated in
ORGANIZATION: BABC0CK & WILC0X LYNCHBURG, VIRGINIA REPORT INSPECTION RESULTS: PAGE 4 of 10 NO.: 99900400/86-01 a report to the B&W NSSS utilities. This computer run also had severe instabilities identified on some of the plots. The analyzed transient was terminated by a reactor trip due to these instabilities.
These instabilities were due to the analysis reaching the point where feedwater was being drawn back through the aspirator ports of the steam generator and thus tripping the reactor.
Based on trends indicated on graphs of DPT outputs in the steam generator (SG) downcomer levels and reactor power level, .it was not possible to independently conclude that 70% full power was a reasonable or conservative estimate of potential power level which would be reached before feedwater flow would be stopped by a control system as a result of water level reaching the aspirator port of B&W steam generators.
The technical justification for the reported results were discussed with the engineering supervisor. He consulted with several individuals and concluded that DPT code instabilities were caused by feedwater '
being drawn back through the aspirator ports of the steam generators.
There was no documented technical evaluation to justify the conclusions reached from the analysis performed and subsequently transmitted to the B&W utilities.
Nonconformances 86-01-01, 02, and 03 were identified in this area of the inspection.
- 2. (Closed) Nonconformance (83-03): This nonconformance involved computer certification files for the CRAFT 2 codes that were reviewed by the supervisor of the originator.
A review of B&W procedure NPG-0403-11, " Technical Document Signatures",
Rev.13, dated April 1,1986 indicated that the revised procedure includes a definition of the meaning and responsibilities related to signatures on technical documents. Unit managers are required to document their decisions that an independent review was properly independent, and to place their initials adjacent to the independent reviewer's signature on technical documents. This nonconformance is considered closed.
In discussions with B&W Quality Assurance (CA) personnel, instances were cited that the requirements in the revised procedure had not been consistently met and there had been noncompliances with other technical documents such as specifications, drawing, EIRs, CI/As, and calcula-tions. B&W issued a memo dated May 14, 1986 to its business unit CA representotives requesting investigations as to the scope of the
ORGANIZATION: BABC0CK & WILC0X LYNCHBURG, VIRGINIA REPORT INSPECTION RESULTS: PAGE 5 of 10 NO.: 99900400/86-01 problem and needed corrective and preventive measures by June 13, 1986.
Based on the above, a review of the implementation of the B&W proce-dure NPG-403-11 may be the subject of a future inspection.
- 3. (Closed) Nonconformance (84-03): This nonconformance involved an uncertified computer code (the CORE code) that was used in a safety-related analysis. The analysis has been independently verified and the computer code has been independently certified.
The NRC inspector reviewed B&W QA procedure NPG-902-06, " Computer Program Development and Certification," Rev.12, dated May 1,1985.
The revised procedure establishes the requirements and responsibil-ities for developing a certified computer program that is used to perform safety related calculations. This nonconformance is considered closed.
4 (Closed) Nonconformance (84-03): This item involved the description of computer code limitations in computer program manuals. ,
B&W CA procedure NPG-903-03 " Development of Certified Computer Program User Manuals", Rev. 11 dated March 1, 1985 includes the identification and description of program limitations that affect the validity and/or accuracy of the programs. The subject title has been changed from " Development and Control of Computer Program Manual" i
to " Development of Certified Computer Program User Manual." This
, nonconformance is considered closed.
- 5. (Closed) Nor.conformance (84-03): This nonconformance, irvolved the ASME code indices used in the T3 PIPE computer program for the calcu-l j
lations related to butt welded fittings which use branch connections in the piping analyses and the prorsr verificatien of program file calculations.
"Re:olution of T3 PIPE Errcrs" dated January 24, 1986, File No.
2A4/(T3 PIPE 18/2) was reviewed with respect to a general review of the certification of "T3 PIPE" program and the associated corrective actions. The NRC inspector observed that the "T3 PIPE" version has incorporated ASTM code indices. In addition, B&W has reviewed all 37 identified pipe analysis documents on the Historical Document List (HDL), and Document Release Notices (CRN's) in the file in the Stress Analysis Unit (SAU), and concluded, that the errors involved have no safety-related effects. This nonconformance is considered closed.
