ML18051A679
| ML18051A679 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 11/30/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| TASK-***, TASK-RR NUREG-0820, NUREG-0820-S01, NUREG-820, NUREG-820-S1, NUDOCS 8312070296 | |
| Download: ML18051A679 (36) | |
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NUREG-0820 Supplement No. 1 Integrated Plant Safety Assessment Systematic Evaluation Program Palisades Plant Consumers Power Company Docket No. 50-255 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation November 1983
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NUREG-0820 Supplement No. 1 Integrated Plant Safety Assessment Systematic Evaluation Program Palisades Plant Consumers Power Company Docket No. 50-255 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation November 1983
ABSTRACT The Nuclear Regulatory Commission (NRC) has published Supplement No. 1 to the Final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0820), under the scope of the Systematic Evaluation Program (SEP), for Consumers Power Company 1 s Palisades Plant located in Covert, Van Buren County, Michigan.
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 a.fter the Final IPSAR for the Palisades 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 Palisades 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.
Palisades SEP iii
CONTENTS c,
\\_
Page ABSTRACT....... *................,.......................................
iii ACRONYMS AND INITIALISMS...............................................
vii 1
INTRODUCTION............................................... ;....... ~
1-1 2 TOPICS THAT REQUIRED REFINED ENGINEERING ANALYSIS OR CONTINUATION OF ONGOING EVALUATION...............................................
2-1 2.1 Topics II-3.B, Flooding Potential and Protection Requirements; II-3.B.l, Capability of Operating Plants To Cope With Design-Basis Flooding Conditions; and II-3.C, Safety-Related Water Supply (Ultimate Heat Sink)......... ~.....
2-1 2.2 Topic III-1, Classification of Structures, Components, and Systems...................................,....
2-2 2.3 Topic III-5.A, Effects of Pipe Break on Structures, Systems, and Components Inside Containment.............. ~......
2-2 2.4 Topic III-6, Seismic Design Considerations.....................
2-3 2.5 Topic III-7.B, Design Codes, Design Criteria, Load Combinations, and Reactor Cavity Design Criteria..............
2-4 2.6 Topic VIII-4, Electrical Penetrations of Reactor Containment...................................................
2-5 2.7 Topic IX-3, Station Service and Cooling Water Systems.........
- 2-6 2.7.1 Cooling of CCW Heat Exchanger... ~.......................
2-6
- 2. 7. 2 Flooding of Safety Systems in Intake Structure..........
2-6 2.8 Topic IX-5, Ventilation Systems................................
2-7 2.8.1 Ventilation of Auxiliary Feedwater Pump Room...........
2-7 2.8.2 Ventilation of Cable-Spreading, Switchgear, and Battery Rooms...................................... ;
2-7 3 STATUS
SUMMARY
OF IMPLEMENTATION ACTIONS............................
3-1 3.1 Topic XV-2, Spectrum of Steam System Piping Failures Inside and Outside Containment (PWR)...............................,..
3-1 4
REFERENCES..........................................................
4-1 APPENDIX A REFERENCES TO CORRESPONDENCE FOR EACH TOPIC EVALUATED APPENDIX B -- NRC STAFF CONTRIBUTORS AND CONSULTANTS Palisades SEP v
CONTENTS (continued)
LIST OF TABLES 2.1 Summary of Supplement Evaluations...............................
2-9 3.1 Integrated Assessment Summary...................................
3-3 Palisades SEP vi
ac ccw CFR CPCo de FSAR FTOL GDC gpm IEEE IPSAR ISI LOCA LPSI MCC MSIV msl NRC POL RG SEP SER SRP Std UHS ACRONYMS AND INITIALISMS alternating current component cooling water Code of Federal Regulations Consumers Power Company direct current Final Safety Analysis Report full-term operating license General Design Criterion(a) gallons per minute Institute of Electrical and Electronics Engineers Integrated Plant Safety Assessment Report inservice inspection loss-of-coolant accident low-pressure safety injection motor control center main steam isolation valve mean sea level U.S. Nuclear Regulatory Commission provisional operating license Regulatory Guide Systematic Evaluation Program Safety Evaluation Report Standard Review Plan Standard ultimate heat sink Palisades SEP vii
1 INTRODUCTION INTEGRATED PLANT SAFETY ASSESSMENT REPORT SUPPLEMENT NO. 1 SYSTEMATIC EVALUATION PROGRAM PALISADES PLANT The Systematic Evaluation Program (SEP) was initiated by the U.S. Nuclear 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 technical 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 in an integrated plant review, and (3) a documented evaluation of plant safety.
