ML20212L153
| ML20212L153 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 08/31/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| TASK-***, TASK-RR NUREG-0827, NUREG-0827-S01, NUREG-827, NUREG-827-S1, NUDOCS 8608250229 | |
| Download: ML20212L153 (46) | |
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NUREG-0827 Supplement No.1 Integrated Plant Safety Assessment Systematic Evaluation Program La Crosse Boiling Water Reactor Dairyland Power Cooperative Docket No. 50-409 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation August 1986
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Availability of Refermce Materials Cited in NRC Pub liutions Most documents cited in NRC publicatio'ns wi i be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.VI.
Washington, DC 20555 s 1 4
- 2. The Superintendent of Dochments, U.S. Government Printing 0ffice, Post Of fice Box 37082, Washington, DC 20013-7082 W
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, 3. The National Technical Information Service, Springfield NA 22161 Although the listing that follo' w s represents the majority of documents cited in NRC publications, it is not intended to be exhaustive. <
Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoran'da; 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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l NUREG-0827 Supplement No.1 Integrated Plant Safety Assessment Systematic Evaluation Program La Crosse Boiling Water Reactor Dairyland Power Cooperative Docket No. 50-409 E
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation August 1986 p= ~e 9 N. ()
ABSTRACT The Nuclear Regulatory Commission (NRC) has published Supplement No. I to the Final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0827), under the scope of the Systematic Evaluation Program (SEP), for Dairyland Power Cooperative's La Crosse Boiling Water Reactor (LACBWR), located in Vernon County, Wisconsin. The SEP was initiated by the NRC to review the design of older operating nuclear power plants to reconfirm and document their safety.
This report documents the review completed under the SEP for those issues that required refined engineering evaluations or the continuation of ongoing evaluations after the Final IPSAR for the LACBWR plant was issued. The review has provided for (1) an assessment of the significance of differences between current technical positions on selected safety issues and those that existed when the LACBWR plant was licensed, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety when the supplement to the Final IPSAR and the Safety Evaluation Report for converting the license from a provisional to a full-term license have been issued. The Final IPSAR and its supplement will form part of the bases for considering the conversion of the provisional operating license to a full-term operating license.
~
La Crosse SEP iii
CONTENTS Page ABSTRACT...;......................................................... iii ACRONYMS AND INITIALISMS............................................. vii 1 INTR 00VCTION................................................... 1-1 2 TOPICS THAT REQUIRED REFINED ENGINEERING ANALYSIS OR CONTINUATION OF ONG0ING' EVALUATION............................. 2-1 2.1 Topic III-1, Classification of Structures, Components, and Systems................................................... 2-1 2.2 Topic III-2, Wind and Tornado Loadings; Topic III-4.A, Tornado Missiles; and Topic III-6, Seismic Design C o n s i de ra ti o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.3 Topic III-3.A, Effects of High Water Level on Structures................................................ 2-2 2.3.1 Containment Stability.............................. 2-2 2.3.2 Stack Stability.................................... 2-2 2.4 Topic III-4.B Turbine Missiles........................... 2-3 2.5 Topic III-5.A, Effects of Pipe Break on Structures, Systems, and Components Inside Containment......................... 2-3 2.6 Topic III-5.B, Pipe Break Outside Containment............. 2-4 2.7 Topic III-7.B, Design Codes, Design Criteria, Load Combinations, and Reactor Cavity Design Criteria.......... 2-5 2.8 Topic VI-4, Containment Isolation System.................. 2-5 2.8.1 Shutdown Condenser Atmospheric Vent (Penetration M-34)................................. 2-5 2.8.2 Shutdown Condenser Sample Line. . . . . . . . . . . . . . . . . . . . . 2-6 2.8.3 Was te Wate r Li nes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-6 2.8.4 Shutdown Condenser Offgas Vent (Penetration M-19)................................. 2-6 2.9 Topic VI-7.C.1, Appendix K - Electrical Instrumentation and Control Re-Reviews.................................... 2-7 2.10 Topic VII-1.A, Isolation of Reactor Protection System From Nonsafety Systems, Including Qualification of Isolation Devices................................................... 2-7 2.11 Topic VIII-3.B, DC Power System Bus Voltage Monitoring and Annunciation............................... 2-8 2.12 Topic IX-5, Ventilation Systems........................... 2-9 2.12.1 011 Vapors in Oil Storage Room..................... 2-9 2.12.2 Emergency Diesel Generator 1A Ventilation System................................. 2-9 La Crosse SEP v
. _ . . __-______ L
CONTENTS (Continued)
P_ag_e 3 TOPICS RESOLVED BY CHANGES TO PLANT TECHNICAL SPECIFICATIONS. . . 3-1 3.1 Topic II-1.A Exclusion Area Authority and Control......... 3-1 3.2 Topic II-3.B.1, Capability of Operating Plants To Cope With Design-Basis Flooding Conditions. . . . . . . . . . . . . 3-1 3.3 Topic II-3.C, Safety-Related Water Supply (Ultimate Heat Sink)...................................... 3-1 3.4 Topic V-12. A, Water Purity of BWR Primary Coolant................................................... 3-2 3.5 Topic VI-6, Containment Leak Testing...................... 3-2 3.6 Topic VI-7.A.3, Emergency Core Cooling System Actuation System.......................................... 3-3 4 IPSAR TOPIC RESOLUTIONS CONFIRMED BY THE NRC REGIONAL 0FFICE................................................ 4-1 4.1 Shutdown Condenser Shell-Side Level Control............... 4-1 4.2 Offgas Vent From Shutdown Condenser....................... 4-1 5 REFERENCES..................................................... 5- 1 Appendix A - REFERENCES TO CORRESPONDENCE FOR EACH TOPIC EVALUATED Appendix B - NRC STAFF CONTRIBUTORS LIST OF TABLES 2.1 Summary of IPSAR and Supplement Evaluations.................... 2-11 3.1 Modifications to La Crosse Technical Specifications as a Result of SEP......................................................... 3-3 4.1 Items for Confirmation by NRC Regional Office. . . . . . . . . . . . . . . . . . 4-2 La Crosse SEP vi
ACRONYMS AND INITIALISMS ac alternating current ASME American Society of Mechanical Engineers BTP Branch Technical Position BWR boiling-water reactor CFR Code of Federal Regulations dc direct current DPC Dairyland Power Cooperative FSAR Final Safety Analysis Report FTOL full-term operating license GDC General Design Criterion (a)
HPCS high pressure core spray IEEE Institute of Electrical and Electronics Engineers IPSAR Integrated Plant Safety Assessment Report LACBWR La Crosse Boiling Water Reactor NRC U.S. Nuclear Regulatory Commission POL provisional operating license RCPB reactor coolant pressure boundary RG Regulatory Guide RPS reactor protection system SEP Systematic Evaluation Program SER Safety Evaluation Report SRP Standard Review Plan TS Technical Specifications UHS ultimate heat sink La Crosse SEP vii
4 INTEGRATED PLANT SAFETY ASSESSMENT REPORT SUPPLEMENT NO. 1 SYSTEMATIC EVALUATION PROGRAM LA CROSSE BOILING WATER REACTOR 4
1 INTRODUCTION The Systematic Evaluation Program (SEP) was initiated by the U.S. Nuclear i Regulatory Commission to review the designs of older operating nuclear power plants to reconfirm and document their safety. The review provides (1) an assessment of the significance of differences between current tecnnical posi-
- tions on safety issues and those that existed when a particular plant was licensed, (2) a basis for deciding on how these differences should be resolved j in an integrated plant review, and (3) a documented evaluation of plant safety.
The initial review of the La Crosse plant as part of the SEP was published in ,
- NUREG-0827, the Integrated Plant Safety Assessment Report (IPSAR), dated June
- 1983. The review compared the as-built plant design with current review cri-teria in 137 different areas defined as " topics." During the review, 54 of the
, topics were deleted from consideration in the SEP because a review was being j conducted under other programs.(Unresolved Safety Issues or Three Mile Island i Action Plan tasks), the topic was not applicable to the La Crosse plant, or the items to be reviewed under that topic did not exist at the site.
! Of the original 137 topics, 83 were, therefore, reviewed for La Crosse; of
? these, 52 met current criteria or were acceptable on another defined basis.
! The review of'the 31 remaining topics found that certain aspects of plant design
! differed from current criteria. These topics were considered in the integrated assessment of the plant, which consisted of evaluating the safety significance and other factors of the identified differences from current design to arrive
~
at decisions on whether modification was necessary from an overall plant safety
! viewpoint. To arrive at these decisions, engineering judgment was used as well as the results of a limited probabilistic risk assessment study.
l In general, the staff's position in the integrated assessment fell into one
! or more of the following categories: (1) equipment modification or addition, <
(2) procedure development or Technical Specification changes, (3) refined engi-
- neering analysis or continuation of ongoing evaluation, and (4) no modification i necessary. Table 4.1 of the IPSAR summarizes the staff's integrated assessment
! positions and documents the licensee's agreement with those positions.
