ML20055B279
| ML20055B279 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 07/31/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0519, NUREG-0519-S04, NUREG-519, NUREG-519-S4, NUDOCS 8207210136 | |
| Download: ML20055B279 (36) | |
Text
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NUREG-0519 Supplement No. 4 Safety Evaluation Report related to the operation of LaSalle County Station, Units 1 and 2 Docket Nos. 50-373 and 50-374 Commonwealth Edison Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1982
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NOTICE Availability of Reference Materials Cited in NRC Publications Co-scuments cited in NRC publications will be available (rom one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
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Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.
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NUREG-0519 Supplement No. 4 Safety Evaluation Report related to the operation of LaSalle County Station, I
Units 1 and 2 Docket Nos. 50-373 and 50-374 Commonwealth Edison Company l
U.S. Nuclear Regulatory Commission l
Office of Nuclear Reactor Regulation l
July 1982 p
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TABLE OF CONTENTS Pajte 1.
INTRODUCTION AND GENERAL DISCUSSION............................
1-1 1.1 Introduction..............................................
1-1 5.
REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS...................
5-1 5.2 Integrity of Coolant Pressure Boundary....................
5-1 5.2.4 leactor Coolant Pressure Boundary Preservice and Inservice Inspection and Testing...................
5-1 6.
ENGINEERED SAFETY SYSTEMS......................................
6-1
- 6. 2 Containment Systems.......................................
6-1 6.2.3 Containment Isolation System.......................
6-2 6.2.3.6 Control Rod Drive and Withdrawl Lines.....
6-2 6.3 Emergency Core Cooling....................................
6-2 6.3.2 Evaluation.........................................
6-2 6.3.2.3 Functional Design.........................
6-2 6.6 Inservice Inspection of Class 2 and 3 Components..........
6-6 6.6.2 Evaluation of Compliance for Unit No. 2 to 10 CFR 50.55a(g)...................................
6-6
- 17. QUALITY ASSURANCE..............................................
17-1 17.4 Assurance of Proper Design and Construction...............
17-1 17.4.1 Background......................................
17-1 17.4.2 Verification Program............................
17-1 17.4.3 Assessment by Teledyne..........................
17-2 17.4.4 Assessment by NRC Staff.........................
17-2 17.4.5 Conclusions.....................................
17-3 22.
TMI-2 Requirements.............................................
22-1 22.2 TMI Action Plan Requirements for Applicants for Operating Licenses.................................................
22-1 La Salle SSER 4 i
TABLE OF CONTENTS Pane III Emergency Preparations and Radiation Protection......
22-1 III.A.2 Improving Licensee Emergency Preparedness Long-Term...................................
22-1 APPENDICES A.
CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW FOR THE LA SALLE COUNTY STATI0N...............................
A-1 8.
PRESERVICE INSPECTION PROGRAM FOR UNIT NO.
2..............
B-1 l
C.
NRC STAFF CONTRIBUTORS AND CONSULTANT.....................
C-1 j
l 1
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La Salle SSER 4 ii
l 5
1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction On March 5, 1981, the Nuclear Regulatory Commission (NRC) staff issued its Safety Evaluation Report (NUREG-0519) regarding the application by Commonwealth Edison Company (hereinafter referred to as the licensee) for licenses to i
operate the La Salle County Station, Unit Nos. 1 and 2 (hereinafter referred to as La Salle or facility), Docket Nos. 50-373 and 50-374.
On June 12, 1981, the Safety Evaluation Report was supplemented by Supplement No. 1 which documented the resolution of several outstanding issues in futher support of the licensing activities.
On February 12, 1982, the NRC staff issued Supplement No. 2 to the Safety Evaluation Report in which we addressed the open items identified in the Safety Evaluation Report and Supplement No. 1.
On April 17, 1982, we issued Supplement No. 3 to the Safety Evaluation Report in which we addressed those items required to be addressed for low power license.
In addition or, April 17, 1982, a license, NPF-11, was issued to allow Unit 1 operation at power' levels not to exceed 5 percent of rated power.
This report is Supplement No. 4 to our Safety Evaluation Report.
Since issuance of License No. NPF-11, several license conditions have been met.
License Condition 2.C.(5)(a) was satisfied by revising the Technical Specifica-tions to include a revised list of safety-related snubbers.
This was accom-plished by License Amendment No. 1 dated June 18, 1982.
The preoperational tests, startup tests, and other items identified in Attach-ment 1 to Licensee No. NPF-11, and referenced in License Condition 2.C.(1) have all been confirmed as complete.
The licensee has been notified of this by a June 19, 1982 letter from the Administrator of NRC Region III.
The NRC staff has also confirmed that the following installations, tests, and other items have been completed as required prior to exceeding 5 percent power:
(1) Additional fire protection in the diesel generator corridor has been added as required by License Condition 2.C.(25)(b).
(2) General Electric has reviewed the power ascension test procedures and the General Electric recommendations have been incorporated by the licensee as reqaired by License Condition 2.C.(30)(b).
(3) Post accident sampling capability as required by License Condition 2.C.-
(30)(f) has been installed and tested.
(4) Additional accident monitoring instrumentation as required by License Condition 2.C.(30)(h) has been installed and associated procedures have been approved by the NRC staff (see letter dated June 19, 1982 from the Administrator of NRC Region III).
La Salle SSER 4 1-1
This report addresses the resolution of safety-related issues, in addition to those identified above, that required further evaluation prior to authorizing operation above 5 percent power.
These issues are:
(1) Verfication review of proper design and construction This is License Condition 2.C.(29).
The staff's overview of this veri-fication review is provided in Section 17.4 of this report.
Based on this overview, the staff concludes that License Condition 2.C.(29) has been met.
(2) Improving the licensee's emergency preparedness This is License Condition 2.C.(3)(p), and is addressed in Section 22.2.-
III.A.2 of this report.
Finally, this report also addresses resolution of additional La Salle safety-related issues which were not required to be resolved prior to authorizing operation of Unit 1 above 5 percent power:
(1) Preservice inspection relief from code requirements for certain Unit 2 components This is discussed in Sections 5.2.4 and 6.6.2 and Appendix B of this report.
(2) Scram discharge volume piping failure Section 6.3.2.3 of this report evaluates the licensee's response to NUREG-0803 which is covered by License Codition 2.C.(14)(b).
With respect to License Condition 2.C.(4) regarding rebar damage and adequacy of the off gas building roof and restricting the license to zero power until NRC approval, a discussion of the NRC staff investigation on these matters will I
be provided in a separate report.
l In summary, based on the safety evaluations completed to date and on the
{
completion of installation, tests and procedures required to be completed prior j
to exceeding 5 percent power, the NRC staff concludes that subject to continued
)
compliance with all of the conditions and requirements of Operating License No. NPF-11, as amended, the licensee can safely operate La Salle County Station, Unit No. 1 up to full rated power.
The items addressed in this report are covered in sections having the dame number and title as the section of the Safety Evaluation Report and its supple-ments in which they were previously discussed.
Appendix A of this report is a continuation of the chronology of the radiological review of La Salle.
Appen-dix B is the preservice inspection program for Unit No. 2.
Appendix C is a list of the principal NRC staff and consultant reviewers who contributed to this supplement.
The NRC project manager for La Salle is Dr. Anthony Bournia.
Dr. Bournia may be contacted by writing to the Division of Licensing, U.S.
Nuclear Regulatory Commission, Washington, D.C.
20555.
La Salle SSER 4 1-2
5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.4 Reactor Coolant Pressure Boundary Preservice and Inservice Inspection and Testing 5.2.4.2 Evaluation of Compliance for Unit No. 2 to 10 CFR 50.55(g)
In our Safety Evaluation Report, we specified that a preservice inspection pro-gram for Unit No. 2 was being performed by the licensee based on the 1974 Edi-tion through the Summer 1975 Addenda of Section XI of the American Society of Mechanical Engineers Code.
This preservice inspection program was completed and was submitted by the licensee in a letter dated September 1, 1981.
Speci-fic written relief from Code requirements was requested and supported by infor-mation in a letter dated April 5, 1982. We determined that certain American Society of Mechanical Engineers Code Section XI examination requirements defined in 10 CFR Part 50, Section 50.55a(g)(2) are impractical.
We have evaluated the American Society of Mechanical Engineers Code required examinations that have been determined to be impractical and, pursuant to 10 CFR Part 50, Section 50.55a(a)(2), have allowed deviations from the require-ments that have been determined to be impractical and that if implemented would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.
Based on the granting of relief from these preservice examination requirements, we conclude tht the preservice inspection program for Unit No. 2 is in compliance with 10 CFR Part 50, Section 50.55a(g)
(2).