4
ORGANIZATION: BABC0CK & WILC0X LYNCHBURG, VIRGINIA REPORT INSPECTION RESULTS: PAGE 6 of 10 NO.: 99900400/86-01 E. Other Findings and Comments
- 1. Oconee Pressurizer Internal Spray Piping Analyses The B&W analysis of the Oconee pressurizer internal spray piping which was performed as a result of the addition of Target Rock Valves and the modified cooldown transient were reviewed. B&W calculation No. 32-1163920-00, " Pressurizer Internal Spray Piping Analysis" dated June 3,1986, analyzed the internal pressurizer spray piping and a pipe support for all applicable loads. The calculation indicated that the spray-head and internal pipe meet the stress criteria of the applicable design code. However, the U-bolt, bolting plate, angle hanger and attachment welds exceeded the allowable stress valves for the applicable design code.
B&W's supervising engineer indicated that the calculations contained preliminary results based on .very conservative assumptions and B&W would look into the problem further with refined computer modeling -
boundary conditions and weld analysis. Duke Power Company, the Oconee licensee, had been notified of the problem and was to meet with B&W to discuss the approach to the problem.
B&W stated that they would notify the NRC of the final results as soon as the reanalysis was completed.
/ 2. Limitorque Valve Actuator Weight Discrepancies TVA reported a 10 CFR Part 21 defect in December,1984 which evolved when an evaluation of valve and valve actuator seismic analyses for the pipe in the Bellefonte nuclear plant identified actuator weight discrepancies between pipe stress analyses values and values contained in vendor drawings for Limitorque valve actuators. B&W identified this potential defect as Potential Safety Concern (PSC) 4-85. At a meeting with B&W all valve ve.: dors involved were invited to discuss the problem and follow-up actions. Limitorque, Anchor-Darling, Copes-Vulcon and Rockwell manufacturers had supplied valves for Bellefonte. Additional meetings were held between B&W and each valve vendor to discuss specifics of the discrepancies. Further analyses showed that Copes-Vulcan valve weights were actually less than calcu-lated values and B&W, Limitorque, and Copes-Vulcan, stated that existing seismic analyses were adequate and no reanalyzes would be required.
ORGANIZATION: BABC0CK & WILC0X LYNCHBURG, VIRGINIA REPORT INSPECTION RESULTS: PAGE 7 of 10 NO.: 99900400/86-01 PSC 4-85 was still open as of the date of the inspection. A letter to the B&W Owners Group (BWOG) was reviewed by the NRC inspector which indicated that no problem existed for B&W 205 Fuel Assembly (FA) plants: Bellefonte and WPPSS-1. Also, the letter indicated that the same problem may potentially exist for B&W 177 FA plants. However, since 177 FA plants had different requirements on valve qualifications, and different valve vendors than the 205 FA plants, B&W could not assess the impact of the valve actuator weight discrepancies for 177 FA plants. B&W stated their belief that licensees of 177 FA. plants should investigate the problem on their own. B&W made a random review of 177 FA plant records and found the same type of weight discrepancies No further evaluation or analyses has been performed by B&W. B&W stated that additional evaluation will not be performed unless requested by the utilities.
B&W personnel noted that while Regulatory Guide 1.48 recommends assur-ance of valve operability under seismic conditions for safety-related valves, 177 FA plants had already been constructed or were in the -
final stages of construction prior to the issuance of Regulatory Guide 1.48,
- 3. Radcal Gamma Thermometer - ANO-1 The NRC inspector reviewed calculation 32-1163870-00 "Radcal Gamma Thermometer Cable Restraint Stresses" and two associated drawings Nos.116347?;-2, " Service Structure Tie Plate Assembly," and 11611344E-2, " Service Structure Tie Plate Details." A comparison of calculation data and results to drawing details was performed to determine completeness, detail and accuracy of the modification. The following was found as a result of the review:
- a. Calculation page numbers 4-6 contained a statement concerning deflections. There was no reference, basis or assumption for this statement, and therefore, it could not be independently verified.