For those positions classified as either Category (1) or (2), Table 4.1 of the l IPSAR lists the scheduled completion dates agreed upon by the staff and the l
licensee. The NRC Region III Office provides verification of the implementa-
=
tion of these positions.
For those positions classified as Category (3), the licensee has provided the i results of the ongoing evaluation to the staff for review. The purpose of this j supplement to the IPSAR is to provide the staff's evaluation of the Category (3)
La Crosse SEP 1-1
issues and to summarize the status of all actions to be implemented as a result of the IPSAR and this supplement to the IPSAR.
The La Crosse plant is one of the seven SEP plants that had not been issued its full-term operating license (FTOL). A Safety Evaluation Report (SER) to sup-port the conversion of the provisional operating license (POL) to an FTOL will be prepared. The SER will consist of the IPSAR Supplement, a consideration of major plant modifications that have been made and the substantive regulations adopted since the POL was issued, and the Unresolved Safety Issues and Three Mile Island Action Plan tasks. (For the latter, the issues that were deleted ,
from consideration in the SEP review will be discussed in the SER.)
La Crosse SEP 1-2
l 2 TOPICS THAT REQUIRED REFINED ENGINEERING ANALYSIS OR CONTINUATION OF ONGOING EVALUATION The licensee has submitted an evaluation for each of the issues that required refined engineering analysis or further evaluation. The staff reviewed these submittals and concluded that either the licensee evaluation met current cri- i teria, the evaluation was acceptable on another defined basis, a corrective action will be required, or further analysis will be required. Factors con-sidered in reaching this conclusion include the perceived safety significance of the difference from current licensing criteria, a qualitative assessment of the financial and exposure costs to make a modification, and, to a lesser extent, implementation impact and schedule. The evaluation of these issues also considered any applicable risk perspectives, developed for the integrated assessment and described in the IPSAR, and related corrective actions proposed by the licensee as part of the integrated assessment or as a result of the follow-on evaluations.
A brief discussion of each of the outstanding issues is presented below. Each evaluation references the more detailed licensee evaluation and staff topic evaluation. References for correspondence pertaining to safety evaluation reports for each section appear in Appendix A. Appendix B is a listing of the staff contributors.
The final status of each of these issues is summarized in Table 2.1 along with the status of all other SEP issues for the La Crosse Boiling Water Reactor.
2.1 Topic III-1, Classification of Structures, Components, and Systems (NUREG-0827, Section 4.5)
As implemented by Regulatory Guide (RG) 1.26, General Design Criterion (GDC) 1 of Appendix A to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50) requires that structures, systems, and components important to safety be designed, fabricated, erected, and tested to quality standards com-mensurate with the importance of the safety functions to be performed.
In Section 4.5 of the IPSAR, the staff concluded that insufficient information existed to complete the topic review in certain areas and recommended that the licensee supply additional information and analyses on the following subjects:
(1) radiography (2) fracture toughness (3) piping (4) valves (5) pumps (6) storage tanks The staff's position in the IPSAR was that the licensee should complete the evaluations and incorporate the results in the Final Safety Analysis Report (FSAR) update. If the results indicate that modifications are required, those actions should be reported to the staff. Section 3.9 of the updated FSAR i La Crosse SEP 2-1
describes the radiography that was performed and the inservice inspection and testing programs. No inadequate components have been identified.
- 2. 2 Topic III-2, Wind and Tornado Loadings (NUREG-0827, Section 4.6);
Topic III-4. A, Tornado Missiles (NUR EG-0827, Section 4.9); and Topic III-6, Seismic Design Considerations (NUREG-0827, Section 4.13)
In the IPSAR, these issues were identified as requiring further evaluation. To resolve these issues, the licensee proposed to perform a study to determine the consequences-of system failures and the effects if the systems are not modified to resist the seismic event. A cost / benefit assessment would be made to evalu-ate modifications needed to withstand the wind / tornado hazard at the 10 4/ year and 10 5/ year frequency. The licensee proposed to determine a method for decay heat removal that will not be affected by tornado missiles and seismic events.
By letter dated October 11, 1984, the licensee submitted the integrated con-sequence study. Additional information was provided by letter dated March 15, 1985. The analysis of the two ventilation stacks was provided on April 1, 1985. ,
The basic approach proposed by the licensee is to use the stored volume of water in the overhead storage tank (inside containment) as a short-term source of shutdown condenser makeup. Long-term supply would be provided via fire pumper trucks or from the emergency service water supply system (from the river). Some emergency core cooling capability is also being seismically qualified. Plant modifications and procedure changes are needed to implement this approach. This issue remains under staff review; conclusions will be pro-vided in a separate SER.
j 2.3 Topic III-3.A, Effects of High Water Level on Structures (NUREG-0827, 4 Section 4.7) 10 CFR 50 (GDC 2), as implemented by Standard Review Plan (SRP, NUREG-0800)
Sections 3.4 and 3.8 and RG 1.102, requires, in part, that plant structures i be designed to withstand the effects of flooding.
! Two issues were identified in Section 4.7 of the IPSAR.
J 2.3.1 Containment Stability l The licensee provided, by letter dated February 14, 1983, the calculated fac-tors of safety against gross sliding and overturning for the containment. The factors met staff criteria; however, the staff had not completed its review i of the analysis assumptions and techniques at the time of issuance of the IPSAR.
Further information was provided in letters dated September 29, 1983,and July 3, 1984. By letter dated September 6, 1985, the staff issued its safety evalua-tion, which concluded that the licensee's analysis was acceptable.
l 2.3.2 Stack Stability l
The IPSAR noted that the licensee proposed to review stack stability for flood-ing load combinations for both the Genoa Unit 2 (LACBWR) and Genoa Unit 3 stacks.
The results showed that the minimum factor of safety against sliding and over-i turning was 1.7, which is higher than the SRP criterion of 1.5. Therefore, the i
! La Crosse SEP 2-2 I
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! staff concluded that stack stability under flood conditions has been demonstrated.
The staff's evaluation was issued by letter dated September 6,1985.
2.4 Topic III-4.B. Turbine Missiles (NUREG-0827, Section 4.10) 10 CFR 50 (GDC 4), as implemented by SRP Section 2.2.3 and RG 1.115, requires, in part, that structures, systems, and components important to safety shall be appropriately protected from dynamic effects, including the effects of missiles.
In Section 4.10 of the IPSAR, the staff determined that the licensee should compare its program for valve testing and testing of the overspeed protection for both Genoa Unit 2 (LACBWR) and Genoa Unit 3 with staff criteria and either justify the adequacy of the existing program or propose corrective measures.
In letters dated September 29, 1983, April 4, 1984, and June 7, 1984, the li-censee provided information concerning the testing at both units. The staff's evaluation of the proposed inspection program is provided in a letter dated April 19, 1985.
The staff concluded that the inspection and testing program proposed by the licensee for the LAC 8WR turbine meets the intent of staff criteria for protec-tion against turbine overspeed failures. The program consists of (1) periodic disassembly and inspection of the protective valves, (2) exercising the valves during normal operation, and (3) testing of the overspeed trip system.
l The inspection frequency is somewhat longer than the staff-recommenced inter-vals; however, on the basis of the satisfactory results of previous inspec-tions, the staff found the proposal acceptable.
The inspection program for the Genoa Unit 3 turbine satisfies the staff criteria and, therefore, is acceptable. This issue is considered resolved.
2.5 Topic III-5.A Effects of Pipe Break on Structures, Systems, and Components Inside Containment (NUREG-0827, Section 4.11) 10 CFR 50 (GDC 4), as implemented by SRP Sections 3.6.1 and 3.6.2 and Branch Technical Positions (BTPs) MEB 3-1 and ASB 3-1, requires, in part, that struc-tures, systems, and components important to safety be appropriately protected from the dynamic effects of postulated pipe ruptures. , ,
In the topic evaluation dated November 8, 1982, the staff identified six items for which further evaluation was required. These items are also described in the IPSAR, NUREG-0827.
4 By letters dated December 29, 1983, June 29, 1984, and July 26, 1985, the li-l censee responded to the open items. The staff's final evaluation closing out this IPSAR section was issued on August 13, 1985.
In that evaluation, the staff concluded that the six issues from Section 4.11 of the IPSAR were resolved by the following actions:
La Crosse SEP 2-3
(1) Jet impingement on mitigation systems - Two branch lines connected to the high pressure core spray (HPCS) line might be damaged by jet impingement from a break in the alternate core spray line. The licensee will reroute these two lines.
(2) Containment integrity No high energy lines are located in the area where the steel containment does not have concrete shielding. This issue is resolved.
(3) Decay heat cooling system blowdown line - In the IPSAR, the licensee committed to relocate a valve in this line and to establish procedures to close the valve in the event of an accident requiring containment isola-tion (such as a loss-of-coolant accident). These actions have been completed.
(4) Control rod drive mechanisms - The licensee's analysis demonstrates that stresses on the mechanisms from jet impingement are within ellowable limits. This issue is resolved.
