ML20205M969
| ML20205M969 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Sequoyah, Crystal River, San Onofre, Rancho Seco, 05000000 |
| Issue date: | 04/24/1986 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Heltemes C NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| References | |
| REF-GTECI-A-01, REF-GTECI-PI, RTR-NUREG-0090, RTR-NUREG-1154, RTR-NUREG-90, TASK-A-01, TASK-A-1, TASK-OR IEIN-86-019, IEIN-86-19, IEIN86-19, NUDOCS 8605010169 | |
| Download: ML20205M969 (30) | |
Text
_ _ _ _ _ _ _ _ _ _
APR 241986 MEMORANDUM FOR:
C. J. Heltemes, Jr., Director Office for Analysis and Evaluation of Operational Data FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation
SUBJECT:
ABNORMAL OCCURRENCE REPORT TO CONGRESS FOR FIRST QUARTER CY 1986 We have reviewed your memorandum of April 7,1986, containing the first draft of proposed events to be included in the First Quarter CY 1986 Abnormal Occurrence (AO) Report to Congress. We concur with the slate of events proposed as Abnormal Occurrences. We have prepared input for Appendices A, B, and C of. the report as requested. This input is provided in Enclosures 1, 2, and 3 respectively. contains the information you requested regarding license suspensions, orders, and new generic safety issues.
Please contact Gary Holahan (X24420) of the Operating Reactors Assessment Staff if you have any questions.
Origin 15Wf IT J R. Denten Harold R. Denton, Director Office of Nuclear Reactor Regulation i
Enclosures:
As Stated i
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i INPUT TO ABNORMAL OCCURRENCE WRITE-UPS LOSS OF POWER AND WATERHAPHER EVENT (SAN ONOFRE UNIT 1, NOVEMBER 21, 1986) 1.
NRR Follow-Up Actions NRR has the lead responsibility for six of the fifteen major activities in the NRC Action Plan.
They include the following plant-specific and generic items:
1.
Assessment of need to re-evaluate USI A-1, (i.e., waterhammer) to address condensation-induced waterhammers in feedwater piping (Generic) 2.
Assessment of adequacy of certain San Onofre-1 design features other than check valves (Plant Specific) 3.
Assessment of the integrity of the repaired feedwater line (Plant Specific) 4.
Adequacy of abnormal conditions and post-trip reviews (Generic and PlantSpecific) 5.
Evaluation of licensee effort to assure that low levels in steam generators during the event did not damage or leave damaging chemical'
+
materials in the steam generators (Plant Specific).
6.
Determine if a backlog of license amendments delayed approval of Inservice Testing (IST) Program (Plant Specific)
Reviews in each of these areas are currently ongoing. The most recent 1,
licensee submittal containing responses to Staff requests for information was received en April 8, 1986. The Commission has completed its review of Item 6 above and has determined that a backlog of license amendments did not exist and thus did not cause a delay in approval of the licensee's IST program.
Lead responsibility for resolving generic-issues related to check valve j
failures has been assumed by industry per a meeting with the NRC. Executive Director for Operations (ED0) on April 7, 1986. NRR and other offices 1
will monitor and review industry actions. At the meeting the industry group agreed to provide the Staff with an initial program plan for resolution of the generic issues by May 7, 1986.
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LOSS OF INTEGRATED CONTROL SYSTEM POWER AND OVERC00 LING TRANSIENT (RANCHO SECO, DECEMBER 26,1985) 1.
NRR Follow-Up Actions In response to the ED0 memorandum of March 13, 1986, the NRR staff has organized their follow-up activity to address seven issues which are primarily generic in nature. They include:
(1) Adequacy of the auxiliary feedwater system; (2) Completeness of previous Staff and licensee actions associated with control systems; (3) Adequacy of the design of the integrated control system (ICS);
- 4) Adequacy of maintenance programs for m6nual isolation valves;
- 5) Adequacy of procedures and training; i
- 6) Adequacy of the FSAR accident analyses; and (7) Adequacy of required staffing.
Generic and plant-specific actions to resolve these issues have been identified.
Generic actions will be completed within the scope of existing. generic pro-grams or the design reassessment of Babcock & Wilcox plants described below.
Plant-specific actions identified in the March 13, 1986, memorandum will be completed in the course of the Rancho Seco re-start review.
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2.
Summary and Status of B&W Design Reassessment Following the TMI accident there has been a growing realization among the NRC staff of the sensitivity of Babcock and Wilcox (B&W) plants to operational transients. A number of recent events at B&W-designed reactors have reinforced their concerns regarding these designs and lead them
+.0 conclude that there is a need to re-examine the basic design requirements for B&W reactors. While they believe that this reassessment is needed, they also believe that B&W reactors can safely continue to operate in the interim. The B&W Owners Group (B&WOG) has committed to taking a leadership role in the reassessnent.
In a January 24, 1986, letter (Reference 1) to the B&WOG, the NRC Executive Director for Operations (ED0) communicated plans for the design reassess-ment and outlined the scope envisioned for the study. A program plan for the reassessment was subsequently developed by the NRR staff and transmitted to the B&WOG in a March 13, 1986, letter to the B&WOG Chairman (Reference 2).
The plan calls for studies in the area of operating experience, transient analysis, and probabilistic risk assessment. The plan also identified those areas that the Staff expects the B&WOG to take the lead or play a major role in completing.
As outlined in a March 21, 1986, memorandum from the EDO to the Commissioners (Reference 3), the B&WOG will assume a strong leadership role in accomplishing key aspects of the overall effort, where such involvement is appropriate.
In a meeting with the Staff on April 8,1986, they presented their program-plan for reducing the reactor trip frequency and improving the transient response of B&W-designed plants. The NRR staff is currently reviewing this plan.
References 1.
Letter from Victor Stello, Jr., Acting NRC Executive Director for Operations, to Hal Tucker, Chairman, B&W Owners Group; January 24, 1986.
2.
Letter from Dennis M. Crutchfield, Assistant Director for Technical Support, NRC Division of PWR Licensing - 8, to Hal Tucker, Chairman, B&W Cwners Group; March 13, 1986.