Our detailed evaluation supporing this conclusion is provided in Appen-dix B to this report.
The initial inservice inspection program for Unit No. 2 will be evaluated after the applicable American Society of Mechanical Engineers Code Edition and Addenda can be determined based on Section 50.55a(b) of 10 CFR Part 50 and before the first refueling outage when inservice inspections will be performed.
As a part of our review, we will evaluate any relief requests as allowed by paragraph IWB-1220 of the American Society of Mechanical Engineers Code.
A supporting technical justification will be presented in a supplement to this report.
5.2.4.3 Conclusions The conduct of periodic inspections and hydrostatic testing of pressure retain-ing components of the reactor coolant pressure bundary, in accordance with the requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and 10 CFR Part 50, will provide reasonable assurance that evidence of structural degradation or loss of leaktight integ-rity occuring during service will be detected in time to permit corrective action before the safety functions of a component are compromised.
Compliance with the preservice inspections required by this Code and 10 CFR Part 50 LaSalle SSER 4 5-1
constitutes an acceptable basis for satisfying the inspection requirements of Criteria 32 of the General Design Criteria.
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1 6 ENGINEERED SAFETY FEATURES 6.2 Containment Systems Concerns have been raised by Mr. John Humphrey involving the Mark III contain-ment design.
These issues were identified during the time period when Mr. Humphrey was a General Electric employee involved in the detailed design of the standard Mark III containment design known as the STRIDE package.
Mr. Humphrey has since resigned from General Electric and transmitted his concerns to the owner of the lead Mark III plant, Grand Gulf Nuclear Station, Units 1 and 2 (Docket Nos. 50-416 and 50-417).
On May 27, 1982, Mississippi Power and Light Company, the licensee of Grand Gulf, and Mr. Humphrey met with the staff to discuss these issues as they relate to Grand Gulf.
Based on the results of this meeting, we have made a preliminary finding that no significant design deficiencies associated with these concerns have.been uncovered.
Therefore, the issuance of a low power license for Grand Gulf was not withheld.
The concerns, however, have raised certain questions which must be addressed prior to the issuance of a full power license for Grand Gulf, Unit 1 and prior to final resolution of all the issues. The staff's review is con-tinuing in this regard.
Similar reviews will also be conducted on all the other Mark III plants.
Although the concerns raised by Mr. Humphrey were specifically directed to the Mark III STRIDE design, we have evaluated the applicability of these concerns to the Mark I and Mark II containments.
Our preliminary review indicates that several concerns could be applicable to all boiling water reactor pressure suppression containments.
Therefore, the staff will request all boiling water reactor Mark I and Mark II owners to address the issues that are applicable to their designs.
For the interim, however, it is our judgment that these concerns need not delay the full power licensing schedule for la Salle, Unit 1.
The basis for our judg-ment is provided below.
(1) Based on our review of the issues and the Mississippi Power & Light response, we have concluded that the technical issues identified were for the most part considered in the design of the Grand Gulf containment; we have not to date uncovered any deficiency in the containment design; (2) Design differences between Mark II and Mark III containment make many of the issues not pertinent to Mark II containments; and (3) The staff has been informed by the La Salle licensee, during a telecon on June 18, 1982, that they have completed a preliminary evaluation of the concerns.
Based on their initial assessment of Humphrey's concerns, they have not ideatified any design deficiency.
The La Salle licensee also stated that the numerous conservatisms employed in the design of the La Salle containments have not been eroded.
La Salle SSER 4 6-1
l l
l In addition, the applicant has committed to submit their plans for final reso-lution of Mr. Humphrey's concerns to demonstrate the adequacy of the La Salle containment design.
Based on the above, we conclude that the licensee's approach to resolving Mr. Humphrey's concerns is acceptable and that full power licensing of La Salle, Unit 1 may proceed as scheduled.
6.2.3 Containment Isolation Systems 6.2.3.6 Control Rod Drive Insert and Withdrawal Lines In addition to the lines identified in Sections 6.2.3.1 through 6.2.3.4 in our Safety Evaluation Report, the design of the control rod drive insert and withdrawal lines represents a departure from the explicit requirements of the General Design Criteria.
Both the control rod drive insert and withdrawal lines are provided with normally closed, fail-closed, solenoid-operated direc-tional control valves, which open only during routine movement of their as-sociated control rod.
The normally closed, fail-open air-operated scram inlet and exhaust valves open only when required to effect a rapid reactor shutdown i
(scram).
In addition, manual shutoff valves are provided for positive isola-I tion in the unlikely event of a pipe break within a hydraulic control unit.
(These units and the valves described above are located outside containment to satisfy testing, inspection, and maintenance requirements).
In addition, each control rod drive insert line is provided with an automatically actuated ball check valve inside containment. We find that the system design represents a departure from the explicit requirements of General Design Criteria.
- However, in accordance with the provisions of Appendix A to 10 CFR Part 50 and Cri-terion 55 of the General Design Criteria which permits departure from its explicit requirement, we find that the control rod drive containment isolation provision stated above is acceptable on the basis stated in NUREG-0803, " Safety Evaluation Report Regarding Integrity of BWR Scram Systems," dated August 1981.
6.3 Emergency Core Cooling System 6.3.2 Evaluation 6.3.2.3 Functional Design Scram Discharge System Pipe Break In Supplement No. 2 to our Safety Evaluation Report, we indicated that, based on the conclusions of NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping," dated August 1981, the scram discharge piping system would be acceptable, provided the plant-specific recommendations of the NUREG are satisfied on an individual plant basis.
The safety concerns associated with a postulated failure of the system discharge piping do not represent a significant contribution to the risk of core melt, provided the assumptions used in the NUREG risk assessment are validated on an individual plant basis.
We indicated that we would review the licensee's response, "LaSalle Evaluation Report Regarding Integrity of Scram System Piping," dated January 21, 1982, to N'JREG-0803.
La Salle SSER 4 6-2
We have now completed the review of the licensee's submittal.
Our review was primarily concerned with the operability of systems required to mitigate the effects of the postulated event.
Chapter 5 of NUREG-0803 provides the summary of the staff's guidance for an acceptable resolution of the major concerns resulting from the boiling water reactor scram system piping evaluation.
The respective concerns and the associated evaluation for La Salle follows.
Verification of Equipment Designed for Water Impingement During the review of the Final Safety Analysis Report, it was ascertained that the plant design would meet the single failure criterion with regard to flood-ing of safety-related equipment, and that the capability f.,e safe shutdown would be maintained.
Our evaluation is presented in Sections 3.4.1 and 9.3.2 of the Safety Evaluation Report.
In the licensee's January 21, 1982 submittal, infor-mation was provided identifying the equipment required for safe shutdown follow-ing the postulated event, and further that the required equipment would not be disabled due to flooding and other environmental effects caused by the scram discharge piping failure.
The reactor building and equipment and floor drain-age design is such as to make wetdown of emergency core cooling system pumps from the postulated event highly unlikely.
The emergency core cooling system equipment is located 87 feet below and four floors down from the hydraulic con-trol units and scram discharge volume piping.
The corner rooms containing the emergency core cooling system pumps are isolated from the basement inner annu-lus by normally closed water-tight doors with curbs approximately one foot high.
Opening of these doors is annunciated in the control room.
Floor drains on the hydraulic control unit level drain into a sump in the basement inner annulus and not into the emergency core cooling system pump rooms.
Further, the emer-gency core cooling system pump motors are located approximately 9 feet above the floor.
This height precludes any credible flood level resulting from the postulated event from impinging upon the emergency core cooling system pump motors. We conclude that adequate flood and water impingement protection is provided.
Verification of Essential Components Qualified for Service at 212 degrees Fahrenheit and 100 percent Humidity and Verification of Equipment Qualified for Wetdown by 212 degrees Fahrenheit Water The licensee states that the emergency core cooling system pump motors are qualified for wetdown by 212 degrees Fahrenheit water from the top.
The emer-gency core cooling system pump motors and electrical systems are qualified for service with environmental conditions in excess of 212 degrees Fahrenehit and 100 percent humidity.
The licensee reviewed the instruments situated within local racks in the environmental zone which was determined to suffer the most damaging effects from the scram discharge volume rupture.
An adequate number of qualified instruments exist to provide sufficient information to the opera-tor so that safe shutdown and long-term cooling can be achieved to mitigate the accident effects, and thus this recommendation is satisfied.
Evaluation of Availability of High Pressure Core Injection-Low Pressure Core Injection Turbines due to High Ambient Temperature Trips The high pressure core spray at La Salle performs the equivalent function of the high pressure core injection system at other plants.
The high pressure La Salle SSER 4 6-3
core spray is powered by an electric rnotor.