The initial review of the Palisades Plant as part of the SEP was published in NUREG-0820, the Integrated Plant Safety Assessment Report (IPSAR), dated October 1982.
The review compared the as-built plant design with current review criteria in 137 different areas defined as topics.
During the review, 47 of the topics were deleted from consideration in the SEP because a review was being conducted under other programs (Unresolved Safety Issues or Three Mile Island Action Plan tasks), the topic was not applicable to the Palisades Plant, or the items to be reviewed under that topic did not exist at the site.
Of the original 137 topics, 90 were, therefore, reviewed for Palisades; of these, 59 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.
In general, the staff's positions 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
- engineering analysis or continuation of ongoing evaluation, and (4) no modifi-cation-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 IPSAR lists the scheduled completion dates agreed upon by the staff and the licensee.
The NRC Region III will provide verification of the implementation of these positions.
Palisades SEP 1-1
For those positions classified as Category (3), the licensee has provided the results of the ongoing evaluation to the staff for review.
The purpose of this supplement to the IPSAR is to provide the staff 1 s evaluation of the Category (3) issues and to summarize the status of all actions to be imple-mented as a result of the IPSAR and this supplement to the IPSAR.
The Palisades Plant is one of the seven SEP plants that has not been issued its full-term operating license (FTOL).
A Safety Evaluation Report (SER) to support the conversion of the provisional operating license (POL) to an FTOL will be prepared.
The SER will consist of the IPSAR, the IPSAR Supplement, a consideration of major plant modifications that have been made and the substan-tive 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.)
Palisades SEP 1-2
2 TOPICS THAT REQUIRED REFINED ENGINEERING ANALYSIS OR CONTINUATION OF ONGOING EVALUATION Table 2.1 of this report presents the list of issues that were classified as Category (3) in the IPSAR and provides a summary of the corrective actions that were required.
The licensee has submitted an evaluation for each of these topics, which the staff has reviewed.
The staff concluded that either the licensee evaluation met current criteria, the evaluation was acceptable on another defined basis, a corrective action will be required, or further analysis will be required.
Factors considered 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 and their consultants.
2.1 Topics II-3.B, Flooding Potential and Protection Requirements; II-3.B.1, Capability of Operating Plants To Cope With Design-Basis Flooding Conditions; and II-3.C, Safety-Related Water Supply (Ultimate Heat Sink)
(NUREG-0820, Sections 4.2, 4.3, and 4.4)
Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50) (General Design Criterion (GDC) 2), as implemented by Standard Review Plan (SRP)
(NUREG-0800) Section 2.4.5 and Regulatory Guide (RG) 1.59, requires that structures, systems, and components important to safety be designed to with-stand the effects of natural phenomena such as floods.
The topic evaluation issued on February 19, 1982, found that the postulated flood level resulting from a design-basis seiche (a wind/pressure-driven wave) would be 597.1 ft mean sea level (msl).
At this level of flooding, the plant facilities and procedures would not be able to cope with the water.' Specifi-cally, the service water and auxiliary feedwater pumps are subject to flooding at a water level of 594.7 ft msl, a level which is 2.4 ft below the postulated design-basis level.
The licensee, in a letter dated March 23, 1982, submitted an evaluation of flooding caused by a seiche.
The staff reviewed this report and also reanalyzed the design-basis seiche flood level.
The staff, in an evaluation dated October 7, 1982, concluded that, on the basis of current licensing acceptance criteria, the postulated flood level should be 593.5 ft msl.
This level of flooding is not a threat to the service water or auxiliary feedwater pumps.
Therefore, the staff concludes that the emergency procedures for this Palisades SEP 2-1
flood level currently in effect at the Palisades Plant are adequate and that no modification of either the procedures or the plant is necessary.
This issue is considered to be fully resolved.
2.2 Topic III-1, Classification of Structures, Components, and Systems (NUREG-0820, Section 4.5) 10 CFR 50 (GDC 1), as implemented by RG 1.26, requires that structures, systems, and components important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of safety functions to be performed.
The staff, in the topic evaluation dated December 28, 1981 and in Section 4.5 of the IPSAR, concluded that insufficient information existed at that time to complete the topic review and requested that the licensee supply additional information and analyses in the following topic review areas.