(5) Containment ventilation exhaust damper operators - The licensee's analysis shows that the operators can withstand the jet impingement force and per-form their safety function. This issue is resolved.
(6) Pipe whip restraint anchor bolts - In response to staff questions, the licensee reevaluated the adequacy of the pipe whip restraints for the alternate core spray system. This analysis shows that the anchor bolts can withstand the dynamic loads from the pipe whip. This issue is resolved.
2.6 Topic III-5.B. Pipe Break Outside Containment (NUREG-0827, Section 4.12) 10 CFR 50 (GDC 4), as implemented by SRP Secticns 3.6.1 and 3.6.2 and BTPs MEB 3-1 and ASB 3-1, requires, in part, that structures, systems, and compo-nents important to safety be designed to accommodate the dynamic effects of postulated pipe ruptures.
The safety objective of this topic review is to ensure that if a pipe should break outside the containment, the plant can be safely shut down without a loss of containment integrity.
In the IPSAR, the staff noted it could not determine what pipe whip damage cri-teria and jet impingement model were used in the licensee's SRP reevaluation of the effects of pipe break outside containment. The questions raised by the staff related mainly to the orientation of the jet resulting from rupture of the main steam and bypass lines on the mezzanine floor of the turbine building.
The licensee, in a letter dated October 6, 1983, indicated that the lines in question have been restrained to prevent pipe whip from damaging structures and equipment. In addition, the movement of the pipe is constrained and the path of the jet will not impinge on any safety-related equipment.
The staff, in a letter dated October 27, 1983, found that information provided by the licensee fully resolved this issue.
La Crosse SEP 2-4
2.7 Topic III-7.B. Desian Codes, Desicn Criteria, Load Combinations, and Reactor Cavity Design Celteria (NLREG-0827, Section 4.14) 10 CFR 50 (GDC 1, 2, and 4), as implemented by SRP Section 3.8, requires that structures, systems, and ccmponents important to safety conform to applicable codes and standards. In the IPSAR, the staff identified 34 specific areas where design code changes are potentially applicable to La Crosse for which the current code requires substantially greater margins than the earlier codes or where no original code provision existed.
In letters dated December 29, 1983, and December 20, 1984, the licensee discussed these 34 areas. One case was identified by the licensee for more detailed anal-ysis. The structural element under consideration is the main steamline pene-tration and pipe whip restraint, which is subject to cyclic loads. As a result of applying the design requirements of Section III (1980), Article NE-3331(b) of the American Society of Mechanical Engineers, " Bailer and Pressure Vessel Code" (ASME Code), an overstress condition between the seal plate and pipe was identified. The licensee has agreed to further analyz's this case to develop a suitable modification to the penetration.
This issue remains under staff review; staff conclusions will be provided in a separate SER.
2.8 Topic VI-4, Containment Isolation System (NUREG-0827 Section 4.21) 2.8.1 Shutdown Condenser Atmospheric Vent (Penetration M-34) 10 CFR 50 (GDC 57), as implemented by SRP Section 6.2.4 and RG 1.141, requires that lines that penetrate the primary containment boundary and are neither part of the reactor coolant pressure boundary nor connected directly to the contain-ment atmosphere be provided with at least one locked-closed, remote manual, or automatic isolation valve outside containment.
In the topic evaluation, dated December 9, 1982, the staff concluded that the isolation provisions for the shutdown condenser atmospheric vent penetration do not meet the criteria of GDC 57. Remote manual isolation valves are available inside containment to shut off high pressure service water and domineralized water; however, to meet current criteria, an isolation valve would be required on the vent line that goes outside containment.
In the IPSAR, the staff concluded that addition of an isolation valve was not warranted. However, the staff concluded that the licensee should review the radiation monitoring system and plant procedures to ensure that corrective actions could be taken in the event of a tube rupture so that radiological releases would be less than a small fraction of that specified in 10 CFR 100.
The licensee, in letters dated December 29, 1983 (LAC-9514), and April 3, 1984 (LAC-9797), provided an analysis of the capability to detect a radiation leak through the shutdown condenser. The conclusion was that the design is accept-able provided the radiation alarm setpoint was reduced to 5 mrem /hr above back-ground. By letter dated April 26, 1984, the staff issued a safety evaluation report that accepted the licensee's analysis. The licensee has reduced the setpoint; therefore, this issue is resolved.
La Crosse SEP 2-5
l j- 2.8.2 . Shutdown Condenser Sample Line i :
10 CFR 50 (GDC 56), as implemented by SRP Section 6.2.4, requires that lines that connect directly to the containment atmosphere and that penetrate the containment shall be provided with containment-isolation valves as specified, j unless it can be demonstrated that the containment isolation provisions for the ;
4 line are acceptable on some other defined basis.
The 1/2-in.-diameter shutdown condenser sample line is equipped with two manual valves inside the containment that are normally locked closed. The sample is j taken inside the containment and the line does not exit the containment..
) However, should sampling be in progress when a event occurs which pressurizes the containment, the person taking the sample may not be able to shut the valves -
before leaving the containment. If the containment is subsequently pressurized j above the head of water in the line, a direct path to the environment (into shell side of condenser and out-the nonisolable vent line) would exist. As
! discussed in a February 8, 1985, letter, the licensee has agreed to install two check valves in series in this line to provide the required isolation should the scenario occur. By letter dated March 10, 1936, the staff concluded that such a valve configuration provides acceptable isolation for this line. The valves were installed during the 1986 refueling outage.
- 2.8.3 Waste Water Lines j
In Section 4.21.3.3 of the IPSAR, the penetrations for the waste water lines '
4 were identified where manual valves outside containment should be maintained I normally locked closed for containment isolation.
i i By letter dated March 28, 1984, the licensee described a design change for the waste water system. As a result, penetration M-22 (manual valve 54-24-179) will be cut and isolated from the containment atmosphere. For penetration M-27, i
a different manual valve has been chosen as the isolation boundary. This valve, 4 54-24-108, is located in a lower radiation area than the valve originally con-i sidered in the IPSAR. This valve will be open only under controlled procedures
! for waste water transfer. The isolation configuration for penetration M-25 has 2
been changed as described in the IPSAR; that is, manual valve 54-24-162 is now maintained normally locked closed.
l Therefore, the staff concludes that the isolation provisions for the waste j water lines are acceptable.
i j 2.8.4 Shutdown Condenser Offcas Vent (Penetration M-19) L In Section 4.21.3.2 of the IPSAR, the licensee agreed.to lock close two manual 1
valves on branch lines inside containment for this penetration by the end of the refueling outage. This' action was completed during the refueling outage as i
, discussed in an April 8, 1985, letter from the licensee. In addition, a remotely j operated solenoid valve in the pipe tunnel was to be added at the same time, j As discussed in the April 8, 1985, letter, the entire line penetration needs to i be upgraded to a higher pressure rating system to install this valve. The li- "
j censee, therefore, requested that this action be deferred to the next refueling 3 outage. In an August 29, 1985, submittal, the licensee described the planned
- modifications and schedule. The staff considers this schedule acceptable, i
! La Crosse SEP 2-6 i
j' i__._.____ _ _ _ _ _ _ _ _ _
The upgrading of the line, i,ncluding replacement of manual valves, has been completed. The remote manual isolation valve will be installed during the next outage of sufficient duration.
2.9 Topic VI-7.C.1, Appendix K - Electrical Instrumentation and Control Re-Reviews (NUREG-0827, Section 4.24) 10 CFR 50 (GDC 17), as implemented by SRP Sections 8.2 and 8.3 and RG 1.6, requires that redundant load groups and the redundant standby electrical power sources be independent. If means exist for manually connecting redundant load groups, at least one interlock should be provided to prevent an operator error that would parallel the standby power sources.
In Section 4.24.1 of the IPSAR, the staff identifieo a concern that 480-V buses 1A and 1B could be manually connected. Originally, a modification was proposed to provide an interlock to prevent the cross-tie breakers from closing if the diesel generators are supplying the essential buses.
In a letter dated August 31, 1983, the licensee noted that one of the bus tie breakers trips automatically as a part of the loading sequence for diesel generator 18. Furthermore, operating procedures prohibit closing either of the two manual breakers except during cold shutdown and refueling outages. There-fore, by letter dated October 7, 1983, the staff concluded that the present design is acceptable and no further modifications (such as the interlock men-tioned above) are required.
2.10 Topic VII-1.A, Isolation of Reactor Protection System From Nonsafety Systems, Including Qualification of Isolation Devices (NUREG-0827, Section 4.26) 10 CFR 50.55a(h), through Institute of Electrical and Electronic Engineers (IEEE) Std. 279-1971, requires that safety signals be isolated from non-safety signals so that failures in the non-safety instrumentation will not affect required safety-related functions.
In Section 4.26.1 of the IPSAR, the staff concluded that the licensee should demonstrate that a single failure in any process recorder would not disable the reactor protection system (RPS). In Section 4.26.3 of the IPSAR, the staff deter-mined that RPS channels should be shown to have adequate isolation from each other.