3.
Memorandum from Victor Stello, Jr., Acting Executive Director. for Operations, to the Commissioners entitled, "B&W Design Reassessment";
March 21, 1986.
4 e
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UPDATING MATERIAL (APPENDIX B 0F A0 REPORT) 85-7 Loss of Main and Auxiliary Feedwater Systems This abnormal occurrence, which occurred at Davis-Besse on June 9, 1985, was originally reported in NUREG-0090, Vol. 8, No. 2, " Report to Congress on Abnornal Occurrences: April-June 1985", and updated in NUREG-0090, Vol. 8, No. 3 and Vol. 8, No. 4.
It is further updated as follows.
As mentioned in the previous update, based upon the findings of the NRC Incident Investigation Team reported in NUREG-1154 (Reference 1),
the NRC identified the concerns Toledo Edison Company (the licensee) should address for NRC review before' resumption of operation of the plant can be approved. These concerns were identified to the licensee in a letter dated August 14, 1985 (Reference 2).
The licensee has responded in a document submitted to the NRC on September 10, 1985, entitled, " Davis-Besse Course of Action" (Reference 3). The NRC staff has essentially completed its review of this document and is currently preparing a Safety Evaluation Report to address plant re-start. Several open items still exist for which the Staff has requested additional information from Toledo Edison. The current schedule calls for completion of the Safety Evaluation Report by May 31, 1986, and briefing of the ACRS in early July.
References 1.
U.S. Nuclear Regulatory Commission, " Loss of Main and Auxiliary Feedwater
[
Event at the Davis-Besse Plant on June 9, 1985", USNRC Report NUREG-1154, published July 1985.
y 2.
10 CFR f 50.54(f) letter from Harold R. Denton, Director, NRC Office of Nuclear Reactor Regulation, to Joe Williams, Jr., Senior Vice President-Nuclear, Toledo Edison Company, Docket No. 50-346, August 14, 1985.
3.
Letter from John P. Williamson, Chairman and Chief Executive Officer, Toledo I
Edison Company, to Harold R. Denton, Director, NRC Office of Nuclear Reactor Regulation, Docket No. 50-346, September 10, 1985.
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i 85-14 Management Deficiencies at Tennessee Valley Authority This abnormal occurrence was originally reported in NUREG-0090, Vol. 8, No. 3, " Report to Congress on Abnormal Occurrences: July-September 1985".
It is updated as follows.
The NRC staff, led by a senior management team, has identified a number of major TVA issues requiring resolution prior to the restart of any of the TVA reactors. The Staff briefed the Commission on these issues on January 7,1986, and stated that Sequoyah Unit 2 was expected to be the first reactor ready to resume operation. TVA briefed the Commission on a number of TVA activities on January 9,1986, and noted the appointment of Steven White as the new Manager of _
l Nuclear Power. The Staff briefed the Commission on TVA status on February 7,1986. TVA briefed the Commission again on March 11, 1986.
By letter dated March 10, 1986, TVA submitted their revised response to the September 17, 1985, 10 CFR 50.54(f) letter regarding corporate concerns. This submittal is under review by the Staff.
There are five areas where there has been considerable TVA activity-and which have also received a significant level of staff attention.
These areas are equipment qualification, employee concerns, welding, electrical design calculations, and simulator evaluations of Sequoyah J
licensed personnel.
The following paragraphs summarize activity during First Quarter 1986 in these areas, focusing primarily on the Sequoyah facility.
Equipment Qualification d
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.As TVA has completed portions of the equipment qualification (EQ) effort the Staff conducted on-site reviews, including an audit in November 1985, a two-week inspection in January 1986, and an 0
inspection during the week ending February 14, 1986. Additional on-site reviews will be conducted.
Discrepancies identified by both the Staff and TVA remain to be corrected; however the TVA EQ Program appears to be sound. The staff currently estimates that all EQ discrepancies vill be resolved by June 1986 and TVA will be able to certify that Sequoyah is in compliance with the EQ rule.
TVA and the Staff will face additional effort to assure EQ compliance at Browns Ferry and Watts Bar.
Employee Concerns Nearly 5,000 TVA employee concerns have been raised; some of these involve safety-related and intimidation, harassment, or wrongdoing issues. About 400 of these concerns apply to the Sequoyah facility. TVA has established a program for evaluating employee concerns and is making the transition from a program administered by
Quality Technology Company (QTC) to a TVA-administered program. The Staff has conducted several inspections of the QTC administered employee concerns program. Resolution of employee concerns may be a pacing item leading to Sequoyah re-start or Watts Barr licensing.
On February 11, 1986, TVA submitted a summary of the TVA-Administered Employee Concerns Program and the methodology TVA will use for resolution of concerns generated by the Watts Bar Program. The Staff responded to TVA regarding this submittal on February 28, 1986, and requested a TVA staff briefing and response to a number of staff Concerns.
TVA and QTC terminated their contract in April 1986. The Staff has obtained copies of~all the QTC-generated employee concerns records and is preparing a program to review and evaluate many of these issues, as well as provide technical issues to TVA for resolution while maintaining employee confidentiality. The magnitude of this effort is unknown at this time.
The NRC staff is currently interviewing individuals and following up on allegations brought directly to the NRC by concerned individuals.
In addition, the Staff is initiating an effort to review, on an expedited basis, all intimidation and harassment issues and request identification of TVA actions taken on those items determined to be
' safety significant.
f Welding Activities completed by the Staff and TVA in the First Quarter CY 1986 are as follows:
q The staff reviewed the TVA Weld Reinspection Plan for Sequoyah and found it acceptable with comments which were provided to TVA.
TVA completed reinspection of 800 welds at Sequoyah during the week ending February 14, 1986.
s Members of the staff were at the Sequoyah site February 24-28, 1986, conducting independent inspection of about 300 welds using the NRC's mobile nondestructive examination van.
Electrical Design Calculations Activities completed in the First Quarter CY 1986 in this area are as follows:
The Staff conducted an on-site inspection of the Sequoyah facility in mid-January to evaluate the TVA program and on-site progress.