- b. The calculation did not include: a stress analysis for the horizontal clamp bars, swing bolts, gusset plates, tie plate, a reference to the weight of cable being restrained, the dead weight of the structure included in the analysis, or clamp bar bolt torque values. Further, there were no references as to why vertical seismic accelerations which were eliminated from the analysis do not influence loadings on the restraint.
ORGANIZATION: BABCCCK & WILC0X LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900400/86-01 RESULTS: PAGE 8 of 10 Nonconformance 86-01-04 was found in this area of the inspection.
- 4. ANO-1 Task Order 1007: Hot Leg Level Taps This task order was prepared in order to design instrument taps for a hot leg level measurement system. The B&W scope of work did not include the actual instrumentation for the system but was limited to the process point taps and two isolation / root values.
The joint configuration for the four penetration taps into each hot leg is to utilize a sleeve insert which will be rolled into each hole machined in the pipe. Prior to rolling, the sleeve would be seal welded at the pipe 10. The installation contractor is required to qualify the rolling procedure to accomplish this configuration and verifying that the sleeve could resist a 5000lb axial pull force. When asked about the basis for this requirement by the NRC inspector, with only a maximum allowed 5% wall thinning ,
of the sleeve, B&W responded that the installation contractor had also expressed some reservation concerning this requirement. The procedure and its qualification test had not yet been conducted.
Additionally, the instrument process line will be inserted into the sleeve and welded which could reduce the rolling efficiency.
A preliminary seismic analysis had been performed which showed that if root valves weighino more than 8 pounds are utilized, the natural frequency for the installation may be less than 33 hertz. The proposed valves for the taps are 1" Kerotest valves which weigh 13 pounds. B&W had not yet completed the final seismic and stress analysis required by the ccntract. Thus, these records were not available for review.
- 5. ANO-1 Task Order 460: DHR Pump Seal Cooler Flow This task order provided the analysis to justify a reduction in service water flow for the decay heat removal pump seal cooler to 3 GPM at 130 F. The B&W analysis indicated that the seals would be operating at near process steam temperature due to the reduced flow. Contacts with the pump manufacturer had indicated that this would be allowable but would significantly reduce seal life. Thus, B&W recommended that the DHR pump seals be replaced every 2 years.
ORGANIZATION: BABC0CK & WILC0X LYNCHBURG, VIRGINIA REPORT INSPECTION RESULTS: PAGE 9 of 10 N0.: 99900400/86-01
- 6. Ah0-1 Task Order 944 - Instrument Strino Calculations This task order was developed to reanalyze the instrument string error for five instruments. In the process of reviewing this task order, one instrument string for High Pressure Injection /
Low Pressure Injection initiation was evaluated for methodology and input adequacy. This calculation and its results were found to be acceptable.
The B&W reanalysis was found to utilize values for the various types of instrument errors which were different than the values provided by the manufacturer, Bailey Meter Co., on the specification sheets. B&W stated that the new values were determined in a study performed in this area and funded by the Tennessee Valley Authority (TVA). This study (EATF) had determined that, for the instruments tested, most had error tolerances lower than the manufacturer had detailed in the specification. The exceptions to this were found for the square root extractors listed on Table 7.1 of this study. Because of these -
greater inaccuracies, an instrument string utilizing one of the square root extractors was reviewed to verify that the large error values were being utilized.
In addition to the TVA report and manufacturer specification sheets, the setpoint analysis also utilized error values from a Bailey Safety Concern Report (SCR), SCR-001. This SCR dealt with inaccuracies caused by a gain potentiometer in several amplifliers.
While the data from this SCR was utilized in Task Order 944 and several other ANO-1 calculation packages that were reviewed, a
! similar process or an engineering review was not available to l determine whether or not this deficiency would affect other licensed facilities. The B&W program and procedures for review and evaluation of vender-provided safety concerns may be the object of a future inspection.
This task order was developed to change the portion of the control l
circuitry for the Emergency Feedwater and Isolation Control system.
l This change was necessitated to allow by-passing of the system during i plant cooldown af ter a steam generator tube rupture. The original circuit required both steam generators to be less than 750 psig prior to going into by-pass. This prevented system actuation on low steam generator pressure at 600 psig and results in the steam generator with the ruptured tube being isolated and preferentially fed which is l
contrary to the required actions.