The high pressure core spray pump motor is qualified for the postulated scram discharge volume rupture environ-ment, and thus this recommendation is satisfied.
Verification of Feedwater and Condensate System Operation Independent of the Reactor Building Environment The condensate and feedwater systems would be available except in the event of a loss of offsite power.
Operability of the condensate and feedwater pumps would not be affected by the postulated event due to their remote location.
La Salle has a motor-driven feedwater pump that is independent of the main steam system.
Hence, it would be available following a closure of the main steam isolation valves.
We conclude that the La Salle systems have the capa-bility to provide sufficient reactor makeup to more than offset the nominal leakage rate of 550 gallons per minute assumed in NUREG-0803 for the postulated pipe crack.
Scram Discharge Volume Pipe Break-Detection Signals Control room alarms that would alert the operators regarding occurrence of the scram discharge volume pipe failure include reactor building radiation area alarm, reactor building floor drain sump alarm, control rod drive high tempera-ture alarm, reactor building ventilation high radiation alarm, and reactor build-ing differential pressure alarm.
The control rod drive high temperature alarm is powered from a standby bus and scans samples at a rate of 5 seconds per point.
Thus, a scan of ali the control rod drives is completed every 16 minutes.
The licensee estimates that nearly all of the control rod drives would exceed the high temperature setpoint instead of the 5 to 15 percent that are usually seen after a scram. While the radiation monitors in the hydraulic control units-scram discharge volume area are not connected to an emergency power source, radiation monitors on other floors are connected to emergency power, as are the reactor building ventilation high radiation alarms.
The sump high-high level is also not dependent on offsite power.
We conclude that La Salle has adequate instrumentation to detect the postulated event in a timely manner.
Periodic Inservice Inspection and Surveillance for the Scram Discharge Volume System A preservice inspection was made of the control rod drive system in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI (1975 Addenda).
This inspection included nondestructive test-ing of several welds in the scram discharge headers.
The results of the pre-service inspection are described within the Unit 1 and Unit 2 preservice inspec-tion reports.
Inservice inspections are planned for each unit as described within the respective La Salle inservice inspectiun program.
Thus, this recom-mendation is satisfied.
Threaded Joint Integrity There are r.o threaded joints utilized in the La Salle scram discharge volume piping to be assessed for integrity loss due to various forms of potentially damaging stimuli, and thus this recommendation does not apply to La Salle.
La Salle SSER 4 6-4
Seismic Design Verification The scram discharge volume piping, including the control rod drive insert and withdraw lines; the scram discharge volume; the scram discharge instrument volume; and the vent and drain lines hate been stress analyzed with seism-ic loadings.
The drain line stress level meets the requirements of American Society Mechanical Engineers Code Section III, Class B (1974).
The remaining lines were designed to satisfy the requirements of American Society Mechanical EngineersSection III, Class 2 piping.
All of the essential scram discharge volume active components have been seismically qualified.
Thus, we conclude that this recommendation is satisfied.
l Hydraulic Control Unit and Scram Discharge Volume Equipment Procedures Review The scram discharge volume surveillance, maintenance, inspection or modifica-tion actions is strictly controlled by Commonwealth Edison quality assurance procedures and La Salle administrative procedures.
Any work performed on the control rod drives, hyraulic control units, or scram discharge volume is classified as safety-related or American Society of Mechanical Engineers code related.
It is, therefore, required to be subject to multiple levels of review, with final approval from the shift supervisor, prior to work commence-ment.
Thus, we conclude that this recommendation is satisfied.
Environmental Qualification of Prompt Depressurization Function The prompt depressurization function is qualified to meet the postulated hydraulic control unit break environmental conditions.
The automatic depres-surization system is located within the primary containment and consists of safety-related equipment.
Thus, we conclude that the automatic depressuriza-tion system is appropriately qualified and would not be exposed to a degrading environment in the event of a scram discharge volume rupture or leak in the reactor building.
As-Built Inspection of Scram Discharge Volume Piping and Supports The as-built inspection of the Unit 1 scram discharge volume piping was com-pleted in March 1982.
The Unit 2 piping will be similarly inspected prior to the Unit 2 fuel load.
The Unit 2 piping arrangement is equivalent to that to be installed in Unit 1 during the first refueling outage.' The piping arrange-ment for both units will have dual vent valves and dual drain valves along with with redundant level measuring transducers for the instrument volume.
Thus, we conclude that this recommendation is satisfied.
Improvement of Procedures La Salle has developed operating annunciator procedures which incorporate specific direct'ons for the reactor operator to initiate prompt depressuriza-tion upon detection of a significant non-isolable leak outside containment.
The reactor operators will be trained to react to pipe break symptoms outside primary containment; the training will specifically address the scram discharge volume rupture event.
Thus, we conclude that this recommendation is satisfied.
La Salle SSER 4 6-5
Limitation of Coolant Iodine Concentration to Standard Technical Specification Values The La Salle Technical Specification for coolant activity limits the level to less than that allowed by the Standard Technical Specifications.
Thus, we conclude that this recommendation is satisfied.
Conclusion We conclude that the La Salle design meets the requirements of Criterion 4 of the General Design Criteria with regard to the capability of structures, sys-tems and components important to safety to accommodate the effects of flooding and other environmental effects caused by the postulated pipe crack, and that we have reasonable assurance that safe shutdown can still be accomplished.
As concluded in NUREG-0803, occurrence of the postulated event before corrective action is taken has a low probability. We conclude that the La Salle design is acceptable.
6.6 Inservice Inspection of Class 2 and 3 Components 6.6.2 Evaluation of Compliance for Unit No. 2 to 10 CFR 50.55a(g)
In our Safety Evaluation Report, we specified that a preservice inspection pro-gram for Unit No. 2 was being performed by the licensee based on the 1974 Edi-tion through Summer 1975 Addenda of Section XI of the American Society of l
Mechanical Engineeers Code.
This preservice inspection program was completed 1
l and was submitted by the licensee in a letter dated September 1, 1981.
Speci-fic written relief from Code requirements was requested and supported by infor-mation in a letter dated April 5,1982. We determined that certain American Society of Mechanical Engineers Code Section XI examination requirements I
defined in 10 CFR 50.55a(g)(2) are impractical.
We have evaluated the American Society of Mechanical Engineers Code required l
examinations that have been determined to be impractical and, pursuant to l
10 CFR Part 50.55a(a)(2), have allowed deviations from the requirements that have been determined to be impractical and that if implemented would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.
Based on the granting of relief from these preservice examination requirements, we conclude that the preservice inspection program for Unit No. 2 is in compliance with 10 CFR 50.55a(g)(2).
Our detailed evalua-tion supporting this conclusion is provided in Appendix B to this report.
l The initial inservice inspection program for Unit No. 2 will be evaluated after the applicable American Society of Mechanical Engineers Code Edition and Addenda can be determined based on 10 CFR 50.55a(b) and before the first refueling outage when inservice inspections will be performed.
A supporting technical justification will be presented in a supplement to this report.
La Salle SSER 4 6-6
1 1
17 QUALITY ASSURANCE 17.4 Assurance of Proper Design and Construction 17.4.1 Background In Supplement No. 2 of our Safety Evaluation Report, we indicated that in light of the problems with quality assurance for design that have recently been noted at other facilities, we believe that further assurance, based on a more detailed examination of portions of the design process is appropriate.
Accordingly, we requested the licensee to hire an independent contractor, approved by us, to perform a review of the mechanical and structural design of loop C residual heat removal system excluding all branch piping less than 3 inches, in the functioning mode of the low pressure injection system using loads resulting from the actuation of the automatic depressurization system in conjunction with the operating basis earthquake to verify that this system has been designed and constructed in accordance with the application and that the NRC requirements have been satisfied.
Commonwealth Edison contracted the Teledyne Engineering Services (Teledyne) to perform this review with our approval.
Teledyne completed its contracted program.
As part of this program, Teledyne reviewed the quality assurance programmatic control and their~ implementation in the design area, the licensee's audit findings related to the design activities of Sargent & Lundy (architect-engineer), General Electric (vendor), and Quadrex (subcontractor), the "as-built" documentation to actual plant configuration, the design process, and the consistency of the design documents with the Final Safety Analysis Report commitments.
17.4.2 Verification Program The verification program conducted by Teledyne was performed under the fol-lowing tasks:
Task 1 - Review of Design Process and Control, Task 2 - Review of Design Procedures, Task 3 - Review of Interface Procedures, Task 4 - Review of Implementation of Design and Interface Procedures, Task 5 - Determination of "As Built" Configuration of Unit 1 Residual Heat Removal Loop C, Task 6 - Comparison of Unit 1 "As-Built" Documentation to the Residual Heat Removal Loop C Configuration, Task 7 - Review of Design Documents versus Final Safety Analysis Report Commitments, La Salle SSER 4 17-1
Task 8 - Review of Commonwealth Edison's Audit Findings, Task 9 - Project Review Internal Committee, and Task 10 - Reporting.