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I il TVA has completed the Sequoyah electrical design review.
TVA is examining the entire design and configuration control program for Sequoyah, including areas outside of the electrical area. This may result in _ considerable TVA and Staff effort-in the next few months, and could become a pacing item for Sequoyah re-start.
Sequoyah Simulator Evaluations In response to high failure rates for operator. requalification exams at Browns Ferry, the Staff has conducted simulator evaluations of Sequoyah operating personnel. These evaluations were completed during the week ending February 28, 1986, and.the results were that 21 out of 24 personnel performed adequately. The one crew (3 persons) that was weak is receiving additional training and will be re-evaluated by TVA. NRC will audit the licensee's retaining and re-evaluation.
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APPENDIX C ITEM (OTHER EVENTS OF INTEREST)
DEGRADED REACTOR COOLANT PUMP (RCP) SHAFTS On January 1,1986, Crystal River 3 was shutdown because of a problem with reactor coolant pump (RCP) "A".
Examination showed that the RCP "A" shaft had failed completely within the hydrostatic bearing due to fatigue propagation of small cracks. The cause is still uncertain but may involve material properties, design inadequacies, manufacturing cracks, residual stresses, bending moments, or thermal cycling.
Florida Power Corporation is planning to meet with the NRC on April 24, 1986, to provide information as to the root cause of the RCP shaft failure.
The failure occurred rapidly with essentially no warning to operators.
It has been determined that the reactor coolant flow rate decreased from the four pump value to the three pump value within three seconds following failure of the shaft. The post-trip review indicates that the reactor tripped on a power / flow mismatch signal approximately five seconds after the failure.
The RCP seals were damaged but did not leak at the time of shaft failure.
RCP "B" shaft has been removed and is extensively cracked at a slightly different location.
RCP "C" shaft ultrasonic testing (UT) indicated a crack less severe than "B" pump. RCP "D" shaft UT indicated a crack of the same magnitude as "B" pump. All eight cap bolts securing the impeller to the shaft on "A" and "B" pumps were cracked in multiple places (some broken).
Five of eight pins which take the torque between the impeller and shaft on "A" and "B" pumps were also cracked. Their appearance is similar to the defective cap bolts. The reactor coolant pumps at Crystal River-3 were manufactured by Byron-Jackson (BJ).
Toledo Edison, which utilizes RCPs similar to those at Crystal River, took advantage of being shutdown to examine the RCP shafts at Davis-Besse. They used the same team from Babcock & Wilcox (B&W) that examined the Crystal River RCPs shafts. UT examination showed cracks in all four RCP shafts with one crack being a circumferential crack about 1" deep. These RCPs were also manu-factured by BJ, and are of the same horsepower and speed (RPM) as the Crystal River-3 RCPs. Additionally, the pump shafts are made of the same material (ASTM A461 GR 660) and have seal injection cooling.
The Office of Inspection and Enforcement (IE) has transmitted an Infonnation Notice 86-19 on the Crystal River-3 and Davis-Besse RCP shaft failures to all nuclear power reactor facilities holding an operating license (0L) or a con-struction permit (CP).
The Office of Nuclear Reactor Regulation is continuing to investigate the generic implications of multiple RCP shaft cracking in similarly designed pumps. This is because a number of nuclear power plants utilize BJ RCPs, and the failure of an RCP shaft can be similar to a seized shaft accident in which specified acceptable fuel design limits (SAFDL) may be exceeded.
I
ADDITIONAL INFORMATION REQUESTED FROM NRR REGARDING LICENSE SUSPENSIONS, ORDERS, AND NEW SAFETY ISSUES 1.
License Suspensions:
There have been no license suspensions for commercial nuclear power reactors during the First Quarter CY 1986.
2.
Orders Covering License Modifications for Safety Reasons:
There have been no orders covering license modifications for safety reasons issued during the First Quarter CY 1986.
3.
Identification of Those Generic Safety Concerns Approved by the Director of NRR in the First Quarter CY 1986:
a.
Generic Issues 122.1, (A) Common Mode Failure of Isolation Valves in the Closed Position (B) Recovery of Auxiliary Feedwater (C) Interruption of Auxiliary Feedwater Flow b.
Generic Issue 122.2, Initiating Feed and Bleed c.
Generic Issue 124, AFW System Reliability.
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APR 24 EEE MEMORANDUM FOR:
C. J. Heltemes, Jr., Director Office for Analysis and Evaluation of Operational Data FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation
SUBJECT:
ABNORMAL OCCURRENCE REPORT TO CONGRESS FOR FIRST QUARTER CY 1986 We have reviewed your memorandum of April 7,1986, containing the first draft of proposed events to be included in the First Quarter CY 1986 Abnornal Occurrence (A0) Report to Congress. We concur with the slate of events proposed as Abnormal Occurrences. We have prepared input for Appendices A, B, and C of the report as requested. This input is provided in Enclosures 1, 2, and 3, respectively. contains the information you requested regarding license suspensions, orders, and new generic safety issues.
Please contact Gary Holahan (X24420) of the Operating Reactors Assessnent Staff if you have any questions.
i Original UWM E S R. Gtattn Harold R.' Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
As Stated g/
cc:
P. Bobe Distribution Central File w/Incoping NRC PDR w/ Incoming /
NSIC w/ Incoming ORAS Rdg H. Denton D. Eisenhut PPAS NRR #86-200/D. Mossburg G. Holahan R. Wessman
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INPUT TO ABNORMAL OCCURRENCE WRITE-UPS LOSS OF POWER AND WATERHAMMER EVENT (SAN ONOFRE UNIT 1, NOVEMBER 21,1985) 1.
NRR Follow-Up Actions NRR has the lead responsibility for six of the fifteen major activities in the NRC Action Plan. They include the following plant-specific and generic items:
1.
Assessment of need to re-evaluate USI A-1, (i.e., waterhammer) to address condensation-induced waterhammers in feedwater piping (Generic) 2.
Assessment of adequacy of certain San Onofre-1 design features other than check valves (Plant Specific) 3.
Assessment of the integrity of the repaired feedwater line (PlantSpecific) 4.