(1) radiography (2) fracture toughness (3) valves (4) pumps (5) storage tanks The licensee, in a letter dated August 12, 1982, submitted the information and analyses requested by the staff.
The staff, in a subsequent evaluation dated June 27, 1983, concluded that the information supplied by the licensee was adequate to resolve all of the open issues in the five topic review areas, except for the fracture toughness testing of nine components.
The nine components (pressurizer, piping of reactor coolant system hot and cold legs, safety injection tanks, component cooling ~nd service water pumps, main steam piping, atmospheric dump. valves, and feedwater piping) had not been tested and were not exempt from testing.
For these nine components the licensee provided fracture toughness analyses to demonstrate quaiitatively that an adequate margin exists between operating temperature and material transition temperature so that sudden brittle fracture of the components is very unlikely.
The staff found these analyses acceptable; therefore, this issue is considered to be fully resolved.
2.3 Topic III-5.A, Effects of Pipe Break on Structures, Systems, and Components Inside Containment (NUREG-0820, Section 4.9) 10 CFR 50 (GDC 4), as implemented by RG 1.46 and SRP Section 3.6.2, requires, in part, that structures, systems, and components important to safety be appropriately protected against dynamic effects, such as pipe whip and discharging fluids that may result from equipment failures.
The effect of pipe breaks inside containment was not a part of the original design basis of the Palisades Plant.
The topic evaluation, dated December 4, 1981, identified over 200 pipe break locations for which additional information was required to determine if the postulated pipe break could adversely affect safety-related systems or compo-nents. to that evaluation provided guidance.to the licensee on how to evaluate the postulated pipe breaks and their effects.
In that Palisades SEP 2-2
evaluation, the staff also requested information on target p1p1ng acceptability criteria, that is, the criteria the licensee used to judge if a pipe would be able to function after it had been struck by a broken pipe.
The licensee submitted analyses by letters dated August 16, 1982 and December 9, 1982.
The staff reviewed the analyses and issued a safety evaluation report on June 22, 1983.
The conclusions of that review were:
(1) The target piping acceptability criteria proposed by the licensee are acceptable.
(2) _The pipe break locations for which there was no resolution were reduced to 14 from the original number of break locations using methods acceptable to the staff.
Of the 14 remaining breaks, 12 were resolved using fracture mechanics analyses acceptable to the staff.
The analy~es demonstrated that a postulated pipe track, the size of which would result in a 10-gpm leak, would be stable in the presence of seismic and normal operating loads for a period of time well in excess of that necessary to detect the pipe leak.
The remaining two postulated pipe breaks (one in the charging line and one in the letdown line) could cause damage to the instrument lines for steam generator pressure and level indication as well as the instru-ment lines for the pressurizer pressure and level indication.
The staff con-cludes that the loss of these monitoring instruments is unacceptable and will, therefore, require that the licensee provide protection for these lines.
In a related matter, the licensee, in accordance with Section 4.15.2 of NUREG-0820, will submit a request for amendment to modify the Technical Speci-fications concerning.the operability of leak detections systems.
2.4 Topic III-6, Seismic Design Considerations (NUREG-0820, Section 4.10) 10 tFR 50 (GDC ~), as implemented by SRP Sections 2.5, 3.7, 3.8,_3.9, and 3.10 and SEP review criteria (NUREG/CR-0098, "Development of Criteria for Seismic Review of Seletted Nuclear Power Plants 11 }, requires that structures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes.
In the topic evaluation, dated December 18, 1981, the staff concluded that there were three issues that required additional evaluation.
These issues are (1) structural integrity of small-diameter piping with motor-operated valves (2) *structural integrity of electrical components (3) functional capability of equipment The licensee, in letters dated October 13, 1982, January 18, 1983, and June 15, 1983, submitted information concerning the evaluation of the three open items.
The staff reviewed the information and, -in a letter dated September 6, 1983, concluded that open issues concerning the seismic design of the Palisades Plant still exist.
Specifically, the staff will require that:
Palisades SEP 2-3
(1) _ The licensee should either reevaluate the seismic issues using the site-specific spectrum as the input ground motion or demonstrate that*
there is enough margin inherent in the equipment design to make up the difference between the site-specific spectrum and the 0.2 g Housner ground response spectrum used by the licensee.
(2)
The licensee should either show that the effect of the additional amplification of the equipment response (resulting from out-of-plane vibration of floors and walls) is negligible or demonstrate that the equipment can withstand the additional seismic loads without affecting its structural loads.