17.4.3 Assessment By Teledyne The preliminary findings by Teledyne resulted in 21 Error / Deviation and 31 open-items reports which were transmitted to the licensee and the NRC staff.
Upon submittal of all Teledyne's preliminary findings, the licensee transmitted its responses to Teledyne and the NRC staff and, in addition, the licensee recevied permission from the NRC staff to establish a dialogue between Teledyne and its Architect-Engineer (Sargent & Lundy) to discuss the potential errors found in the Teledyne review.
Of these 52 reports which involved various problems in the design area and none in quality assurance, 39 were closed by Teledyne based on the acquisition of additional information and/or clarifica-tion of existing information.
The 13 remaining reports were reviewed by Teledyne's Project Review Internal Committee.
This committee, composed of three senior level Teledyne engineers who together had the expertise to resolve the technical issues, and the Teledyne Project Manager concurred that none of these reports have the potential for significant safety impact.
17.4.4 Assessment by NRC Staff The NRC staff reviewed the Teledyne evaluation reported in Technical Report No. 5539-2, dated May 28, 1982. We reviewed Teledyne's evaluation on the quality assurance and implementation, the adequacy of the design process, design procedures, interface procedures, and the consistency of design docu-ments with the Final Safety Analysis Report.
The major portion of Teledyne's report appears to focus on the evaluation of the analytical and design pro-cedures pertaining to the residual heat removal piping system (loop C), equip-ment, and component supports.
The final report provides a list of all open items, errors, deviations, closed items, observations, and findings that developed through the Teledyne review process and this list involved only in the design area and none in quality assurance.
As indicated above, 52 reports were generated and 39 were closed as a result of additional clarifications and supplemental responses.
The 13 remaining reports were determined not to have the potential for significant safety impact.
The NRC staff reviewed those open-items and error-deviations reports submitted to the Project Review Com-mittee and concluded that these reports can be categorized as not having a significant safety impact on La Salle.
In addition, the NRC staff feels that Teledyne has performed an in-depth review of the analytical procedures and design calculations used in the piping, equipment, and component support design to assure the adequacy of the design bases, the adequacy of the design imple-mentation, and the consistency between the design documents and the Final Safety Analysis Report commitments.
The NRC staff reviewed Teledyne's evaluation of the comparison of "as-built" documentation to actual plant configuration as provided in Sections 5.5 and 5.6 and Appendices H and I of Technical Report No. 5539-2.
The report provides a detailed description of the initial walkdown of the selected piping system and the findings of the field survey, and the subsequent followup field survey with La Salle SSER 4 17-2
1 the resolutions of the concerns noted in the first survey.
The field survey of the "as-built" piping system was general in nature and did not include every component and aimension partly due to the inaccessibility of some of the areas where the piping was routed.
However, the NRC staff feels that a sufficient portion of the major piping system and the amount of detail of those parameters that were checked provide adequate assurance that the "as-built" piping con-figuration has been satisfactorily evaluated.
I In Section 2.1 of the report, Teledyne stated that the development of the response spectra would be excluded from review since it was not part of the program. We reviewed the list of response spectra and the piping stress analyses referenced in the report and have several concerns which were not included in the scope of Teledyne's review regarding the validity of the hydrodynamic response spectra used in the final "as-built" piping analyses.
The NRC staff met with the licensee on June 15, 1982 to discuss this item.
As a result of our discussion with the licensee, we found that the documenta-tion of the final hydrodynamic loads used to generate the containment building response spectra that were used as input for the "as-built" piping analysis was provided in an internal report entitled, "SRV/LOCA Hydrodynamic Loads Revised Design Basis Summary Report - La Salle County Station, Units 1 and 2,"
SL-3876, dated October 1, 1981.
The primary purpose of this report was to document which revisions of the hydrodynamic loads were used in the final reanalysis and redesign of La Salle and to document the load definitions used in the genera-tion of the containment building response spectra.
Based on our review of this internal report (SL-3876), we believe that the hydrodynamic containment building response spectra used in the final piping system analyses have been adequately documented.
Thus, providing further assurance of the adequacy of the inter-disciplinary review and interface control, including the internal and external transmittal distribution, and use of design data among and within all inter-facing design organizations.
17.4.5 Conclusions The independent design verification program conducted by Teledyne on the loop C residual heat removal system indicated that the quality assurance control and implementation, design process, procedures and Final Safety Analysis Report commitments are acceptable except in the area of response spectra.
The results of the limited review provide increased assurance that the quality assurance program established and implemented by the licensee and its principal contractors did effectively control the overall program and construction activities for the La Salle County Station. While several design deficiencies were identified, the overall design and construction activities were adequately performed so that no adverse impact on safety was found.
Therefore, with respect to as-surance of proper design and construction, the NRC staff concludes that there is an acceptable basis for granting authority to operate the facility at power levels up to and including full power.
La Salle SSER 4 17-3
1 22 TMI-2 Requirements 22.2 TMI Action Plan Requirements for Licensees for Operating License III Emergency Preparations and Radiation Protection III.A NRC and Licensee Preparedness III.A.2 Improving Licensee Emergency Preparedness - Long Term Discussion and Conclusion Since the issuance of Supplement No. 2 to our Safety Evaluation Report, the Commonwealth Edison Company conducted a full-scale offsite emergency prepared-ness exercise involving the State of Illinois and local counties within the 10 mile Emergency Planning Zone for La Salle.
The exercise was conducted over a 2-day period on April 14-15, 1982.
The licensee participated on the first day only.
For the second day of the exercise, the licensee provided controlled messages to ensure offsite actions would be implemented. We found relevant to that exercise, that only minor areas for improvement were identified.
This exercise, plus the full-scale exercise (licensee, State and local counties) held on December 4, 1980, and a re-exercise involving Grundy County on September 30, 1981 as described in Supplement No. 2 to our Safety Evaluation Report, have provided a demonstration of onsite preparedness to meet the requirements of 10 CFR Part 50, Appendix E.
Accordingly, we find the licensee is in compliance with License Condition Section 2.C.30(p)(1) of License No. NPF-11.
On January 26, 1982 and again on April 15, 1982, the licensee tested the prompt public notification system to ensure operability and adequacy of the system.
The second test was conducted during the April 14-15, 1982 exercise and was witnessed by the Federal Emergency Management Agency. We find that the licen-see is in compliance with License Condition 2.C.30(p)(2) of License No. NPF-11.
However, if the Federal Emergency Management Agency makes a finding relevant to these tests that indicate a design problem or other deficiencies, we shall per License Condition 2.C.30(p)(3) impose the provisions of 10 CFR 950.54(s)(2) to ensure that any significant deficiencies identified will be corrected.
The Federal Emergency Management Agency provided final findings of the status of offsite preparedness in a letter dated June 4, 1982.
On April 14-15, 1982, in conjunction with the NRC staff, the Federal Emergency Management Agency observed the exercise.
As indicated in the Federal Emergency Management Agency's final findings of June 4,1982 and as demonstrated during the April 14-15, 1982 exercise, the Federal Emergency Management Agency states that there is reasonable assurance that the appropriate protective measures can and will be applied offsite in the event of a radiological emergency and that the plans and preparedness for offsite protection near La Salle are adequate to protect the health and safety of the public.
Accordingly, we find that the licensee is in compliance with License Condition 2.C.30(p)(3) of License No. NPF-11.
La Salle SSER 4 22-1
The licensee has established an acceptable interim meteorological program and has provided the following mechanisms that would lead to their long term improvements:
(i) the licensee has installed a process computer with the capability to retrieve meteorological information that provides a redundant means for data access; (ii) the licensee has proposed a plan for meeting the meteorological and dose assessment capability guidance of NUREG-0654, including access from remote locations, and has committed to full operation of this capability by January 1, 1983; and (iii) the licensee in a letter dated, March 30, 1982, provided a descrip-tion of the dose calculational methodology with Class A transport and diffusion module and has committed to include this description in the radiological emergency plan.
The licensee has included in its radiological emergency plan, the meteorological measurement preventa-tive and corrective maintenance program.
Accordingly, we find that the licensee is in compliance with License Condition 2.C.30(p)(4) of License No. NPF-11.
La Salle SSER 4 22-2
APPENDIX A Continuation of Chronology of Radiological Review For the La Salle County Station April 17, 1982 Letter to licensee indicating Issuance of Facility Operat-ing License NPF-11 for Unit 1 not to exceed 5 percent power.