By letter dated September 29, 1983 (LAC-9324), the licensee proposed the follow-ing modifications:
(1) Provide a separate ac source for each recorder amplifier for iritermediate range flux channels 3 and 4.
(2) Provide qualified isolation devices between the outputs from power range channels 5, 6, 7, and 8 and their recorders.
(3) Provide separate, redundant power supplies for two channels each of the source range, intermediate range, pressure, and power-to-flow channels.
La Crosse SEP 2-7
(4)~ For power range channels 5, 6, 7, and 8, provide independently fused power to each automatic gain circuit and provide an overvoltage alarm on the IB inverter ac output.
By letter dated October 27, 1983a, the staff concluded that the above modifi-cations are sufficient to resolve these issues.
By letter dated January 3, 1985, the licensee modified the completion schedule for the separation of power supplies for the power-to-flow channel to the 1986 refueling outage. The remaining items were completed during the 1985 outage.
The reason for the delay for the power-to-flow channel was to avoid overloading the 1A static inverter; this inverter is being replaced with a larger one dur-ing the 1986 outage, and the power-to-flow channel could then be added as a load. The staff agreed to this schedule change in a letter dated May 16, 1985.
Subsequently, by letter dated November 14, 1985, the licensee requested that the change for the power-to-flow channels be deferred to the 1987 outage. The nuclear instrumentation at LACBWR is being replaced; however, the new equipment will not be available during the 1986 outage. Coordination of modifications to the power supplies for the power-to-flow channels with changes to the power-to-flow logic circuitry resulting from the instrumentation replacement will sim-plify the modification process. The staff accepted the revised completion date by letter dated December 23, 1985.
2.11 Topic VIII-3.8, DC Power System Bus Voltage Monitoring and Annunciation (NUREG-0827, Section 4.28) 10 CFR 50.55a(h), through IEEE Std. 279-1971, and 10 CFR 50 (GDC 2, 4, 5, 17, 18, and 19), as implemented by SRP Section 8.3.2, RGs 1.6, 1.32, 1.47, 1.75, 1.118, and 1.129, and BTP 1C58-21, require that the control room operator be given timely indication of the status of the batteries and their availability under accident conditions.
As discussed in Section 4.28 of the IPSAR, the La Crosse control room does not have some of the alarms and int'! cations for dc power systems that are recom-mended by staff guidance.
The licensee proposed to evaluate the present and NRC proposed de system moni-toring to determine whether plant modifications are warranted. By letters dated August 31, 1983, December 19, 1983, and January 22, 1985, the licensee responded to this issue. The staff's evaluation of the licensee's proposal was issued on April 24, 1985.
The staff concluded that, in conjunction with changes already made, the follow-ing modifications proposed by the licensee are sufficient to resolve this issue:
(1) Replace two " energize to trip" relays with "deenergize to trip" relays.
(2) Transfer power supply for two breakers from generator plant bus to other dc buses.
These changes, which were completed during the 1985 outage, will ensure that a single failure of a de bus will not prevent both trains of ac power from auto-matically functioning if required.
La Crosse SEP 2-8 i
~
2.12 Topic IX-5, Ventilation Systems (NUREG-0827, Section 4.29) 2.12.1 Oil Vapors in Oil Storage Room 10 CFR 50 (GDC 4), as implemented by SRP Section 9.4.4, requires that systems and components important to safety be designed to accommodate the effects of and be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.
In the topic evaluation, dated October 28, 1982, the staff concluded that the oil storage room ventilation system, a subsystem of the turbine building venti-lation system, would fail as a result of a loss of offsite power. Although normally not considered a safety-related system, its loss could allow a poten-tially explosive mixture of vapors to collect, and the effects on plant safety of the detonation have not been evaluated.
4 The licensee, in letters dated December 29, 1983, and February 9, 1984, provided the results of its examination of possible sources of volatile vapors and the potential ignition of these vapors.
The licensee determined that, of all vapor sources in that room, No. 2 fuel oil has the lowest flash point - approximately 100 F. The flash point is defined as the lowest temperature at which vapor pressure of the liquid is just suf-ficient to provide a flammable mixture and an ignition source is necessary for combustion. The licensee has further determined that all No. 2 fuel oil vapors are vented to the outside, thus preventing any accumulation within this room.
With respect to the remaining vapor sources, their flash points are much higher than temperatures that reasonably could be expected to occur even with a loss of room ventilation.
In addition, the licensee indicated that the oil storage room's design pre-cluded ignition sources. On the basis of the above information, the licensee concluded that room ventilation is not necessary to prevent explosive condi-tions from occurring in the storage room.
The staff, in a letter dated May 14, 1984, concurred with the licensee's assess-ment. This issue is considered to be fully resolved.
2.12.2 Emergency Diesel Generator 1A Ventilation System 10 CFR 50 (GDC 4), as implemented by SRP Section 9.4.5, requires that systems and components important to safety be designed to accommodate the effect of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.
In the topic evaluation, dated October 28, 1982, the staff concluded that the ventilation system that services the 1A emergency diesel generator room is subject to disabling single failures. The higher operating ambient conditions that would result from loss of ventilation could adversely affect the perfor-mance of the enclosed diesel generator.
La Crosse SEP 2-9
l The licensee, in a letter dated December 29, 1983, indicated that the La Crosse procedures have been modified to require opening the machine shop door to the 1A emergency diesel generator room if the ventilation system becomes inoperable.
The staff, in a letter dated May 14, 1984, found that the procedure modifica-tion is acceptable to resolve this issue.
l l
l La Crosse SEP 2-10
E Table 2.1 Summary of IPSAR and supplement evaluations n
~
o SEP IPSAR Supplement Requirements 3 topic section section from o no. no. Title Requirements from IPSAR no. supplement II-1. A 4.1 Exclusion Area Authority Incorporate Tecnnical Specification (TS) 3.1 Complete - Amendment No. 41 and Control change to inform NRC of any changes in occupancy of privately owned land.
II-3.B 4.2 Flooding Potential and Put cutouts in parapets of the turbine, 4 Complete - See Inspection Protection Requirements office, and crib house buildings to Report 50-409/83-22.*
limit live load from ponded water to less then 30 psf.
II-3.B.1 4.3 Capability of Operating Develoo or modify emergency procedures 4 Complete - See Inspection Plants To Cope With for site flooding. Report 50-409/83-22."
Design-Basis Flooding Conditions. Propose limiting conditions for 3.2 Complete - Amendment No. 45 operation for site flooding.
II-3.C 4.4 Safety-Related Water Review possible loss of cooling water 3.3/4.2 Complete - Amendment No. 44 g Supply (Ultimate Heat Sink) to plant, and if needed develop TS and a procedures to identify alternate
[ sources of water.
III-1 4.5 Classification of Struc- -Analyze and upgrade, if necessary, 2.1 None tures, Components, and structures, components, and systems Systems (Seismic and as described in Section 4.5.
Quality)
III-2 4.6 Wind and Tornado Loadings Analyze and upgrade, if necessary, 2.2 To be determined the structural capacity for wind 1 and tornado loadings so that safe !
shutdown is ensured.
III-3.A 4.7 Effects of High Water Level on Structures 4.7.1 Containment Stability Calculate factors of safety against 2.3.1 None gross sliding and overturning of containment.
See footnotes at end of table.
l
. . =
~
l r.- Table 2.1 (Continued) ou n
g SEP IPSAR Supplement Requirements '!
e topic section section from
- no. no. Title Requirements from IPSAR no. supplement
" Review stack stability for design-m III-3.A 4.7.2 Stack Stability 2.3.2 None N basis flood level load combinations.
4.7.3 Crib House None -- --
III-3.C 4.8 Inservice Inspection of Perform inservice inspection as 4 Complete - See Inspection Water Control Structures described in Section 4.8. Report 50-409/83-22.*
III-4.A 4.9 Tornado Missiles Compare alterrate methods to achieve 2.2 To be determined safe shutdown to protect against tornado missiles.
III-4.8 4.10 Turbine Missiles Provide comparison of overspeed 2.4 Implement program of periodic protection testing for Genoa Units disassembly and inspection of 2 and 3. valves, valve exercising, and testing of overspeed trip.
7 w
III-5.A 4.11 Effects of Pipe Break on Structures, Systems, and Evaluate open Items 1, 2, 4, 5, and 6 in Section 4.11.
2.5 Reroute two branch line connections on high pressure N Components Inside core spray Containment Item 3 - relocate manual valve 4 56-24-009 and develop procedures to Complete - See Inspection close this valve in the event of Report 50-409/83-22.*
loss-of-coolant accident (see Section 4.21.2.3).
III-5.B 4.12 Pipe Break Outside Containment 4.12.1 Clarification of Pipe Whip Clarify pipe whip damage criteria 2.6 None Damage Criteria and Jet and jet impingement model.