(3)
The licensee should submit evaluation results for control room panels C-llA and C-126 and switchgear lD.
(4)
The licensee should provide justification for qualifying control panel C-33 on the basis of the evaluation of motor control centers (MCCs) 1 and 2 (the structure of the two items is different).
(5)
The acceleration used as the vertical component of seismic input for the MCC 1 and 2 internal device anchorage evaluation is too low.
The licensee should calculate the natural frequencies of the internal device anchorage system first and then obtain the spectral acceleration from the corres-ponding floor response spectra.
(6)
The licensee should develop a plan and schedule to implement the generic cable tray evaluation guidelines developed by the SEP Owners Group.
When this information is submitted to the staff for review, the staff will issue a supplemental evaluation.
2.5 Topic III-7.B, Design Codes, Design Criteria, Load Combinations, and Reactor Cavity Design Criteria (NUREG-0820, Section 4.12) 10 CFR 50 (GDC 1, 2, and 4), as implemented by SRP Section 3.8, requires that structures, systems, and components be designed for the loading that will be imposed on them and that they conform to applicable codes and standards.
In the topic evaluation, dated February 12, 1982, the staff concluded that 22 specific areas of design code changes potentially applicable to the Palisades design had been identified where the current code requires substantially greater safety margins than the earlier version of the code, or where no original code provision existed.
The licensee committed, in the integrated assessment, to review the NRC evaluation to determine applicability and perform, on a sampling basis, an evaluation of the code, load, and load combination changes on existing 11as-built 11 structures to assess the level of conservatism in the design.
The licensee responded to the staff 1s safety evaluation in letters dated October 8, 1982, and September 23, 1983.
A detailed review of the licensee 1 s October 8, 1982 submittal is presented in a staff evaluation dated November 1, 1983.
Additional information was supplied in the licensee 1 s September 23, 1983 submittal; however, clarification and additional information is still required to resolve a number of the open items.
Those open items are identified in the staff safety evaluation.
Palisades SEP 2-4
The licensee has presented analyses and qualitative arguments and concluded that the existing structures have an adequate margin of safety to accommodate the differences identified.
This information has been sufficient to resolve the issues in many cases, while others require clarification or additional information.
In no case has a difference been identified which has required a plant modification to restore a margin of safety.
On this basis the staff concludes that the additional information is confirmatory in nature and is required to provide an adequate documentation of the results of this evalua-tion.
The results of the staff 1 s review of the requested clarifications and information will be addressed in a supplemental evaluation.
2.6* Topic VIII-4, Electrical Penetrations of Reactor Containment (NUREG-0820, Section 4.26) 10 CFR 50 (GDC 50), as implemented by RG 1.63 and Institute of Electrical and Electronic Engineers (IEEE) Std 317-1972, requires that penetrations be designed so that the containment structure can accommodate, without exceeding the design leakage rate, the calculated pressure, temperature, and other environmental conditions resulting from any loss-of-coolant accident (LOCA).
In the topic evaluation, dated July 31, 1981, the staff concluded that with a LOCA environment inside containment, the low-voltage ac and de penetrations do not comply with IEEE Std 317-1972, Paragraphs 4.2.4 and 4.2.5, and RG 1.63, Paragraph Cl, regardless of the initial assumed temperature, because the operating time of the backup circuit breaker is excessive.
However, the staff noted that the licensee, in a letter dated June 15, 1981, proposed to review the containment electrical penetrations an~ document the following:.
(1) appropriate interrupting capacity for all power circuit penetrations (2) appropriate interrupting capacity for sampled control and instrument circuit penetrations (3) appropriate surveillance testing for circuit protective devices (4)
- modifications needed to conform to current licensing criteria The staff found this proposal acceptable, but since the licensee committed to this review before the integrated assessment, this topic was not evaluated in the IPSAR.
However, this topic was listed in the IPSAR to track the progress of the licensee 1s proposed program.
The licensee has kept the staff informed of progress in this area by letters dated October 12, 1982 and February 11, 1983.
The staff has reviewed this information and issued an SER on June 10, 1983~
The staff concluded that the design of the Palisades Plant electrical penetrations is similar to those in other SEP plants, the probability of electrical failure of the penetrations is low, and any resultant leakage path would be small.
The staff based its con-clusions, in part, on evaluations that have been conducted for other SEP plants.