April 22, 1982 Letter from licensee concerning Comments and Clarifications on Meeting Transcript, March 31, 1982.
April 28. 1982 Letter from Illinois Friends of the Earth concerning 2.206 Request to Institute a Show Cause Proceeding.
April 29, 1982 Letter from licensee concerning Interpretation of Technical Specifications-Reporting Requirements when Special Reports are indicated in Action Statements.
May 3, 1982 Letter from Assistant Attcrney General of State of Illinois concerning Amendment to Request for Show Cause Proceeding.
May 3, 1982 Letter to licensee concerning Extension of Construction Completion Date for Unit No. 2.
May 3, 1982 Letter from licensee concerning Fire Stops for Non-segregated Phase Bus Duct Penetrations.
May 4, 1982 Letter from licensee concerning Completion of Vendor Review of Emergency Procedures.
May 4, 1982 Letter from licensee concerning Interpretation of Technical Specifications - Plant Staff Working Hours.
May 7, 1982 Letter to licensee concerning Revised Reactor Vessel Materials Surveillance Program Withdrawal Schedule for Unit No. 1.
May 7, 1982 Letter from licensee concerning Final Report on Allegations Regarding Rebar Damage and Off-Gas Roof Building Thickness.
May 7, 1982 Letter from licensee concerning Teledyne Open Item and Error / Deviation Reports for the La Salle Independent Design Review - First Transmittal.
May 13, 1982 Letter from licensee concerning Teledyne Open Item and Error / Deviation Reports for the La Salle Independent Design Review - Second Transmittal.
La Salle SSER 4 A-]
'I L
__ 9
3.
s
~.
May 14, 1982
~ Letter from licensee concerning Teledyne Open Item and Error / Deviation Reports for the La Salle Independent Design Review - Final Transmittal; and its 1st Transmittal on Responses.
May 18, 1982
' Letter from licensee transmitting Report on Additiona'l Allegations Regarding Rebar Damage.
May 20, 1982 Letter from 1!censee concerning Teledyne Open Item and Error / Deviation Reports for the La Salle Independent Design Review Responses to Remaining Items.
1 May 21, 1982 Letter to 1icensee concerning Interpretation of Technical Specification - Reporting Requirements when Special Reports are indicated in Action Statements.
May 21, 1982 Letter from licensee concerning Preoperational Fog and Ice Observation Program.
i May 24, 1982 Letter from licensee concerning Proposed Amendment to 1
NPF-11 Appendix A Technical Specifications.
l May 26, 1982 Letter from licensee concerning Teledyne Open Item and
[
Error / Deviation Re orts for the La Salle Independent Design Review, Supplemental Responses.
May 26, 1982 Letter from FEMA concerning Evaluation to the December 4, 1980 La Salle Exercise.
May 28, 1982
~
Letter from licensee concerning Teledyne Open Item and Error / Deviation Reports for the La Salle Independent Design Review, Supplemental Responses.
June 1, 1982 Letter from licensee concerning Drywell/Welwell Vacuum 8reakers.
' June.1, 1982 Letter from licensee concerning Proposed Amendment to NPF-11 Appendix A, Technical Specifications.
~ June 1, 1982 Letter from licensee concerning Final Report for the La Salle Independent Design Review.
June 1, 1982 Letter from FEMA indicating status of La Salle Formal Submission under 44 CFR 350 (proposed).
June 2, 1982 Letter to the Attorney General of Illinois responding to the Amendment to the Request for Show Cause Proceeding.
. June 2, 1982 Letter from licensee concerning Request for NRC Approval of N
Change in Acceptance Criteria for RCIC Startup Test.
June 7, 1982 Letter from licensee concerning Proposed Amendment to NPF-11 Appendix " Technical Specifications
,\\
La Salle SSER 4 A-2 L
l f
June 7, 1982 Letter to licensee concerning Technical Specifications Change Requested by the Licensee for Isolating LPCI in the Special Test.
June 7, 1982 Letter from licensee concerning Separation of Class 1E and non-Class 1E Cable Tray.
June 10, 1982 Letter from licensee concerning Forty-Year Operating License.
June 11, 1982 Letter from licensee concerning Drywell/Wetwell Vacuum Breakers.
June 14, 1982 Letter from licensee concerning Proposed Amendments to NPF-11 Appendix "A" Technical Specifications.
June 16, 1982 Letter from licensee concerning Security Plan Revisions.
June 17, 1982 Letter to licensee concerning Change in Acceptance Criteria for the Reactor Core Isolation Cooling System Startup Test for La Salle.
June 17, 1982 Letter from licensee concerning Independent Design Review on SRV/LOCA Hydrodynamic Loads Revised Design-Basis Summary Report.
June 18, 1982 Letter to licensee forwarding Amendment No. I to NPF-11.
June 18, 1982 Addendum to Technical Report No. TR-5539-2, Independent Design Review of La Salle, Unit 1.
June 21, 1982 Letter to licensee concerning Amendment No. 2 to Facility Operating License No. NPF-11.
June 24, 1982 Letter from licensee concerning Equipment Qualification of Safety-Related Electrical Equipment in Harsh Environments.
June 29, 1982 Letter from licensee concerning Compliance with Regulatory Guide 1.97.
June 30, 1982 Letter from licensee concerning Environmental Qualification.
July 2, 1982 Letter from licensee concerning Proposed Amendment to NPF-11 Appendix "A" Technical Specifications.
July 2, 1982 Letter to licensee concerning Concerns Regarding the Adequacy of the Design Margins of the Mark I and II Con-tainment Systems.
July 2, 1982 Letter to licensee concerning Control of Heavy Loads.
July 6, 1982 Letter to licensee concerning Request for Additional Information on Two Additional Instrumentation and Control Concerns.
La Salle SSER 4 A-3
I APPENDIX 8 PRESERVICE INSPECTION PROGRAM FOR UNIT No. 2 I.
INTRODUCTION For nuclear power facilities whose construction permits were issued on or after January 1,1971, but before July 1, 1974, 10 CFR 50.55a(g)(2) specifies that components shall meet the preservice examination requirements set for in editions of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and Addenda in effect six months prior to the date of issuance of the construction permit.
The provisions of 10 CFR 50.55a(g)(2) also state that the components (including supports) nay meet the requirements set forth in subsequent editions of this Code and addenda which are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein.
On September 1, 1981 and April 5, 1982, Commonwealth Edison submitted the La Salle Unit 2 preservice inspection plan and preservice inspection report based upon the 1974 Edition through the Summer 1975 Addenda of the American Society of Mechanical Engineers Code.
Specific written relief from Code requirements was requested and supported by information pursuant to 10 CFR 50.55a(a)(2)(i).
Therefore, our evaluation consisted of reviewing the licensee's submittal to the requirements of the 1974 Edition of Section XI through Summer 1975 Addenda and determining if deviations from the Code re-quirements were justified.
As a result of our review of this information, we have determined that certain preservice examinations are impractical and performing these required examina-tions would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.
Our basis for this conclusion is discussed in the subsequent paragraphs of this Appendix.
II. TECHNICAL REVIEW CONSIDERATIONS A.
La Salle Unit 2 received a construction permit on September 10, 1973.
In accordance with 10 CFR 50.55a, the preservice inspection program must conform to the 1971 Edition of the American Society Mechanical Engineers Code,Section XI through the Winter 1972 Addenda.
The American Society of Mechanical Engineers first published rules for inservice inspection in the 1970 Edition of Section XI.
No preservice or inservice inspection require-ments existed prior to that date.
Since the plant system design and ordering of long lead-time components were well underway by the time the Section XI rules became effective, full compliance with the exact Section XI access and inspectability requirements of the Code were not always practical.
Commonwealth Edison revised the preservice program based on the requirements of the 1974 Edition through Summer 1975 Addenda in consideration of the updating requirements of 10 CFR 50.55a(g).
La Salle SSER 4 B-1
B.
Verification of as-built structural integrity of the primary pressure boundary is not dependent on the Section XI preservice examination.
The applicable construction codes to which La Salle Unit 2 primary pressure boundary was fabricated contain examination and testing requirements which by themselves provide the necessary assurance that the pressure boundary components are capable of performing safely under all operating conditions reviewed in the Final Safety Analysis Report and described in the plant design specification.
As a part of these examinations, all of the primary pressure-boundary full penetration welds were volumetrically inspected (radiographed) and the system was subjected to hydrostatic pressure tests.
C.
The intent of the preservice examination is to establish a reference or i
baseline prior to the initial operation of the facility.