Adequacy of abnormal conditions and post-trip reviews (Generic and PlantSpecific) 5.
Evaluation of lict e effort to assure that low levels in steam generators during the event did not damage or leave damaging chemical materials in the steam generators (Plant Specific).
6.
Determine if a backlog of license amendments delayed approval of Inservice Testing (IST) Program (Plant Specific)
Reviews in each of these areas are currently ongoing. The most recent
- [
licensee submittal containing responses to Staff requests for information was received on April 8, 1986. The Commission has completed its review of Item 6 above and has determined that a backlog of license amendments did not exist and thus did not cause a delay in approval of the licensee's IST program.
Lead responsibility for resolving generic issues related to check valve failures has been assumed by industry per a meeting with the NRC Executive Director for Operations (ED0) on April 7,1986..NRR and other offices will monitor and review industry actions. At the meeting the industry group agreed to provide the Staff with an initial program plan for resolution of the generic issues by May 7,1986.
9
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LOSS OF INTEGRATED CONTROL SYSTEM POWER AND OVERC00 LING TRANSIENT (RANCHO SECO, DECEMBER 26,1985) 1.
NRR Follow-Up Actions In response-to the EDO memorandum of March 13, 1986, the NRR staff has organized their follow-up activity to address seven issues which are primarily generic in nature. They include:
(1) Adequacy of the auxiliary feedwater system;.
(2) Completeness of previous Staff and licensee actions associated with-control systems; (3) Adequacy of the design of the integrated control system'(ICS);
- 4) Adequacy of maintenance programs for manual isolation valves;
~
- 5) Adequacy of procedures and training;
- 6) Adequacy of the FSAR accident analyses; and (7) Adequacy of required staffing.
Generic and plant-specific actions to resolve these issues have been identified.'
Generic actions will be completed within the scope of existing generic pro-grams or the design reassessment of Babcock & Wilcox plants described below.
Plant-specific actions identified in the March 13, 1986, memorandum will be completed in ti. course of the Rancho Seco re-start review.
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2.
Summary and Status of B&W Design Reassessment Following the TMI accident there has been a growing realization among the NRC staff of the sensitivity of Babcock and Wilcox (B&W) plants to operational transients. A number of recent events at B&W-designed reactors have reinforced their concerns regarding these designs and lead them to conclude that there is a need to re-examine the basic design requirements for B&W reactors. While they believe that this reassessment is needed, they also believe that B&W reactors can safely continue to operate in the interim. The B&W Owners Group (B&WOG) has committed to taking a leadership role in the reassessment.
In a January 24, 1986, letter (Reference 1) to the B&WOG, the NRC Executive Director for Operations (ED0) communicated plans for the design reassess-ment and outlined the scope envisioned for the study. A program plan for the reassessment was subsequently developed by the NRR staff and transmitted to the B&WOG in a March 13, 1986, letter to the B&WOG Chairman (Reference 2).
The plan calls for studies in the area of operating experience, transient analysis, and probabilistic risk assessment. The plan also identified those areas that the Staff expects the B&WOG to take the lead or play a major role in completing.
As outlined in a March 21, 1986, memorandum from the ED0 to the Commissioners (Reference 3), the B&WOG will assume a strong leadership role in accomplishing key aspects of the overall effort, where such involvement is appropriate.
In a meeting with the Staff on April 8, 1986, they presented their program plan for reducing the reactor trip frequency and improving the transient response of B&W-designed plants. The NRR staff is currently reviewing this plan.
References j
1.
Letter from Victor Stello, Jr., Acting NRC Executive Director for Operations, to Hal Tucker, Chairman, B&W Owners Group; January 24,-1986.
2.
Letter from Dennis M. Crutchfield, Assistant Director for Technical Support, NRC Division of PWR Licensing - 8, to Hal Tucker, Chairman, B&W Owners Group; March 13, 1986.
3.
Memorandum from Victor Stello, Jr., Acting Executive Director for Operations, to the Commissioners entitled, "B&W Design Reassessment";
March 21, 1986.
- j i
~,
UPDATING MATERIAL (APPENDIX B 0F A0 REPORT) 85-7 Loss of Main and Auxiliary Feedwater Systems This abnormal occurrence, which occurred at Davis-Besse on June 9, 1985, was originally reported in NUREG-0090, Vol. 8, No. 2, " Report to Congress on Abnormal Occurrences: April-June 1985", and updated in NUREG-0090, Vol. 8, No. 3 and Vol. 8, No. 4.
It is further updated as follows.
As mentioned in the previous update, based upon the findings of the NRC Incident Investigation Team reported in NUREG-1154 (Reference 1),
the NRC identified the concerns Toledo Edison Company (the licensee) should address for NRC review before resumption of operation of the plant can be approved. These concerns were identified to the licensee in a letter dated August 14, 1985 (Reference 2).
The licensee has responded in a document submitted to the NRC on September 10, 1985, entitled, " Davis-Besse Course of Action" (Reference 3). The NRC staff has essentially completed its review of this document and is currently preparing a Safety Evaluation Report to address plant re-start.
Several open items still exist for which the Staff has requested additional information from Toledo Edison.
The current schedule calls for. completion of the Safety Evaluation Report by May 31, 1986, and briefing of the ACRS in early July.
1 References 1.
U.S. Nuclear Regulatory Commission, " Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9,1985", USNRC Report NUREG-1154, published July 1985.
2.
10 CFR 6 50.54(f) letter from Harold R. Denton, Director, NRC Office of Nuclear Reactor Regulation, to Joe Williams, Jr., Senior Vice President-Nuclear.
Toledo Edison Company, Docket No. 50-346, August 14, 1985.
3.
Letter from John P. Williamson, Chairman and Chief Executive Officer, Toledo Edison Company, to Harold R. Denton, Director, NRC Office of Nuclear Reactor Regulation, Docket No. 50-346, September 10, 1985.
19 l
4
85-14 Management Deficiencies at Tennessee Valley Authority This abnormal occurrence was originally reported in NUREG-0090, Vol. 8, No. 3, " Report to Congress on Abnormal Occurrences: July-September 1985";
It is updated as follows.