Impingement Model 4.12.2 Verification of Potential None -- --
Releases From the Worst High-Energy-Line Break See footnotes at end of table.
[7 Table 2.1 (Continued) n g SEP IPSAR Supplement Requirements vi ' topic section section from
.Xl no. no. Title Requirements from IPSAR no. supplement
@@ III-5.B 4.12.3 Failure of Steam Heating None -- ' --
'o System in Electrical Equipment Room III-6 4.13 Seismic Design Analyze structures, systems, and com- 2. 2 To be determined Considerations ponents as described in Section 4.13.
III-7.B 4.14 Design Codes, Design Assess structural code changes on 2.7 To be determined Criteria Load Combinations, safety margins in "as-built" and Reactor Cavity Design structures as described in Criteria Section 4.14.
III-8.A 4.15 Loose-Parts Monitoring None -- --
and Core Barrel Vibration Monitoring III-10.A 4.16 Thermal-Overload Protection None -- --
h? for Motors of Motor-Operated Valves
{3 V-5 4.17 Reactor Coolant Pressure Boundary (RCP8) Leakage Detection 4.17.1 Leakage Sensitivity None -- --
4.17.2 Seismic Qualification Develop procedure identifying 4 Complete - See Inspection actions to be taken following Report 50-409/83-22.*
a seismic event and the failure of leakage detection equipment.
V-10.A 4.18 Residual Heat Removal See Section 4.20. -- --
System Heat Exchanger Tube Failures See footnotes at end of table.
g Table 2.1 (Continued)
) n Supplement Requirements y SEP IPSAR section from un topic - section
'- no. no. Requirements from IPSAR no. supplement
$ Title 7
V-10.B 4.19 Residual Heat Removal System Reliability 4.19.1 Use of Safety-Grade Systems None -- --
for Safety Shutdown 4.19.2 Shutdown Condenser Shell- Add a second level controller. 4.1 Complete (Open Item 83-22-14)
Side Level Control 4.19.3 Additional Emergency None -- --
Procedures V-12.A 4.20 Water Purity of BWR Primary Coolant 4.20.1 Chloride and pH Limits Revise chloride and pH limits to 3.4 Complete - Amendment No. 41 conform with RG 1.56.
E 4.20.2 Conductivity Limits Reestablish conductivity limits 3.4 Complete - Amendment No. 41 and sampling frequency following review of system capability.
VI-4 4.21 Containment Isolation System 4.21.1 Valve Location 4.21.1.1(1) Penetration M-8, High- Develop procedures to isolate 4 Complete - See Inspection Pressure Service Water penetration M-8 from outside Report 50-409/84-09.** '
Line containment coincident with failure of containment check valves.
( 4.21.1.1(2) Penetration M-11, None -- --
I Demineralized Water
! System Line See footnotes at end of table.
i
E Table 2.1 (Continued) n o SEP IPSAR Supplement Requirements
- topic section section from o no. no. litle Requirements from IPSAR no. supplement m
m VI-4 4.21.1.2 Penetrations M-21 and M-31, None -- --
" Vent Exhaust Damper and Ventilation Supply 4.21.1.3 Penetration M-23. Resin Develop procedures to close 4 Cosplete - See Inspection Sluice to Atmosphere new valve if loss-of-coolant Report 50-409/84-16.t accident occurs when resin transfer is in progress, normally locked closed at other times.
4.21.1.4 Penetration M-34, Evaluate present radiation 2.8.1 Complete Shutdown Condenser monitoring system and procedures Atmospheric Vent for detecting a radiation leak.
4.21.1.5 Penetration I-A, Alter- None -- --
nate Core Spray High-
. Pressure Service Water
'T Line 5 4.21.1.6 Penetration I-A, Con- None -- --
tainment Building Drain Suction Line 4.21.2 Valve Type 4.21.2.1 Penetrations M-9 and M-10, Develop procedures to close valves 4 Complete - See Inspection Component Cooling Water 57-24-012 and 57-24-034 ba uf an Report 50-409/84-09.**
Lines component cooling water surge tank level alare or other indication of line break.
4.21.2.2 Penetration M-12. Con- None -- --
trol Air System Line See footnotes at end of table.
I i
Table 2.1 (Continued)
E n Supplement Requirements 2 SEP IPSAR section from
.g topic section Requirements from IPSAR no. supplement p no. no. Title 4 Complete - See Inspection Penetration M-17, Decay Relocate manual valve 56-24-009 and VI-4 4.21.2.3 Heat Removal Line develop procedures to close this Report 50-409/84-09.**
valve in the event of loss-of-coolant accident.
4.21.2.4 Penetration M-18, Seal None Injection Line 4.21.2.5 Penetration M-28, Reactor None Cavity Purge Air Line Develop procedure to close remote 4 Complete - See Inspection 4.21.2.6 Penetration M-29, Offgas Report 50-409/84-09.**
vent to Chimney manual valve 55-25-004 outside containment if a high containment building activity automatic closure signal is sent to automatic valve 55-25-003 inside containment, or if a major primary break condition exists.
7 5 4.21.3 Valve Type and Locked-Closed valves 4 Complete - See Inspection 4.21.3.1 Penetration M-13, Station Lock close manual isolation valve Report 50-409/84-03.**
Air 70-24-30 outside containment and provide procedures to open and relock.
4.3 Install remotely operated valve 4.21.3.2 Penetration M-19, Off- Lock close manual valves 55-24-101 (0 pen Item 84-09-14).
and 62-28-013 and install a remotely 2.8.4 gas Vent From Shutdown Condenser operated solenoid valve outside containment.
2.8.3 Complete - See Inspection 4.21.3.3(1) Penetration M-22, Waste Valve 54-24-179 will be maintained Report 50-409/84-09.**
Water Line in a normally locked-closed position.
Its use will be controlled by administrative procedure.
See footnotes at end of table.
E Table 2.1 (Continued) n Supplement Requirements 2
m SEP topic IPSAR section section from Requirements from IPSAR no, supplement U no. . no. Title Complete - See Inspection S
VI-4 4.21.3.3(2) Penetration M-25, Waste Water Line Valve 54-24-162 will be maintained in a normally locked-closed position.
4 Report 50-409/84-09.**
Its use will be controlled by administrative procedure.
4.21.3.3(3) Penetration M-27, Waste Valve 54-24-160 will be maintained :.6.3 Complete - See Inspection Water Line in a normally locked-closed position. Report 50-409/84-09.**
4.21.3.4 Penetration M-26, Heat- Maintain valves 73-24-009 and 4 Complete - See Inspection ing System Supply and 73-24-057 in a normally locked- Report 50-409/84-09.**
Return closed position.
4.21.4 Instrument Lines None 4.21.5 Insufficient Indication Develop procedures to specify under 2.8.2 Complete for Operation of Remote which conditions remote manual 4 Manual va lves valves will be closed.
4.22 Containment Leak Testing Visually inspect airlock seals 72 hr 3.5 Complete - Amendment No. 41 and
[ VI-6 Inspection Report 50-409/83-22.*
after opening and replace seals in accordance with manufacturer's recommendations.
VI-7.A.3 4.23 Emergency Core Cooling Incorporate Test Procedures 17.5.1 3.6 T5 submitted System Actuation System and 17.5.2 into TS.
VI-7.C.1 4.24 Appendix K - Electrical Instrumentation and Con-trol Re-Reviews 4.24.1 480-V Essential Buses IA Develop interlocking method to prevent 2.9 None and IB both diesels from being paralleled.
4.24.2 120-V AC Circuit Breakers None See footnotes at end of table.
g Table 2.1 (Continued)
Q SEP IPSAR Supplement Requirements O topic section section from no. no. Requirements from IPSAR no. supplement g Title u VI-10.A 4.25 Testing of Reactor Trip Modify one power range channel 4 Complete - See Inspection System and Engineered to compensate for feedwater Report 50-409/83-22.*
o Safety Features, Including temperature deviation inaccuracies.
Response-Time Testing VII-1.A 4.26 Isolation of Reactor l Protection System from 1 Nonsafety Systems, j Including Qualification i of Isolation Devices 4.26.1 Channel Isolation Demonstrate by analysis and/or tests 2.10.1 Complete !
that a single failure in any one of the process recorders does not affect any other channel of the reactor protection system.
N 4.26.2 Qualification as None d Class IE Equipment cn 4.26.3 Isolation Between Reactor Demonstrate by analysis and/or test 2.10.2 Provide redundant power supplies Protection System (RPS) which RPS channels have inadequate for power-to-flow channel.
Channels and Power Supplies power supply isolation and correct at least one redundant channel.
4.26.4 Power Source for Scram Provide separate power supplies for 4 Complete Channels the full scran channels.
4.26.5 Isolation of Range. None Common Modules VIII-1.A 4.27 Potential Equipment Failures Associated with Degraded Grid Voltage 4.27.1 480-V Buses IA and IB None See footnotes at end of table.