The results of subsequent inservice examinations can then be compared to the original condition to determine if changes have occurred.
If review of the inservice inspection results shows no change from the original condition, no action is required.
In the case where baseline data are not available, all indications must be treated as new indications and evaluated accordingly.
Section XI of the American Society Mechanical Engineers Code contains acceptance standards which may be used as the basis for evaluating the acceptability of such indications.
D.
Other benefits of the preservice examination include providing redundant or alternative volumetric inspection of the primary pressure boundary using a test method different from that employed during the component fabrication.
Successful performance of a preservice examination also demonstrates that the welds so examined are capable of subsequent inservice examination using a similar test method.
In the case of La Salle Unit 2, a large portion of the American Society of Mechanical Engineers Code requiring preservice examinations was performed.
We have concluded that failure to perform a 100 percent preservice examination of the welds identified below will not significantly affect the assurance of the initial structural integrity.
E.
In some instances where the required preservice examinations were not performed to the full extent specified by the applicable American Society l
of Mechanical Engineers Code, we will require that these or supplemental examinations be conducted as part of the inservice inspection program.
We have concluded that requiring these supplemental examinations to be performed at this time (before plant startup) would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.
The performance of supplemental examinations, such as surface examinations, in areas where volumetric inspection is difficult will be more meaningful after a period of operation.
Acceptable preopera-tional integrity has already been established by similar Section III fabrication examinations.
In cases where parts of the required examination areas cannot be effectively examined because of a combination of component design or current inspection technique limitations, we will continue to evaluate the development of new or improved volumetric examination techniques.
As improvements in these areas are achieved, we will require that these new La Salle SSER 4 B-2
t techniques be made a part of the inservice examination requirements of those components or welds which received a limited preservice examination.
III.
EVALUATION OF RELIEF REQUESTS The licensee requested relief from specific preservice inspection requirements for La Salle Unit No. 2 in the Preservice Inspection Plan and the Preservice Inspection Report of September 1, 1981. The licensee's request to use the requirements of subsequent editions and addenda of the Code has been evaluated and found to be acceptable.
These requests are identified in Section K of this Appendix.
Relief request 6 was withdrawn by Commo1 wealth Edison.
Evaluation of the remaining relief requests follows:
unless otherwise stated, references to the Code refer to the American Society of Mechanical Engineers Code,Section XI, 1974 Edition, including Addenda through Summer 1975.
Based on the information submitted by Commonwealth Edison and our review of the design, geometry, and materials of construction of the components, certain preservice requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI have been determined to be impractical and imposing these requirements would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.
Therefore, pursuant to 10 CFR Part 50, paragraph 50.55a(a)(2), our conclusions that these preservice requirements are impractical and that the relief requests are justified as provided below:
A.
All Class 1 and 2 Pressure Retaining Welds in Piping, Pumps and Valves Examination, Category B-J, Items 4.5 and 4.6 and Examination Categories C-F and C-G, Items C2.2 and C2.3 (Relief Request RI-04).
Code Requirement:
All indications which produce a response greater than 20 percent of the reference level shall be investigated to the extent that the operator can evaluate the shape, identity and location of all such reflectors in terms of the acceptance-rejection standards of the referenc-ing Code section.
Code Relief Request:
A deviation was requested from recording indications of 20 percent of reference (distant amplitude correction-DAC) sensitivity.
As an alternative examination, the licensee requests the use of a 50 percent DAC recording sensitivity.
Reason for Request: The geometry of the pipe weld does not allow meaning-ful ultrasonic recording sensitivity of 20 percent DAC.
An ultrasonic recording sensitivity of 50 percent DAC will provide a more meaningful and reliable means of detectiny defects in these welds.
The American Society of Mechanical Engineers,Section XI, Summer 1975 Addenda does not specifically address volumetric inspection of welds in piping, but references Article V of Section V.
The 1975 Section V pro-visions do not specifically address piping welds either, nor does it stipulate recording levels for weld indications.
Commonwealth Edison, La Salle SSER 4 B-3
u therefore, usesSection V to the extent applicable for pipe welds but uses Appendix I of the American Society of Mechanical EngineersSection XI, Summer 1975 Addenda, for data to be recorded.
This article sets 50 percent DAC as the minimum recording level.
Staff Evaluation:
Recording and evaluating indications between 20 and 50 percent DAC is difficult for the following reasons:
(a) The welded joints in nuclear piping frequently contain Code allowable wall thickness differences (12 percent of nominal thickness) as well as weld drop-through, counterbore taper, crown height, and other surface conditions which generate a large number of geometric reflec-i tors which produce ultrasonic test (UT) indications greater than 20 percent DAC.
(b) Weld meta' in stainless steel piping contains reflectors due to tne metallurgical structure which produce a large number of UT indica-l tions.
j Based upon the fabrication examination required by Section III, we conclude that a recording level of 50 percent is acceptable for the Preservice Inspertion.
However, for future inservice inspections, we have determined that recording at a 50 percent level is acceptable with the following conditions:
]
l (1) All indications 50 percent of DAC or greater shall be recorded.
(2) All indications 100 percent of DAC or greater shall be investi-gated by a Level II or Level III examiner to the extent neces-sary to determine the shape, identity, and location of the reflection.
(3) Any crack-like indication, 20 percent DAC or greater, discovered during a UT examination of piping welds and base metal materials shall be recorded and investigated by a Level II or Level III examiner to the extent necessary to determine the shape, identity, and location of the reflector.
(4) The owner shall evaluate and take corrective action for the disposition of any indication investigated and found to be nongeometric in nature.
Based on our review of the preservice inspection requirements, we conclude that imposition of exact Code requirements would result in hardships or unusual difficulties without a compensating increase in quality and safety since the initial preservice integrity has been demonstrated by the fabrication examinations.
B.
Circumferential Butt Welds in Primary Containment Penetrations, Category B-J, Item B4.5 (Relief Requests RI-05 and RI-11).
Code Requirement:
The volumetric examination performed during the inspec-tion interval shall cover all the area of 25 percent of the circumferential La Salle SSER 4 B-4
i joints including the adjoining 1 foot sections of longitudinal joints and 25 percent of the pipe branch connections.
The volumetric examinations shall cover 100 percent of the weld.
Code Relief Request:
A relief was requested from performing the entire volumetric examination required by the Code.
Reason for Request
Due to its design, the primary containment penetra-tion assembly leaves one pressure-retaining piping weld inaccessible for examination by either surface or volumetric methods.
A volumetric exami-nation consisting of a radiographic examination was performed during fabrication to meet Section III requirements.
These fabrication dot ments are available for audit at La Salle County Station.
The welds can only be examined by inspecting for evidence of leakage during system hydrotests.
The licensee proposed a visual examination for leakage during a system hydrostatic test as an alternative examination or a surface examination for the circumferential butt weld on the reactor core isolation cooling penetration.
Staff Evaluation:
The circumferential butt weld that is required by Code to be examined is physically inaccessible for inspection because of the existing design.
Durir, our review of the inservice inspection program, we will evaluate augmente( inspections to maintain the examination sample size of circumferential butt welds between the containment penetration sleeve and the fluid head fitting for the following penetration lines:
2FW02FA 2MS02BA 2MS02ED
- 1A*
2MS02BB
- lA*
2FWO2FB 2MS02BC
- 1B*
2MS02BD We have determined that part of the required Section XI examination is impractical because the existing geometric configuration limits the extent of the examination.
We conclude that the limited Section XI surface examination, the radiographic examination performed during fabrication, and the hydrostatic test demonstrates an acceptable level of preservice structurai integrity.
C.
American Society of Mechanical Engineers Class 2 Pipe Branch Connections Exceeding Four Inches in Diameter, Examination Category C-F, Item C2.3 (Relief Request RI-07).
Code Requirement:
The examination areas shall include essentially 100 percent of the longitudinal and circumferential welds and the base metal for one wall thickness beyond the edge of the weld.
Longitudinal welds shall be examined for at least one foot from the intersection with the edge of the circumferential weld selected for examination.
In case of pipe branch connections, the areas shall include the weld on the main pipe run, and at least two inches of the base metal along the branch run.
La Salle SSER 4 B-5
Code Relief Request:
A relief was requested from performing 100 percent of the Code required volumetric examination.
Reason for Reguest:
Some of the branch pipe connections on carbon steel I
piping are constructed with reinforcement saddles.
These saddles are j
fillet welded over the actual branch connection weld. A meaningful volumetric examination cannot be done on either the branch connection weld or the two saddle welds.
Staff Evaluation:
The branch pipe connection in question are physically inaccessible.
The pressure retaining weld required by IWC to be examined is totally covered by a reinforcement saddle.