The NRC staff, led by a senior management team, has identified a number of major TVA issues requiring resolution prior to the restart of any of the TVA reactors. The Staff briefed the Commission on these issues on January 7,1986, and stated that Sequoyah Unit 2 was expected to be the first reactor ready to resume operation.
TVA briefed the Commission on a number of TVA activities on January 9,1986, and noted the appointment of Steven White as the new Manager of Nuclear Power. The Staff briefed the Commission on TVA status on February 7,1986. TVA briefed the Commission again on March 11, 1986.
By letter dated March 10, 1986, TVA submitted their revised response to the September 17, 1985, 10 CFR 50.54(f) letter regarding corporate concerns. This submittal is under review by the Staff.
l There are five areas where there has been considerable TVA activity and which have also received a significant level of staff attention.
These areas are equipment qualification, employee concerns, welding, electrical design calculations, and simulator evaluations of Sequoyah licensed personnel. The following paragraphs summarize activity during First Quarter 1986 in these areas, focusing primarily on the Sequoyah facility.
Equipment Qualification As TVA has completed portions of the equipment qualification (EQ) effort the Staff conducted on-site reviews, including an audit in November 1985, a two-week inspection in January 1986, and an inspection during the week ending February 14, 1986. Additional on-site reviews will be conducted.
Discrepancies identified by both the Staff and TVA remain to be corrected; however the TVA EQ Program appears to be sound. The staff currently estimates that all EQ discrepancies will be resolved by June 1986 and TVA will be able to certify that Sequoyah is in compliance with the EQ rule.
t TVA and the Staff will face additional effort to assure EQ compliance at Browns Ferry and Watts Bar.
j Employee Concerns Nearly 5,000 TVA employee concerns have been raised; some of these
}
involve safety-related and intimidation, harassment, or wrongdoing i
issues. About 400 of these concerns apply to the Sequoyah facility. IVA has established a program for evaluating employee concerns and is making the transition from a program administered by
-u!
Quality Technology Company (QTC) to a TVA-administered program.
The Staff has conducted several inspections of the QTC-administered employee concerns program. Resolution of employee concerns may be a pacing item leading to Sequoyah re-start or Watts Barr licensing.
On February 11, 1986, TVA submitted a suninary of the TVA-Administered Employee Concerns Program and the methodology TVA will use for resolution of concerns generated by the Watts Bar Program. The Staff responded to TVA regarding this submittal on February 28, 1986, and requested a TVA staff briefing and response to a number of staff Concerns.
TVA and QTC terminated their contract in April 1986. The Staff has -
obtained copies of all the QTC-generated employee concerns records and is preparing a program to review and evaluate many of these issues, as well as provide technical issues to TVA for resolution while maintaining employee confidentiality. The magnitude of this effort is unknown at this time.
The NRC staff is currently interviewing individuals and following up on allegations brought directly to the NRC by concerned individuals.
In addition, the Staff is initiating an effort to review, on an expedited basis, all intimidation and harassment issues and request-identification of TVA actions taken on those items determined to be safety significant.
Welding Activities completed by the Staff and TVA in the First Quarter CY 1986 are as follows:
The staff reviewed the TVA Weld Reinspection Plan for Sequoyah and found it acceptable with. comments which were provided to TVA.
TVA completed reinspection of 800 welds at Sequoyah during the week ending February 14, 1986.
Members of the staff were at the Sequoyah site February 24-28, 1986, conducting independent inspection of about 300 welds using the NRC's mobile nondestructive examination van.
Electrical Design Calculations Activities completed in the First Quarter CY 1986 in this area are as follows:
The Staff conducted an on-site inspection of the Sequoyah facility in mid-January to evaluate the TVA program and on-site progress.
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TVA has completed the Sequoyah electrical design review.
TVA is examining the entire design and configuration control program for Sequoyah, including areas outside of the electrical area. This may result in considerable TVA and Staff effort in the next few months, and could become a pacing item for Sequoyah re-start.
Sequoyah Simulator Evaluations In response to high failure rates for operator requalification exams-at Browns Ferry, the Staff has conducted simulator evaluations of Sequoyah operating personnel. These evaluations were completed during the week ending February 28, 1986, and the results were that 21 out of 24 personnel performed adequately. The one crew (3_ persons) that was weak is receiving additional training and will be re-evaluated by TVA. NRC will audit the licensee's retaining and re-evaluation.
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APPENDIX C ITEM (OTHER EVENTS OF INTEREST)
DEGRADED REACTOR COOLANT PUMP (RCP) SHAFTS On January 1,1986, Crystal River 3 was shutdown because of a problem with reactor coolant pump (RCP) "A".
Examination showed that the RCP "A" shaft had failed completely within the hydrostatic bearing due to fatigue propagation of small cracks. The cause is still uncertain but may involve material properties, design inadequacies, manufacturing cracks, residual stresses, bending moments, or thermal cycling. Florida Power Corporation is planning to meet with the NRC on April 24, 1986, to provide information as to the root cause of the RCP shaft failure.
The failure occurred rapidly with essentially no warning to operators.
It has been determined that the reactor coolant flow rate decreased from the four pump value to the three pump value within three seconds following failure of the shaft. The post-trip review indicates that the reactor tripped on a power / flow mismatch signal approximately five seconds after the failure.
The RCP seals were damaged but did not leak at the time of shaft failure.
RCP "B" shaft has been removed and is extensively cracked at a slightly different location. RCP "C" shaft ultrasonic testing (UT) indicated a crack less severe than "B" pump.
RCP "D" shaft UT indicated a crack of the same magnitude as "B" pump. All eight cap bolts securing the impeller to the shaft on "A" and "B" pumps were cracked in multiple places (some broken).
Five of eight pins which take the torque between the impeller and shaft on "A" and "B" pumps were also cracked. Their appearance is similar to the defective cap bol ts. The reactor coolant pumps at Crystal River-3 were manufactured by Byron-Jackson (BJ).