E Table 2.1 (Continued)
E SEP IPSAR Supplement Requirements S topic section section from g no. no. Title Requirements from IPSAR no. supplement
$o VIII-1.A 4.27.2 Single Source of Offsite Mone -- --
Power VIII-3.8 4.28 DC Power System Bus Evaluate dc system monitoring and 2.11 Complete Voltage Monitoring and any necessary modifications.
Annunciation.
IX-5 4.29 Ventilation Systems 4.29.1.1 1A-480-V Essential Mone -- --
Switchgear 4.29.1.2 011 Vapors in Oil Storage Demonstrate that no ventilation is 2.12.1 None Room necessary for this room or develop procedures to provide ventilation.
4.29.2 Electrical Equipment Roon None -- --
to O
o 4.29.3 Emergency Diesel Generator Demonstrate that the 1A emergency 2.12.2 Complete 1A Ventilation System diesel generator can continue to function without ventilation or propose corrective seasures.
4.29.4 Diesel Building Develop procedures for monitoring 4 Complete - See Inspection Ventilation System ventilation in diesel building Report 50-409/83-22.*
rooms.
4.29.5 Intake Structure None -- --
IX-6 4.30 Fire Protection -- -- --
XV-20 4.31 Radiological Consequences None -- --
of Fuel-Damaging Accidents
" Letter dated January 25, 1984.
- Letter dated October 5,1984.
tLetter dated January 28, 1985.
3 TOPICS RESOLVED BY CHANGES TO PLANT TECHNICAL SPECIFICATIONS During the integrated assessment for La Crosse, a number of issues were resolved by commitments from the licensee to perform evaluations in order to determine whether modifications to plant Technical Specifications are warranted.
This section describes the actions taken regarding resolution of IPSAR issues involving Technical Specification changes. A summary of all changes to Tech-nical Specifications and the associated license amendment is given in Table 3.1.
3.1 Topic II-1.A, Exclusion Area Authority and Control (NUREG-0827, Section 4.1) 10 CFR 100.3(a), as implemented by SRP Section 2.1.7, requires the licensee to have the authority to determine all activities, including exclusion or removal of personnel and property, in the exclusion area.
In IPSAR Section 4.1, the staff noted that two parcels of land under private ownership with restrictive easements were within the exclusion area. The licensee proposed to add a Technical Specification (TS) requirement to notify the staff of any changes in occupancy of this land. The TS change was sub-mitted December 19, 1983, and was issued in Amendment 41 to the license on May 28, 1985.
3.2 Topic II-3.B.1, Capability of Operating Plants To Cope With Design-Basis Flooding Conditions (NUREG-0827, Section 4.3) 10 CFR 50 (GDC 2), as implemented by SRP Section 2.4.10, requires that struc-tures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as flooding.
Section 4.3 of the IPSAR concluded that emergency procedures and limiting con-ditions for operation for site flooding should be developed. The licensee has completed work on emergency procedures for site flooding as noted in Inspection Report 83-22. By letter dated December 19, 1983, as revised June 25, 1985, the licensee submitted a proposed Technical Specification requiring that whenever the river level adjacent to the site reaches 639.2 ft and is predicted to reach 640 ft, the plant will be shut down. Supporting information was pro-vided in a February 22, 1985, letter from the licensee. This TS change was approved by Amendment 45 to the license, dated January 6, 1986.
3.3 Topic II-3.C, Safety-Related Water Supply (Ultimate Heat Sink)
(NUREG-0827, Section 4.4) 10 CFR 50 (GDC 2), as implemented by SRP Section 2.4.11 and RG 1.27, requires that structures, systems, and components important to safety be designed to withstand the effects of natural phenomena. IPSAR Section 4.4 concluded that the ultimate heat sink function could be disrupted because of low water level La Crosse SEP 3-1
I on the Mississippi River should dams downstream of the site fail. The licensee has identifitd that the shutdown condenser with sufficient makeup sources could provide long-term decay heat removal. Therefore, the licensee proposed a TS change to require that a minimum volume'fre raintained in the virgin water tank to provide immediate condenser makeup until other sources, such as well l
water or the emergency service water supply system', can be aligned. Operating i procedures for use of these systems would also be provided. . The licensee's approach is discussed in letters dated March 8, 1984, March 21, 1984, and February 22, 1985.
The proposed TS change was approved by Amendment 44 to the license, dated October 8, 1985.
3.4 Topic V-12.A, Water Purity of BWR Primary Coolant (NUREG-0827, Section 4.20) 10 CFR 50 (CDC 10, as implemented by RG 1.56, requires that the reactor cool-ant pressure boundny (RCPB) have minimal probability of rapidly propagating failure. Thi,s includes corrosion-induced failures from impurities in the reac-tor coolant system. The safety objective of this review is to ensure that the plant reactor coolant chemistry is adequately controlled to minimize the possi-bility of corrosion-induced failures.
In IPSAR Section 4.20.1, the licensee proposed to submit TS changes to add limits on pH and to reduce the limit on chloride concentration to the recom-mended values.in FG 1.56. In Section 4.20.2, the licensee agree to reestablish conductivity ifmits, based on cleanup system capacity, and to establish a sam-pling frequency in the Technical Specifications.
These proposed changes were submitted by letter dated December 19, 1983, and werrr approved by the staff as Amendment 41 to the license, dated May 28, 1985.
3.5 Topic VI-6, Containment Leak Testing (NUREG-0327, Section 4.22) 10 CFR 50, Appendix J, requires that test" be performed to ensure that leakage through the primary reactor containment snd systemsiand components penetrating primary containment shall not exceed allowable leakage rate values as specified in the Technical Specifications or associated bases.
In Section 4.22 of the IPSAR, the need for TS chang,es on inspection of airlock door seals was identified. By letter dated September 27, 1984, the staff is-sued an exemption to Section III.D.2.b.iii of Appendix J to 10 CFR 50. This section requires airlock leakage tests within 3 days of each opening. The ex-emption was approved provided that (1) leakage tests of containment airlocks ~
are performed every 4 months (current TS requirement) (2) door seals on con-tainment airlocks are visually inspected for degradation after each opening but not required more often than once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and (3) door seals on con-tainment airlocks are replaced periodically in accordance with manufacturer's recommendations. By letter dated December 19, 1983, the licensee submitted pro-posed TS changes to implement (2) and (3) above. These changes were approved in Amendment 41 to the license, dated May 28, 1985.
La Crosse SEP 3-2
3.6 Topic VI-7.A.3, Emergency Core Cooling System Actuation System (NUREG-0827, Section 4.23) 10 CFR 50.55a(h), as implemented by IEEE Std. 279-1971, and 10 CFR 50 (GDC 37),
as implemented by RG 1.22, require that equipment important to safety be tested periodically to ensure the operability of the system as a performance of the i full operational sequence that brings the system into operation, including operation of the associated cooling water system.
The emergency core cooling system actuation system is tested in accordance with Test Procedures 17.5.1 and 17.5.2. These procedures fulfill the requirements for emergency core cooling system actuation tests but are not currently re-quired in the Technical Specifications. The licensee submitted, in a letter dated September 29, 1982, a proposed revision to the LACBWR Technical Specifica-tions which incorporates Test Procedures 17.5.1 and 17.5.2 into the Technical Specifications surveillance requirements. The proposed revision is under staff review.
Table 3.1 Modifications to La Crosse Technical Specifications as a result of SEP SEP IPSAR Supplement topic section section Amendment no. no. Title no. no.
II-1.A 4.1 Exclusion Area Authority 3.1 41 and Control II-3.8.1 4.3 Capability of Operating Plants To 3.2 45 Cope With Design-Basis Flooding Conditions II-3.C 4.4 Safety-Related Water Supply 3.3 44 (Ultimate Heat Sink)
V-12.A 4.20.1 Chloride and pH Limits 3.4 41 4.20.2 Conductivity Limits 3.4 41 VI-6 4.22 Containment Leak Testing 3.5 41 VI-7.A.3 4.23 Emergency Core Cooling 3.6 Pending System Actuation System
- La Crosse SEP 3-3
4 IPSAR TOPIC RESOLUTIONS CONFIRMED BY THE NRC REGIONAL OFFICE During the integrated assessment for La Crosse, a number of issues were resolved by commitments made by the licensee for specific plant modifications or pro-cedural changes.
Subsequent to issuance of the IPSAR far La Crosse, the NRC regional office was requested through Task Interface Agreement 83-79 to verify that plant modifica-tions had been implemented and to review changes to plant operating procedures made by the licensee.
Table 4.1 provides a list of IPSAR actions for which regional confirmation was requested.
Region 3 staff conducted onsite inspections for each item identified in Ta-ble 4.1. The inspections consisted of examinations of installed equipment as well as a review of supporting procedures and other documentation. Region 3 staff concluded that the licensee had met the commitments documented in the IPSAR for the items in Table 4.1 except for the open items discussed below.
Inspection findings with the results of this review are documented in Inspec-tion Reports 50-409/83-22, 50-409/84-09, and 50-409/84-16 (see letters dated January 25, 1984, October 5, 1984, and January 28, 1985, respectively).