I
ORGANIZATION: BABC0CK & WILC0X LYNCHBURG, VIRGINIA REPORT INSPECTION NO.: 99900400/86-01 RESULTS: PAGE 10 of 10 This design change was reviewed and found to be satisfactory. In this review, the ANO-1 Technical Specifications, FSAR, and the EFW System Design Description (SDD) were utilized. The EFW minimum flow rate of 500 GPM specified in the Technical Specifications is based on the transient analysis for a loss of main feedwater (LOMF) event.
The transient analysis assumes that the non-safety anticipatory trip is not present. In this case the pressurizer can go water solid.
In the SDD, however, the specified minimum EFW flow rate is 700 gpm per steam generator.
While this higher value is based on the necessity to keep the pres-surizer from going solid in the LOMF event, the two values could lead to confusion when compared to the FSAR. B&W stated that an effort to ensure design data is correctly stated and agreement between documents is underway. However, the SDD for the EFW system had an approval date of June 1986.
- 8. Recommendation Tracking System Report ,
During discussions with B&W cn providing service information to clients ,
a copy of a newly prepared Recommendation Tracking System Report, cated May 1986, Document I.D. 47-1163743-00, was provided for review.
The report provides each of B&W owners with recommendations from the Owner's Group along with a status of the responses from each utility.
While briefly reviewing this document, one recommendation, Number TR-027-ADM, was noted which discussed a problem with overconservative gain settings in the Power / Flow Imbalance trip circuit. The gain setting problem had resulted in several reactor trips in 1983. This recomrrendation also stated that it was an item that had resulted from" misadjustment during the alignment procedures.
! Since this item had been first recognized in 1983, the NRC inspector inquired whether the information had been provided by an alternate means to the utilities to ensure proper performarce by the maintenance personnel. No guidance was provided other than this recommendation i
and the affect the utility's trip assessment provided to the owner's l
group af ter each trip.
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Docket No. E Inspector R. G. Ambrosek Documents Examined Report No. h, Page 1 of 2 Item Type of Document -
No. Document No. Rev. Date Title / Subject !?
8 205/T4.4 Close-out notice, PSC 17-83, Overcooling Events at Low Power
. 1 Ltr PSC 17-83 3/27/86 w/ attachment " Evaluation Report PSC 17-83 Overcooling Events at Low Power ESC-171 h S
2 Ltr APL-86-158 3/12/86 Dvercooling Events at low Power (PSC 17-83)
E Personal 4/21/86 3 Files D. Mars 8/21/83 PSC 17-83 Vol. I and Vol 2 9 4 Manual UPGD-TM-45 B May 1985 Digital Power Train - Digital Computer Simulation of a B8N Nuclear Power Plant
! 5 Manual NPGD-TM-435 B May 1983 Power Train - Hybrid Computer Simulation of 88N Nuclear Power Station t 5
7 EIR 51-1152370-00 6/4/84 Resolution of PSC 41-79 Concern 4 8 TRE 8/3/83 Topical Report Evaluation on TRAP 2 9 TRE 8/31/83 Topical Report Evaluation on CADDS o
&l 10 ADM NPG-0402-01 20 9/3/85 Adm. Manual Procedures / Preparing and Processing WD Calculations R. W. Moore 11 LTR to D. Mars 10/1/85 PSC 17-83 12 File Task 286 - RCP Seal Fluid Systems Upgrade 13 File Task 500 - RCP Pump Seal Residue Analysis
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Inspector R. G. Ambrosek Docket No.
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Report No. m Page 2 of 2 Item Type of Document t No. Document No. Rev. Date Title / Subject a 0
14 File Task 885 - RCP Seal Integrity Following Loop 15 File Task 1005 - Spec. for Modified RCP Seal g M
16 Cert File DPT/3/0 0 S/20/85 DPT/3/0 - Computer Code Certification File g to 17 ADN IFG-1717-D2 3 6/3/85 Corrective Action / Reporting and Resolving Internally ~"
Identified Design Deficiencies 18 ADM NPG-0903-D3 11 3/1/85 Ngat. Info. Systems-Development of Certified Computer Program User Manuals
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