Toledo Edison, which utilizes RCPs similar to those at Crystal River, took advantage of being shutdown to examine the RCP shafts at Davis-Besse. They used the same team from Babcock & Wilcox (B&W) that examined the. Crystal River RCPs shafts. UT examination showed cracks in all four RCP shafts with one crack being a circumferential crack about 1" deep. These RCPs were also manu-factured by BJ, and are of the same horsepower and speed '. RPM) as the Crystal River-3 RCPs. Additionally, the pump shafts are made of the same material (ASTM A461 GR 660) and have seal injection cooling.
The Office of Inspection and Enforcement (IE) has transmitted an Information-Notice 86-19 on the Crystal River-3 and Davis-Besse RCP shaft failures to all nuclear power reactor facilities holding an operating license (OL) or a con-struction permit (CP).
The Office of Nuclear Reactor Regulation is continuing to investigate the generic implications of multiple RCP shaft cracking in similarly designed pumps. This is because a number of nuclear power plants utilize BJ RCPs, and the failure of an RCP shaft can be similar to a seized shaft accident in which
~
specified acceptable fuel design limits (SAFDL) may be exceeded.
r r
f ADDITIONAL INFORMATION REQUESTED FROM NRR REGARDING ETTENSE SUSPENSIONS, ORDERS, AND NEW SAFETY ISSUES 1.
License Suspensions:
There have been no license suspensions for commercial nuclear power rea'ctors during the First Quarter CY 1986.
2.
Orders Covering License Modifications for Safety Reasons:
There have been no orders covering license modifications for safety reasons issued during the First Quarter CY 1986.
3.
Identificattor, of Those Generic Safety Concerns Approved by the Director of NRR in the First Quarter CY 1986:
a.
Generic Issues 122.1, (A) Common Mode Failure of Isolation Valves in the Closed Position (B) Recovery of Auxiliary Feedwater (C) Interruption of Auxiliary Feedwater Flow b.
Generic Issue 122.2, Initiating Feed and Bleed c.
Generic Issue 124, AFW System Reliability.
8
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APR 241986 MEMORANDUM FOR:
C. J. Heltemes, Jr., Director Office for Analysis and Evaluation of Operational Data FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation
SUBJECT:
ABNORMAL OCCURRENCE REPORT TO CONGRESS FOR FIRST QUARTER CY 1986 We have reviewed your memorandum of April 7,1986, containing the first draft of proposed events to be included in the First Quarter CY 1986 Abnormal Occurrence (A0) Report to Congress. We concur with the slate of events proposed as Abnormal Occurrences. We have prepared input for Appendices A, B, and C of the report as requested. This input is provided in Enclosures 1, 2, and 3, respectively. contains the information you requested regarding license suspensions, orders, and new generic safety issues.
Please contact Gary Holahan (X24420) of the Operating Reactors Assessment Staff if you have any questions.
Oriscal 6! F g E.Gtattn Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
As Stated cc:
P. Bobe Distribution Central File w/ Incoming NRC PDR w/Incomiryg NSIC w/ Incoming /
ORAS Rdg H. Denton D. Eisenhut PPAS NRR #86-200/D. Mossburg G. Holahan Q)g R. Wessman T 4 M. Carusp J U E*MR B,.01(k"S 'Q q
00:1 0:
ORAS MCARUS0: dim RWESSMAN GbOLAHAN HUT 1R TON y/D/86 4/p /86 ef/g/86
/86
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i INPUT TO ABNORMAL OCCURRENCE WRITE-UPS LOSS OF POWER AND WATERHAMMER EVENT (SAN ONOFRE UNIT 1, NOVEMBER 21, 1986) 1.
NRR Follow-Up Actions NRR has the lead responsibility for six of the fifteen major activities in the NRC Action Plan. They include the following plant-specific and generic items:
1.
Assessment of need to re-evaluate USI A-1, (i.e., waterhammer) to address condensation-induced waterhammers in feedwater piping (Generic) 2.
Assessment of adequacy of certain San Onofre-1 design features other than check valves (Plant Specific) 3.
Assessment of the integrity of the repaired feedwater line (Plant Specific) 4.
Adequacy of abnormal conditions and post-trip reviews (Generic and PlantSpecific) 5.
Evaluation of licensee effort to assure that low levels in steam generators during the event did : sot damage or leave damaging chemical 4
materials in the steam generators (Plant Specific).
6.
Determine if a backlog of license amendments delayed approval.of Inservice Testing (IST) Program (Plant Specific)
Reviews in each of these areas are currently ongoing. The most recent
[
licensee submittal containing responses to Staff requests for information was received on April 8, 1986. The Commission has completed its revie, of Item 6 above and has determined that a backlog of license amendments did not exist and thus did not cause a delay in approval of the licensee's IST program.
Lead responsibility for resolving generic issues related.to check valve failures has been assumed by industry per a meeting with the NRC Executive Director for Operations (ED0) on April 7, 1986. NRR and other offices will monitor and review industry actions. At the meeting the industry group agreed to provide the Staff with an initial program plan for resolution-of the generic issues by May 7, 1986.
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e LOSS OF INTEGRATED CONTROL SYSTEM POWER AND OVERC00 LING TRANSIENT (RANCHO SECO, DECEMBER 26,1985) 1.
NRR Follow-Up Actions In response to the ED0 memorandum of March 13, 1986, the NRR staff has organized their follow-up activity to address seven issues which are primarily generic in nature. They include:
(1) Adequacy of the auxiliary feedwater system; (2) Completeness of previous Staff and licensee actions associated with control systems; (3) Adequacy of the design of the integrated control system (ICS);
(4) Adequacy of maintenance programs for manual isolation valves;.
(5) Adequacy of procedures and training; (6) Adequacy of the FSAR accident analyses; and (7) Adequacy of required staffing.
Generic and plant-specific actions to resolve these issues have been identified.
Generic actions will be completed within the scope of existing generic pro-grams or the design reassessment of Babcock & Wilcox plants described below.
Plant-specific actions identified in the March 13, 1986, memorandum will be completed in the course of the Rancho Seco re-start review, n
3 Y
a ea
n 2.