4.1 Shutdown Condenser Shell-Side Level Control (NUREG-0827, Section 4.19.2)
In IPSAR Section 4.19.2, the staff recommended that a second level controller for the shutdown condenser shell-side water be installed. The controller was installed during the 1986 refueling outage.
4.2 Offaas Vent From Shutdown Condenser (NUREG-0827, Section 4.21.3.2)
The licensee committed to lock close two manual valves (55-24-101 and 62-?8-013) and to install a remotely operated solenoid valve outside contain-ment for this penetration. The procedural changes for the locking of the valves have been completed. By letter dated April 8, 1985, the licensee noted that installation of an additional valve on this line would require upgrading of the entire penetration line. Therefore, the licensee requested a deferral of this modification as discussed in Section 2.8.4 of this supplement. This item is Open Item 409/84-09-14.
La Crosse SEP 4-1 l
Table 4.1 Items for confirmation by NRC regional office Item IPSAR no. Description section no. Status
-1 Put cutouts in parapets of th'e 4.2 Complete turbine, office, and crib house buildings to limit live load from pounded water to less than 30 psf.
2 Develop or modify emergency 4.3 Complete; see also procedures for site flooding. Section 3.2 3 Review possible loss of cooling 4.4 Complete; see also water to plant, and if needed Section 3.3 develop procedures to identify alternate sources of water.
4 Develop procedures for performing 4.8 Complete inservice inspection of water control structures.
5 Develop procedures identifying 4.17.2 Complete actions to be taken following a seismic event and the failure of leakage detection equipment.
6 Add a second level controller 4.19.2 See Section 4.1 for shutdown condenser shell-side level control.
7 Containment isolation systems 4.21 See Sections 4.2 and 2.8.4 (a) Make several hardware modifications.
(b) Develop procedures to ensure valve closures.
8 Containment airlock seals should 4.22 Complete ,
be visually inspected and seals replaced in accordance with '
manufacturer's recommendations.
9 Modify one power range channel 4.25 Complete to compensate for feedwater temperature deviation inaccuracies.
10 Provide separate power supplies 4.26.4 Complete for the full scram channels.
11 Develop procedures for monitoring 4.29.4 Complete ventilation in diesel building rooms.
La Crosse SEP 4-2
5 REFERENCES Code of Federal Regulations, Title 10, " Energy," U.S. Government Printing Office, Washington, D.C. (includes General Design Criteria).
Dairyland Power Cooperative, " Final Hazards Summary Report," October 1962.
Letter, September 29, 1982, from F. Linder (DPC) to D. Crutchfield (NRC), Sub-ject: Application for License Amendment.
-- , October 28, 1982, from D. Crutchfield (NRC) to F. Linder (DPC),
Subject:
SEP Topic IX-5, Ventilation Systems.
-- , November 8,1982, from D. Crutchfield (NRC) to F. Linder (DPC),
Subject:
SEP Topic III-5.A, Effects of Pipe Break on Systems, Structures and Components Inside Containment.
-- , December 9, 1982, from D. Crutchfield (NRC) to F. Linder (DPC),
Subject:
SEP Topic VI-4, Containment Isolation Systems.
-- , February 14, 1983, from F. Linder (DPC) to D. Eisenhut (NRC),
Subject:
SEP Topic III-3.A (LAC-8861).
-- , February 16, 1983, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
Response to Summary of Differences (LAC-8881).
-- , August 31, 1983, from F. Linder (DPC) to D. Crutchfielo gdRC),
Subject:
SEP Topics VI-7.C.1 and VIII-3.B (LAC-9302).
-- , September 29, 1983, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
Summary of Proposed Integrated Assessmen't Actions (LAC-9324).
-- , October 6,1983, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
SEP Topic III-5.B, Pipe Break Outside Containment (LAC-9351).
-- , October 7,1983, from D. Crutchfield (NRC) to F. Linder (DPC),
Subject:
SEP Topic VI-7.C.1.
-- , October 27, 1983, from D. Crutchfield (NRC) to F. Linder (DPC),
Subject:
IPSAR Section 4.12,_ Pipe Break Outside Containment.
-- , October 27,1983a, from D. Crutchfield (NRC) to F. Linder (DPC),
Subject:
Integrated Plant Safety Assessment Report (IPSAR), Sections 4.26 and 4.26.3 -
Isolation of Reactor Protection System..
-- , December 19, 1983, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
Application for License Amendment (LAC-9480).
La Crosse SEP 5-1
-- , December 19,1983a, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
SEP Topic VIII-3.B (LAC-9490).
-- , December 29, 1983, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
Integrated Plant Safety Assessment (LAC-9514).
-- , January 25, 1984, from W. Shafer (NRC) to F. Linder (DPC),
Subject:
Inspection Report 50-409/83-22.
-- , February 9,1984, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
SEP Topic IX-5, Ventilation System (LAC-9578).
-- , March 8,1984, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
IPSAR Section 4.4 (LAC-9627).
-- , March 21, 1984, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
IPSAR Section 4.4, License Amendment Application (LAC-9725).
-- , March 28, 1984, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
IPSAR Section 4.21.3.3, Waste Water Lines (LAC-9742).
-- , April 3,.1984, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
Shutdown Condenser Tube Leakage (LAC-9797).
-- , April 4, 1984, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
SEP Topic III-4.B, Turbine Missiles (LAC-9686).
-- , April 26, 1984, from D. Crutchfield (NRC) to F. Linder (DPC),
Subject:
IPSAR Section 4.21.1.4, Shutdown Condenser Atmospheric Vent, Penetration M-34.
-- , May 14, 1984, from D. Crutchfield (NRC) to F. Linder (DPC),
Subject:
IPSAR Section 4.29, Oil Vapors in Oil Storage Room.
-- , June 7, 1984, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
SEP Topic III-4.B.
-- , June 29, 1984, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
IPSAR Section 4.11 (LAC-10011).
-- , July 3, 1984, from F. Linder (DPC) to D. Crutchfield (NRC),
Subject:
IPSAR Section 4.7.
-- , September 27, 1984, from W. Paulson (NRC) to F. Linder (DPC),
Subject:
Exemption From Appendix J Airlock Leak Testing Requirements.
-- , October 5,1984, from W. Shafer (NRC) to F. Linder (DPC),
Subject:
Inspection Report 50-409/84-09.
-- , October 11, 1984, from F. Linder (DPC) to D. C'utchfield r (NRC),
Subject:
Integrated Consequence Study.
-- , December 20, 1984, from F. Linder (DPC) to W. Paulson (NRC),
Subject:
IPSAR Section 4.14.
l l
La Crosse SEP 5-2
-- , January 3,1985, from F. Linder (DPC) to W. Paulson (NRC),
Subject:
IPSAR Section 4.26.3 Schedule.
-- , January 22, 1985, from F. Linder (DPC) to J. Zwolinski (NRC),
Subject:
IPSAR Section 4.28.
-- , January 28, 1985, from R. Warnick (NRC) to F. Linder (DPC),
Subject:
Inspection Report 50-409/84-16.
-- , February 8,1985, from F. Linder (DPC) to J. Zwolinski (NRC),
Subject:
Integrated Assessment (4.21.5).
-- , February 22, 1985, from F. Linder (DPC) to J. Zwolinski (NRC),
Subject:
SEP Topics II-3.B.1 and II-3.C (LAC-10594).
-- , March 15, 1985,, from F. Linder (DPC) to J. Zwolinski (NRC),
Subject:
Dairyland Power Cooparative, (DPC) La Crosse Boiling Water Reactor (LACBWR)
Provisional Operating License No. DPR Integrated Consequence Study, NUREG-0827, Integrated Plant Safety Asnessment.
-- , April 1,1985, from F. Linder (DPC) to J. Zwolinski (NRC),
Subject:
SEP Topic III-2, Wind and Tornado Loadings (LAC-10673).
-- , April 8, 1985, from F. Linder (DPC) to J. Zwolinski (NRC),
Subject:
IPSAR Change of Commitment Date for Section 4.21.3.2 (LAC-10683).
-- , April 19, 1985, from J. Zwolinski (NRC) to F. Linder (DPC),
Subject:
IPSAR Section 4.10, Turbine Missiles.
-- , April 24, 1985, from J. Zwolinski (NRC) to F. Linder (DPC),
Subject:
NUREG-0827, Section 4.28, DC Power System Bus Voltage Monitoring and Annunci-ation - LACBWR.
-- , May 16, 1985, from J. Zwolinski (NRC) to F. Linder (DPC),
Subject:
Com-mi+xent Date for Section 4.21.3.2.
-- , May 28, 1985, from J. Zwolinski (NRC) to F. Linder (DPC), transmitting Amendment No. 41 to License DPR-45.
-- , June 25, 1985, from F. Linder (DPC) to J. Zwolinski (NRC), transmitting revised amendment application on site flooding.