The licensee is proposing to perform a surface examination of the saille fillet welds as an alter-nate examination.
l We have evaluated the design drawings for accessibility of the following l
IMS-2044-4 IMS-2044-17 IMS-2044-42 IMS-2044-5 IMS-2044-18 IMS-2044-50 IMS-2044-8 IMS-2044-38 IMS-2044-51 IMS-2044-9 IMS-2044-39 IMS-2044-52 IMS-2044-15 IMS-2044-40 IMS-2044-53 IMS-2044-16 IMS-2044-41 I
We have determined that examination of these welds to the extent required by Section XI is impractical due to the location of piping reinforcement saddles and the geometry of fillet welds attaching the saddles to the main pipe.
The licensee has performed surface examinations on all fillet attachment welds.
We conclude that the limited Section XI surface examination, the radi-ographic examinations performed during fabrication, and the hydrostatic test demonstrate an acceptable level of preservice structural integrity.
D.
American Society Mechanical Engineers Class 2 Pressure-Retaining Welds in Underground Piping, Examination Category C-F and C-G, Items 2.1, 2.2, and 2.3 (Relief Request RI-08).
i Code Reguirement:
Volumetric examination of circumferential butt welds and branch connections exceeding four inches diameter including the weld metal and base metal for one wall thickness by a sampling procedure defined by IWC-2520.
Code Relief Request:
A relief was requested from performing the volu-metric examination required by Code.
Reason for Request
The underground location of these welds makes a volumetric or surface examination impractical.
Staff Evaluation:
Section IWC-1230 of subsequent editions of Section XI that are referenced by 10 CFR Part 50.55a(b) recognizes the impracticality of examination of components encased in concrete. We have determined that La Salle SSER 4 B-6
even though IWC-1230 specifies concrete-encased components, its intent may be applied to components that are underground.
This request applies to the following lines:
2HP01A 2RI16A*2A*
The licensee will perform a hydrostatic test of these lines.
We have determined that the Code-required examination is impractical because of the underground location of these welds.
Imposing the Code requirement would result in hardships or unusual difficulties without a compensating increase in quality and safety because the preservice struc-tural integrity has been demonstrated by the fabrication examination.
E.
Closure Head Nozzle Inner Radius (NIR), Examination Category B-D, Item Bl.4 (Relief Request RI-09).
Code Requirement:
The extent of examination of each nozzle shall cover 100 percent of the volume to be inspected as shown in Figure IWB-2500D.
All nozzles shall be examined during each inspection interval.
Code Relief Request:
Request relief from performing 100 percent volu-metric examination requirements for nozzle-to-vessel welds N7, N8 and N18.
Reason for Request
The reactor pressure vessel (RPV) closure head is removed during refueling activities allowing access to the RPV closure head nozzle inner radius (NIR).
A surface examination is more sensitive in detecting surface defects at the NIR.
Approximately 40 man-hours are required to volumetrically inspect the 3 RPV closure head NIR and bore areas during inservice inspection.
This UT technique employs five specially contoured variable-angle lucite wedges which track both the radial and circumferential marking lines.
By comparison, only 4 man-hours are required to examine all these areas via liquid penetrant test (PT) techniques.
With the equivalent radiation levels, the UT exposure would be ten times the PT exposure for these inspections.
Staff Evaluation: We conclude that a surface examination of the nozzle inner radius region and bore areas is more sensitive for detecting service-induced defects than UT.
In addition, radiation exposure of inspection personnel in the future will be reduced using PT techniques.
Therefore, we find the surface examination an acceptable alternative examination.
F.
Pressure-Retaining Welds in Class 2 Piping Less than 0.250 Inch in Wall Thickness (Relief Request RI-10).
Code Requirement:
Volumetric examination of circumferential butt welds and branch connections exceeding four inches diameter including the weld metal and base metal for one wall thickness by a sampling procedure defined by IWC-2520.
Code Relief Request:
A relief was requested from performing the volu-metric examination required by Code.
La Salle SSER 4 B-7
Reason for Request
The wall thickness of the piping (less than 0.250 inch) makes meaningful contact pulse-echo ultrasonic examination im-practical.
These welds are in lines which draw pump suction from the condensate storage tanks with a maximum operating pressure of 30 pounds per square inch gauge and a maximum operating temperature of 120 degrees Fahrenheit.
Staff Evaluation:
The welds in question are on line 2RI16A which has a I
wall thickness of 0.165 inch.
The ultrasonic examination required by IWC-2500 cannot be expected to produce meaningful results. We conclude that the surface examination proposed by the licensee is an acceptable alternative examination that may be used in lieu of the required volu-metric examination.
G.
Recirculation Pump Internal Surface and Valve Internal Surface Visual Examination (Relief Requests RI-13 and RI-14)
Code Requirement:
(a) Examination Category B-L Visual examination of 1 of 2 pumps each 10 year interval.
(b) Examination Category B-M Visual examintion of 1 valve in each group of valves (as specified in Category B-M-2) each 10 year interval.
Code Relief Request:
A relief was requested to delete the visual examina-tion as required by Code.
Reason for Request
(a) Relief Request 13 - The recirculation pump casing material is cast stainless steel.
This material type has performed very well in nuclear service and has demonstrated substantial resistance to such chemical processes as pitting corrosion and stress corrosion cracking l
which would decrease the structural integrity of the pump.
To do the required visual inspection of this pump's internal surfaces, large amounts of radiation exposure and time would be required.
This was demonstrated at a similar nuclear station, where an expenditure of approximately 1000 man-hours and 50 man rem was required to complete the visual inspection of a similar designed pump.
The large expenditure of man-rem and man-hours to complete the visual inspection of this pump is impractical and not commensurate to the increased safety achieved by the inspection.
Commonwealth Edison believes that adequate safety margins are inherent in the basic pump design and that the health and safety of the public will not be i
adversely affected by not performing the visual examination of the pump casing internal surfaces solely for the purpose of inspection.
If, however, a pump requires disassembly for maintenance, then a visual inspection, to the extent practical, will be performed.
(b) Relief Request 14 - In Class 1 systems, there are 69 valves in 18 groups which are greater than 4 inches nominal pipe size.
The i
La Salle SSER 4 B-8
t requirement to disassemble an operable valve for the sole purpose of performing a visual examination of the internal pressure boundary is impractical and not commensurate to the increased safety achieved by the inspection and has only a very small potential of increasing plant safety margins with a very disproportionate impact on expen-ditures of plant manpower and radiation exposure.
Performing these visual examinations under such adverse conditions as high dose rate (10 R/hr) and poor as-cast surface condition provides little addi-tional information as to the valve casing integrity.
The performance of both carbon and stainless steel cast valve bodies has been excellent in all boiling water reactor applications.
Commonwealth Edison believes that adequate safety margins are in-herent in the basic valve design and that the health and safety of the public will not be adversely affected by not performing the visual inspections of the valve body internal surfaces solely for the purpose of inspection.
If, however, a valve requires disassembly for maintenance, then a visual inspection, to the extent practical, will be performed.
Staff Evaluation: We have determined that disassembly of these valves and pump at this time solely to perform the required Section XI preservice visual examination of internal surfaces is impractical. We conclude that the nondestructive examinations performed to date significantly exceed the Section XI visual requirements.
H.
Reactor Pressure Vessel Welds at the Vessel Bottom Head (Relief Request RI-15).
Code Requirement:
The volumetric examination performed during each inspection interval shall cover at least 10 percent of the length of each longitudinal weld and 5 percent of the length of each circumferential shell weld and head weld.
Code Relief Request:
A relief was requested from performing 100 percent of the Code volumetric examinations.
Reason for Request
These welds are partially or completely inaccessible due to the control rod drive housings.
A volumetric examination consisting of a radiographic examination was performed during fabrication to meet Section III requirements.
These fabrication documents are available at the vessel manufacturer for audit.
Staff Evaluation:
The welds in question are GEL-1061-DG and GEL-1061-DR on the vessel bottom head. A total of 28 inches was accessible and examined on both welds.
Complete volumetric examination of the subject welds is physically not possible due to the design and location of control rod drive penetrations from the exterior and core shroud plate from the interior.
We have determined that the required Section XI examination is impractical because the vessel design limits the extent of examination. We conclude that the limited Section XI examination, the radiographic examination
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La Salle SSER 4 B-9 j
performed during fabrication, and the hydrostatic test demonstrate an acceptable level of preservice structural integrity.
I.
Circumferential and Longitudinal Pipe Welds with Access Limitations (Relief Request RI-16).
Code Requirement:
3 (a) Examination-Category-B-J - The examination areas shall include essentially 100 percent of the longitudinal and circumferential welds and the base metal for one wall thickness beyond the edge of the weld.