Summary and Status of B&W Design Reassessment Following the TMI accident there has been a growing realization among the NRC staff of the sensitivity of Babcoci: and Wilcox (B&W) plants to operational transients. A number of recent events at B&W-designed reactors have reinforced their concerns regarding these designs and lead them to conclude that there is a need to re-examine the basic design requirements for B&W reactors. While they believe that this reassessment is needed, they also believe that B&W reactors can safely continue to.
operate in the interim. The B&W Owners Group (B&WOG) has committed to taking a leadership role in the reassessment.
In a January 24, 1986, letter (Reference 1) to the B&WOG, the NRC Executive Director for Operations (ED0) communicated plans for the design reassess-ment and outlined the scope envisioned for the study. A program plan for the reassessment was subsequently developed by the NRR staff and transmitted to the B&WOG in a March 13, 1986, letter to the B&WOG Chairman (Reference 2).
The plan calls for studies in the area of operating experience, transient analysis, and probabilistic risk assessment. The plan also identified those areas that the Staff expects the B&WOG to take the lead or play a major role in completing.
As outlined in a March 21, 1986, memorandum from the EDO to the Commissioners (Reference 3), the B&W0G will assume a strong leadership role in accomplishing key aspects of the overall effort, where such involvement is appropriate.
In a meeting with the Staff on April 8,1986, they presented their program plan for reducing the reactor trip frequency and improving the transient response of B&W-designed plants.
The NRR staff is currently reviewing this plan.
7 References 1.
Letter from Victor Stello, Jr., Acting NRC Executive Director for Operations, to Hal Tucker, Chairman, B&W Owners Group; January 24, 1986.
2.
Letter from Dennis M. Crutchfield, Assistant Director for Technical
?
Support, NRC Division of PWR Licensing - 8, to Hal Tucker, Chairman, B&W Owners Group; March 13,1986.
3.
Memorandum from Victor Stello, Jr., Acting Executive Director for Operations, to the Commissioners entitled, "B&W Design Reassessment";
March 21, 1986.
1 e
UPDATING MATERIAL (APPENDIX B 0F A0 REPORT) 85-7 Loss of Main and Auxiliary Feedwater Systems This abnormal occurrence, which occurred at Davis-Besse on June 9,1985, was originally reported in NUREG-0090, Vol. 8, No. 2, " Report to Congress
~
4 on Abnormal Occurrences: April-June 1985", and updated in NUREG-0090, Vol. 8, No. 3 and Vol. 8, No. 4.
It is further updated as follows.
As mentioned in the previous update, based upon the findings of the NRC Incident Investigation Team reported in NUREG-1154 (Reference 1),
the NRC identified the concerns Toledo Edison Company (the licensee) should address for NRC review before resumption of operation of the plant can be approved. These concerns were identified to the licensee in a letter dated August 14, 1985 (Reference 2).
3 The licensee has responded in a document submitted to the NP,C on September 10, 1985, entitled, " Davis-Besse Course of Action" (Reference 3). The NRC staff has essentially completed its review of this document and is currently preparing a Safety Evaluation Report to address plant re-start. Several open items-still exist for which the Staff has requested additional information from Toledo Edison. The current schedule calls for completion of the Safety Evaluation Report by May-31, 1986, and briefing of the ACRS in early July.
References 1.
U.S. Nuclear Regulatory Commission, " Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9, 1985", USNRC Report NUREG-1154, published July 1985.
2.
10 CFR 6 50.54(f) letter from Harold R. Denton, Director, NRC Office of Nuclear Reactor Regulation, to Joe Williams, Jr., Senior Vice President-Nuclear, Toledo Edison Company, Docket No. 50-346, August 14, 1985.
3.
Letter from John P. Williamson, Chairman and Chief Executive Officer, Toledo Edison Company, to Harold R. Denton, Director, NRC Office of Nuclear Reactor Regulation, Docket No. 50-346, September 10, 1985.
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85-14 Management Deficiencies at Tennessee Valley Authority This abnormal occurrence was originally repo. ted in NUREG-0090, Vol. 8, No. 3, " Report to Congress on Abnormal Occurrences: July-September 1985".
It is updated as follows.
The NRC staff, led by a senior management team, has identified a number of major TVA issues requiring resolution prior to the restart of any of the TVA reactors. The Staff briefed the Commission on these issues on January 7,1986, and stated that Sequoyah Unit 2 was expected to be the first reactor ready to resume operation. TVA briefed the Coninission on a number of TVA activities on January 9,1986, and noted the appointment of Steven White as the new Manager of Nuclear Power. The Staff briefed the Commission on TVA status on February 7,1986. TVA briefed the Commission again on March 11, 1986.
By letter dated March 10, 1986, TVA submitted their revised response to the September 17, 1985, 10 CFR 50.54(f) letter regarding corporate concerns. This submittal is under review by the Staff.
There are five areas where there has been considerable TVA activity and which have also received a significant level of staff attention.
These areas are equipment qualification, employee concerns, welding, electrical design calculations, and simulator evaluations of Sequoyah licensed personnel. The following paragraphs summarize activity during First Quarter 1986 in these areas, focusing primarily on the Sequoyah facility.
Equipment Qualification As TVA has completed portions of the equipment qualification (EQ) effort the Staff conducted on-site reviews, including an audit in November 1985, a two-week inspection in January 1986, and an inspection during the week ending February 14, 1986. Additional
'i on-site reviews will be conducted.
Discrepancies identified by both the Staff and TVA remain to be 1
corrected; however the TVA EQ Program appears to be sound. The staff currently estimates that all EQ discrepancies will be resolved by June 1986 and TVA will be able to certify that Sequoyah is in compliance with the EQ rule.
TVA and the Staff will face additional effort to assure EQ compliance at Browns Ferry and Watts Bar.
Employee Concerns Nearly 5,000 TVA employee concerns have been raised; some of these involve safety-related and intimidation, harassment, or wrongdoing issues. About 400 of these concerns apply to the Sequoyah facility. TVA has established a program for evaluating employee concerns and is making the transition from a program administered by
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D Quality Technology Company (QTC) to a TVA-administered program.