-- , July 26, 1985, from F. Linder (DPC) to J. Zwolinski (NRC),
Subject:
SEP Topic III-5.A (LAC-11053).
-- , August 13, 1985, from J. Zwolinski (NRC) to F. Linder (DPC), subject:
IPSAR Section 4.11, Effects of Pipe Breaks in Systems, Structures, and Com-ponents Inside Containment.
-- , August 29, 1985, from J. Taylor (DPC) to J. Zwolinski (NRC),
Subject:
IPSAR Section 4.21.3.2 (LAC-11095).
La Crosse SEP 5-3
-- , September 6,1985, from J. Zwolinski (NRC) to F. Linder (DPC),
Subject:
IPSAR Sections 4.7.1 Containment Stability and 4.7.2 Stack Stability.
-- , October 8, 1985, from J. Zwolinski (NRC) to J. Taylor (DPC), transmitting Amendment No. 44 to License DPR-45.
t
--4, November 14, 1985, from J. Taylor (DPC) to J. Zwolinski (NRC),
Subject:
IPSAR 4.26.3 (LAC-11234).
-- , December 23, 1985, from J. Zwolinski (NRC) to J. Taylor (DPC),
Subject:
IPSAR Section 4.26.3.
-- , January 6,1986, from J. Zwolinski (NRC) to J. Taylor (DPC),
Subject:
Amendment 45 to License DPR-45.
-- , March 10, 1986, from J. Zwolinski (NRC) to J. Taylor (DPC),
Subject:
Shutdown Condenser Sample Line.
U.S. Nuclear Regulatory Commission, NUREG-0800 (formerly NUREG-75/087), " Stand-ard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," June 1981 (includes Branch Technical Positions).
-- , NUREG-0827, " Integrated Plant Safety Assessment, Systematic Evaluation Program, La Crosse Boiling Water Reactor, Final Report," June 1983.
-- , Regulatory Guide (RG) 1.6, " Independence Between Redundant Standby (Onsite)
Power Sources and Between Their Distribution Systems."
-- , RG 1.22, " Periodic Testing of Protection System Actuation Functions."
-- , RG 1.26, " Quality Group Classifications and Standards for Water , Steam ,
and Radioactive-Waste-Containing Components of Nuclear Power Plants."
-- , RG 1.27, " Ultimate Heat Sink for Nuclear Power Plants."
-- , RG 1.32, " Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants."
-- , RG 1,47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems."
-- , RG 1.56, " Maintenance of Water Purity in Goiling Water Reactors."
-- , RG 1.75, " Physical Independence of Electric Systems."
-- , RG 1.102, " Flood Protection for Nuclear Power Plants."
-- , RG 1.115, " Protection Against Low-Trajectory Turbine Missiles."
-- , RG 1.118, " Periodic Testing of Electric Power and Protection Systems."
-- , RG 1.129, " Maintenance, Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants."
-- , RG 1.141, " Containment Isolation Provisions for Fluid Systems."
La Crosse SEP 5-4
Industry Codes and Standards American Society of Mechanical Engineers, " Boiler and Pressure Vessel Code" (ASME Code),Section III, " Nuclear Power Plant Components," Article NE-3331(b),
1980.
Institute of Electrical and Electronics Engineers (IEEE) Std. 279-1971, " Criteria for Protection System for Nuclear Power Generating Stations."
La Crosse SEP 5-5
\
APPENDIX A REFERENCES TO CORRESPONDENCE FOR EACH TOPIC EVALUATED La Crosse SEP
IPSAR supplement section no. Date Reference 2.3 9/6/85 Letter from J. A. Zwolinski (NRC) to J. W. Taylor (DPC),
Subject:
Integrated Plant Safety Assessment Report Sections 4.7.1, Containment Stability and 4.7.2, Stack Stability 2.4 4/19/85 Letter from J. A. Zwolinski (11RC) to F. Linder (DPC),
Subject:
IPSAR Section 4.10, Turbine Missiles 2.5 8/13/85 Letter from J. A. Zwolinski (NRC) to F. Linder (DPC),
Subject:
IPSAR Section 4.11, Effects of Pipe Break in Systems, Structures and Compo-nents Inside Containment 2.6 10/27/83 Letter from D. Crutchfield (NRC) to F. Linder (DPC),
Subject:
IPSAR Section 4.12, Pipe Break Outside Containment 2.8.1 4/26/84 Letter from D. Crutchfield (NRC) to F. Linder (DPC),
Subject:
IPSAR Section 4.21.1.4, Shutdown Condenser Atmospheric Vent, Penetration M-34.
2.8.2 3/10/86 Letter from J. Zwolinski (NRC) to J. Taylor (DPC),
Subject:
Shutdown Condenser Sample Line 2.9 10/7/83 Letter from D. M. Crutchfield (NRC) to F. Linder (DPC),
Subject:
Integrated Plant Safety Assessment Report (IPSAR) 4.24.1, Appendix K, Electrical Instrumentation and Control Reviews for the La Crosse Boiling Water Reactor 2.10 10/27/83 Letter from D. M. Crutchfield (NRC) to F. Linder (DPC),
Subject:
Integrated Plant Safety Assessment Report (IPSAR), Sections 4.26 and 4.26.3 - Isolation of Reactor Protection System La Crosse SEP A-1
.IPSAR supplement Section no. Date Reference 2.11 4/24/85 Letter from J. A. Zwolinski (NRC) to F. Linder (DPC),
Subject:
NUREG-0827, Section 4.28, DC Power System Bus Voltage Monitoring and Annunciation - LACBWR 2.12 5/14/84 Letter from D. M. Crutchfield (NRC) to F. Linder (DPC),
Subject:
IPSAR Section 4.29, Oil Vapors in Oil Storage Room La Crosse SEP A-2
APPENDIX B NRC STAFF CONTRIBUTORS f
La Crosse SEP
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o This supplement to the IPSAR is a product of the NRC staff. The NRC staff mem-bers listed below were principal contributors to this report.
NRC Staff Title Branch / Division J. Chen Geotechnical Engineer RIB /DSR0 D. Persinko Maintenance and Surveillance Engineer HFMT/DHFT P. Y. Chen Senior Mechanical Engineer ISAPD/DPL-B L. Bell Nuclear Engineer RSB/DPL-A R. Scholl Senior Reactor Systems Engineer ORAS S. Brown Reactor Systems Engineer PAF0 E. McKenna Senior Project Manager PD#1/DPL-A J. Stang Project Manager BD#1/ DBL C. Grimes Project Director ISAPD/DPL-B La Crosse SEP B-1
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I BIBLIOGRAPHIC DATA SHEET Supplement No.1 g
- 4. TITLE AND SUBTITLE (Add Vo/ume No., #f apprqanare) 2, (Leave star,4) @
[ Integrated Plant Safety Assessment / h e
Systematic Evaluation Program 3. RECIPle[FS ACCESSION NO.
[ La Croke Boiling Water Reactor / 3-7 AUTHOR (S) -
- 5. DAT[ REPORT COMPLE TED M[TH l YEAR i
- 9. PE RFORMING ORG 12ATION N AME AND MAILING ADDRESS (Include 2,0 Codel fJgust A'E REPORT ISSUED 1986 M i
Division of B Licensing
- Office of Nucle Reactor Regulation . F Ifigu"st I^f986 A r U.S. Nuclear Reg atory Commission 6 't**
- o'**>
m i Washington, D.C.
555 8 (Lene Nwk) T
- 12. SPONSORING ORGANIZATION AME AND M AILING ADDRE SS (include l'a Codel __
Same as #9 above.
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11 FIN NO. 1_
l 13 TYPE OF REPORT
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[PE RIOD COVE RE D (inclus,ve darvsl s
- Technical Evaluation -
-m k 'Ye"rfYi"n's"'t'o"*D"ocIIe*t No. 50-409
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- 16. ABSTR ACT (200 words or lessJ -
R The Nuclear Regulatory Commission (N as published its Supplement No.1 to the :
- Integrated Plant Safety Assessment Rep t (IPSAR) (NUREG-0827) under the scope of the F Systematic Evaluation Program (SEP),
! Dairyland Power Cooperative La Crosse Boiling
- review the design of older operatio , nucle power plants to reconfirm and document '
their safety. This report documen ~ the re 'ew completed under the SEP for those issues de-that required refined engineering ' valuation or the continuation of ongoing evaluations after the Final IPSAR for the LA WR was issue The review has provided for (1) an -
assessment of the significance differences b ween current technical positions on 1 l selected safety issues and tho that existed wh LACBWR was licensed, (2) a basis for .:-
I deciding on how these differe 'es should be resolv in an integrated plant review, =-
[ and (3) a documented evaluati n of plant safety whe the supplement to the Final IPSAR _
=
and the Safety Evaluation Re ort for converting the ' cense from a provisional to a B-l full-term license have been issued. The IPSAR and it supplements will form part of m E the bases for considering e conversion of the license -
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- 17. KEY WORDS AND DOCUMENT AN A YSIS 17a DE SCRIPTORS . ;__
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