Longitudinal welds shall be examined for at least one foot from the intersection with the edge of the circumferential weld selected for examination.
In the case of pipe branch connections, the areas shall include the weld metal, the base metal for one pipe wall thickness beyond the edge of the weld on the main pipe run, and at least 2 inches of the base metal along the branch run.
(b) Examination Category C-F - Volumetric examination of circumferential butt welds and branch connections exceeding 4 inches diameter in-cluding the weld metal and base metal for one wall thickness by a sampling procedure defined by IWC-2520.
Code Relief Request:
A relief was requested from performing 100 percent of the Code-required examination.
Staff Evaluation: We have evaluated the accessibility and inspectability of welds on the following lines:
2FW02EA 2LP028*1A*
2RH04A 2FW02EC 2LP06B 2RH19CB 2FW02EF 2MS01AB We have determined that examination of these welds to the extent required by the Code is impractical due to the design of the piping system and/or examinations on those areas which cannot be completely scanned by the ultrasonic. inspection.
We conclude that the limited Section XI examina-tions, the volumetric examinations performed during fabrication, and the hydrostatic test demonstrate an acceptable level of preservice structural integrity.
J.
Class 2 Pump Casing Welds (Relief Request RI-17)
Code Requirements:
The volumetric examination shall include 100 percent of the selected welds.
Code Relief Request:
A relief was requested from performing 100 percent of the volumetric examination required by Code.
Staff Evaluation:
The following pumps and associated welds are involved:
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La Salle SSER 4 B-10
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High Pressure Core Spray (HPCS) Pump Welds:
IHP-PU2-4 IHP-PU2-8A IHP-PU2-7A IHP-PU2-8B IHP-PU2-7B IHP-PU2-8C IHP-PU2-7C IHP-PU2-9 Low Pressure Core Spray (LPCS) Pump Welds:
ILP-PU2-4 ILP-PU2-8 ILP-PU2-7 ILP-PU2-9 Residual Heat Removal (RHR) Pump Welds:
Pump A IRH-PU2A-4 IRH-PU2A-8A IRH-PU2A-7A IRH-PU2A-8B IRH-PU2A-7B IRH-PU2A-8C IRH-PU2A-7C IHR-PU2A-9 Pump B IRH-PU2B-4 IRH-PU2B-8A IRH-PU2B-7A IRH-PU28-8B IRH-PU2B-7B IRH-PU2B-8C IRH-PU28-7C IRH-PU2B-9 Pump C IRH-PU2C-4 IRH-PU2C-8A IRH-PU2C-7A IRH-PU2C-8B IRH-PU2C-78 IRH-PU2C-8C IRH-PU2C-7C IRH-PU2C-9 Weld 4 on both the HPCS and LPCS pumps have limitations due to the suction flange and concrete base. We estimate the licensee examined 75 percent of these welds according to the Code. Welds 7, 8 and 9 on the HPCS and LPCS are totally inaccessible because the pump is encased in a concrete floor.
Weld 4 on the RHR pumps also have limitations due to suction flange and concrete floor.
Welds 7, 8 and 9 are totally inaccessible because the pump is encased in a concrete floor.
We have determined that examination of these welds to the extent required by Code is impractical due to pump design and configuration.
We conclude that the limited Section XI nondestructive examination plusSection XI pump testing required ensures an acceptable level of preservice integrity.
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La Salle SSER 4 B-11
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K.
Additional Relief Requests In addition to the relief requests evaluated in Section III, parts A through J, the licensee submitted three other requests for relief which involved updating examination requirements to the 1977 Edition through Summer 1978 Addenda of Section XI of the American Society of Mechanical Engineers Code.
Updating is permitted by 50.55a(g)(2), provided all of the related requirements of the respective editions or addenda are met.
We have evaluated relief requests submitted by the licensee and have determined that the relief requests are acceptable and in accordance with subsequent editions of Section XI referenced by 50.55a(g)(2):
Relief Request Examination Identification Category _
Component 1
B-K- 1 Welded supports I
2 C-D Bolting 3
IWC-1220(a)
All Class 2 pres-sure retaining components 12 C-F Fillet welds i
l l
IV.
CONCLUSIONS Based on the foregoing, we have determined, pursuant to 10 CFR 50.55a(a)(2),
that certain Section XI required preservice examinations are impractical, and i
I compliance with the requirements would result in hardships or unusual dif-ficulties without a compensating increase in the level of quality and safety.
Our technical evaluation has not identified any practical method by which the existing La Salle Unit 2 can meet all the specific preservice inspection requirements of Section XI of the American Society of Mechanical Engineers Code.
Requiring compliance with all the exact Section XI required inspections would delay the startup of the plant in order to redesign a significant number of plant systems, obtain sufficient replacement components, install the new components, and repeat the preservice examination of these components.
Examples of components that would require redesign to meet the specific preservice examination provisions are the reactor vessel, residual heat removal pumps, and a significant number of the piping and component support systems.
Even after the redesign effort, complete compliance with the preservice examination requirements probably could not be achieved.
However, the as-built structural integrity of the existing primary pressure boundary has already been established by the construction code fabrication examinations.
Based on our review and evaluation, we conclude that the public interest is not served by imposing certain provisions of Section XI of the American Society of Mechanical Engineers Code that have been determined to be impractical.
Pur-suant to 10 CFR 50.55a(a)(2), we have allowed deviations from these require-ments which are impractical to implement and would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
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La Salle SSER 4 B-12
l I
APPENDIX C This supplement to the Safety Evaluation Report was prepared by the NRC staff and its consultant.
The NRC staff members listed below were principal con-tributors to this report.
Included in this list is a consultant that was utilized by the NRC staff in its preservice inspection review.
NRC Staff Title Affiliation W. L. Axelson Section Chief Emergency Preparedness Section Region III R. A. Gramm Mechanical Engineer Auxiliary Systems Branch M. R. Hum Senior Materials Engineer Materials Engineering Branch B. Mann Senior Auxiliary System Auxiliary Systems Branch Engineer J. G. Spraul Senior Quality Assurance Quality Assurance Branch Engineer D. Terao Mechanical Engineer Mechanical Engineering Branch R. C. Van Niel Section Leader Emergency Preparedness Licensing Branch I. T. Yin Inspection Specialists Engi.eering Inspection Branch Consultant Name Organization T. Taylor Pacific Northwest Laboratories La Salle SSER 4 C-1 1
- U j
U.S NUCLE AR REGULATORY COMMISSION
""8 NUREG-0519
- BIBLIOGRAPHIC DATA SHEET Supplement No. 4 4 TITLE AND SU8 TITLE (Add Votume No.. dappropreare) 2.(Leave 6/m t)
Safety Evaluation Report related to the operation of La Salle County Station, Units 1 and 2
- 3. RECIPIENT *S ACCESSION No 7 CauTHORISl
- 5. DATE REPOH f COMPI( f tp jyEAR MONTH July 1982
- 9. PERF ORMING ORGANIZATION NAME AND M AILING ADD 9ESS (factude lip Codel DATE REPORT ISSUED uOsin jvEA.
U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation
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Washington, D. C.
20555 s it,,7, eian s 8 Iteave blankl
- 13. SPONSORING ORGANIZATION N AME AND MAILING ADDRESS (incluor 2,p Codel
- 10. PROJECT / TASK / WORK UNIT NO.
Same as 9 alove
- 11. CohTR ACT NO.
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- 13. TYPE OF REPORT PE RIOD COVE RED finclusere dates)
Safety Evaluation Report April 1982 to July 19fl2
- 15. SUPPLEMENT ARY NOTES
- 14. (teave nic A J Docket Nos. 50-373 and 50-374
- 16. ABSTR ACT (200 *ords or less)
Supplement No. 4 to the Safety Evaluation Report of Commonwealth Edison Company's application for licenses to operate its la Salle County Station, Units 1 and 2, located in Brookfield Township, La Salle County, Illinois has been prepared by the Office of Nuclear Reactor Regulation of the U. S. Nuclear Regulatory Commission.
On April 17, 1982, we issued a license, NPF-ll, to allow Unit 1 operation at power levels not to exceed 5 percent of rated power. This supplement addresses matters for proceeding to full power.
- 17. KE Y WORDS ANO DOCUMENT AN ALYSIS 8 7a DESCRIPTORS I ?b.10E N TI F IE RSc OPE N.E N DE D T E R MS SELURITY CLASS (Tn,s reportl 21 No OF P Af,ES 1"Uncl as si fied 18 AV AILABILITY ST ATEMENT Uniimited 20 SECURITY CL ASS (Th,s papf 22 PRICE Unclassified S
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