The Staff has conducted several inspections of the QTC-administered employee concerns program. Resolution of employee concerns may be a pacing item leading to Sequoyah re-start or Watts Barr licensing.
On February 11, 1986, TVA submitted a summary of the TVA-Administerea Employee Cr7cerns Program and the methodology TVA will use for resolution of concerns generated by.the Watts Bar Program. The Staff responded to TVA regarding this submittal on February 28, 1986, and requested a TVA staff briefing and. response to a number of staff Concerns.
TVAandQTCterminatedtheircontract_inApri]1986. The Staff has obtained copies of all the QTC-generated employee concerns records and is preparing a program to review and evaluate many of these issues, as well as provide technical issues to TVA for resolution while maintaining employee confidentiality. The magnitude of this effort is unknown at this time.
The NRC staff is currently interviewing individuals and following up on allegations brought directly to the NRC by concerned individuals.
In addition, the Staff is initiating an effort to review, on an i
expedited basis, all intimidation and harassment issues and request identification of TVA actions taken on those items determined to be safety significant.
Welding Activities completed by the Staff and TVA in the First Quarter CY 1986 are as follows:
The staff reviewed the TVA Weld Reinspection Plan for Sequoyah
-l and found it acceptable with comments which were provided to i
TVA.
TVA completed reinspection of 800 welds at Sequoyah during the week ending February 14, 1986.
Members of.the staff were at the Sequoyah site February 24-28, 1986, conducting independent inspection of about 300 welds using the j
NRC's mobile nondestructive examination van..
Electrical Design Calculations Activities completed in the First Quarter CY 1986 in this area are
- i as follows:
4 The Staff conducted an on-site inspection of the Sequoyah facility in mid-January to evaluate the TVA program and on-site progress.
.A l
i-i TVA has completed the Sequoyah electrical design review.
TVA is examining the entire design and configuration control program for Sequoyah, includir,g areas outside of the electrical area. This may. result in considerable TVA ard Staff effort in the next few months, and could become a pacing item for Sequoyah re-start.
Sequoyah Simulator Evaluations In response to high failure rates for operator requalification exams at Browns Ferry, the Staff has conducted simulator evaluations of Sequoyah operating personnel. These evaluations were completed during the week ending February 28, 1986, and the results were that i
21 out of 24 personnel performed adequately. The one crew (3 persons) that was weak is receiving additional training and will be re-evaluated by TVA. NRC will audit the licensee's retaining and re-evaluation.
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s e
k APPENDIX C ITEM (OTHER EVENTS-0F INTEREST)
DEGRADED REACTOR COOLANT PUMP (RCP) SHAFTS On January 1,-1986, Crystal River 3 was shutdown beelise of a problem with reactor coolant pump (RCP) "A".
Examination snoweo : tat the RCP "A" shaft had failed completely within the hydrostatic bea' ring due to fatigue propagation of small cracks.
The cause is still uncertain but may involve material properties, design inadequacies, manufacturing cracks, residual stresses, bending moments, or thermal cycling. Florida Power Corporation is planning to meet with the NRC on April 24, 1986, to provide information as to the root cause of the RCP shaft failure.
The failure occurred rapidly with essentially no warning to operators.
It has been determined that the reactor coolant flow rate decreased from the four pump value to the three pump value within three seconds following failure of the shaft. The post-trip review indicates that the reactor tripped on a power / flow mismatch signal approximately five seconds after the failure.
The RCP seals were damaged but did not leak at the time of shaft failure.
RCP "B" shaft has been removed and is extensively cracked at a slightly different location. RCP "C" shaft ultrasonic testing (UT) indicated a crack less severe than "B" pump. RCP "D" shaft UT indicated a crack of the same magnitude as "B" pump. All eight cap bolts securing the impeller to the shaft-on "A" and "B" pumps were cracked in multiple' places (some broken).
Five of-eight pins which take the torque between.the impeller and shaft on "A" and "B" pumps were also cracked. Their appearance is similar to the defective cap bolts. The reactor coolant pumps at Crystal River-3 were manufactured by fc Byron-Jackson (BJ).
3 Toledo Edison, which utilizes RCPs similar to those at Crystal River, took advantage of being shutdown 'to examine the RCP. shafts at Davis-Besse. They used the same team from Babcock & Wilcox (B&W) that examined the Crystal River RCPs shafts. UT examination showed cracks in all four RCP' shafts with one crack being a circumferential crack about 1"-deep. These RCPs were also manu-factured by BJ, and are of the same horsepower and speed ~(RPM) as the Crystal River-3 RCPs. Additionally, the pump shafts are made of the same material (ASTM A461 GR 660) and have seal injection cooling.
The Office of Inspection and Enforcement (IE);has transmitted an Information Notice 86-19 on the Crystal River-3 and Davis-Besse RCP shaft failures to all nuclear power reactor facilities holding an operating: license (OL) or a con-struction permit (CP).
The Office of Nuclear Reactor Regulatibn is continuing to investigate the generic. implications of multiple RCP. shaft cracking in similarlyfdesigned pumps. This is because a number of nuclear. power plants utilize BJ RCPs, and the failure of an RCP. shaft can be similar-to a seized shaft accident in which specified acceptable fuel design limits (SAFDL) may.be exceeded.
-d %'$
c ADDITIONAL INFORMATION REQUESTED FROM NRR REGARDING LICENSE SUSPENSIONS, ORDERS, AND NEW SAFETY ISSUES 1.
License Suspensions:
There have been no license suspensions for commercial nuclear power reactors during the First Quarter CY 1986.
2.
Orders Covering License Modifications for Safety Reasons:
There have been no orders covering license modifications for safety reasons issued dur-ing the First Quarter CY 1986.
3.
Identification of Those Generic Safety Concerns Approved by the Director of NRR in the First Quarter CY 1986:
a.
Generic Issues 122.1, (A) Common Mode Failure of Isolation Valves-in the Closed Position (B) Recovery of Auxiliary Feedwater (C) Interruption of Auxiliary Feedwater Flow b.
Generic Issue 122.2, Initiating Feed and Bleed c.
Generic Issue 124, AFW System Reliability.
4
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