The staff concluded that no further action was required for R. E. Ginna (NUREG-0821), Oyster Creek (NUREG-0822), Dresden Unit 2 (NUREG-0823), Millstone Unit 1 (NUREG-0824), and Yankee (NUREG-0825).
That conclusion was based on the very small contribution of electrical penetration failure to containment failure Palisades SEP 2-5
by leakage (i.e., the size of a potential leakage path *is small and the proba-bility of penetration electrical failure resulting in leakage is low).
Subse-quently, the staff reached the same conclusion for the Palisades Plant.
Therefore, this issue is considered to be fully resolved.
2.7 Topic IX-3, Station Service and Cooling Water Systems (NUREG-0820, Section 4.27) 2.7.l Cooling of CCW Heat Exchanger 10 CFR 50 (GDC 44) requires that for onsite electric power system operation (assuming offsite power is not available), the ultimate heat sink cooling water system safety function can be accomplished, assuming a single failure.
In the topic evaluation, dated February 22, 1982, the staff concluded that with a loss of offsite power and the single failure of diesel generator 1-2, suf-ficient service water flow might not be provided to prevent design temperatures from being exceeded in the component cooling water (CCW) system.
Loss of diesel generator 1-2 results in the loss of two of the three service water pumps.
The licensee performed an analysis of this scenario and submitted the results in letters dated October 15, 1982 and January 16, 1983.
The staff issued an evaluation of this issue on May 31, 1983.
The staff found that the licensee 1 s proposal to revise the plant emergency operating procedures to require the alignment of the fire water system to the service water system, if there is only one service water pump available, acceptable.
The licensee has revised the procedures.
This issue is considered to be fully resolyed.
2.7.2 Flooding of Safety ;Systems in Intake Structure 10 CFR 50 (GDC 4), as implemented by SRP Section 3.6.1, requires, in part, that structures, systems, and components important to safety be appropriately protected against dynamic effects, such as pipe whip and discharging fluids, that may result from equipment failures.
In the topic evaluation, dated February 22, 1982, the staff concluded that the current licensing criteria were not met because flooding or spraying of safety systems, such as the service water pumps which are located in the intake structure, may occur if the circulating water piping, or other piping or pumps, ruptured.
In order to preclude this flooding from occurring, the licensee, in a letter dated March 17, 1982, proposed the following plant modifications, which the staff, as detailed in NUREG-0820, found acceptable:
(1)
Provide adequate drainage in the intake structure for postulated leaks or breaks.
(2)
Provide alarms in the control room to indicate occurrence of flooding in the water intake building.
(3)
Ensure that the operator can react in sufficient time to prevent inundation of the service water pump.
Palisades SEP 2-6
(4)
Provide spray protection for the service water pumps.
Subsequent to this proposal, the licensee analyzed the drainage in the intake structure and provided information to the staff in letters dated August 25, 1982 and February 10, 1983, to show that adequate drainage exists with only the existing 4-in. floor drains.
The licensee concluded that Items (2) and (3) were, therefore, not necessary.
However, to increase the margin between calculated drainage flow and the maximum postulated break flow, the licensee has proposed replacing one 26-in. manhole cover with a grated cover to increase drainage flow.
The staff, in a letter dated May 31, 1983, found the proposed actions accept-able.
Implementation of the proposed actions is scheduled for the 1983 refuel-ing outage.
This issue is constdered to be fully resolved.
2.8 Topic IX-5, Ventilation Systems (NUREG-0820, Section 4.28) 2.8.1 Ventilation of Auxiliary Feedwater Pump 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 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 February 11, 1982, the staff concluded that a loss of offsite power may potentially create a common-mode failure~ that is,.
loss of offsite power would result in a loss of ventilation to the auxiliary feed pump room, which could potentially cause the failure of both auxiliary feed pumps.
The licensee, in a letter dated November 1, 1982, provided the results of tests conducted in the auxiliary feed pump room in which the ventilation damper was shut.
The test results indicated that the temperatures at the pump motors would not exceed 156°F given the design outside ambient temperature of 95°F sp~cified in the Final Safety Analysis Report.
The staff, in a [[letter::05000255/LER-1983-035-03, /03L-0:on 830505,safety Injection Tank (SIT) C Reached Tech Spec Limit of 198 Inches.Caused by Minor Leakage Past Loop Check Valve & SIT Check Valve or Fill & Drain Valve.Leak Rate Being Monitored|letter dated June 3, 1983]], found that the licensee 1 s test results were adequate to resolve this issue if the licensee qualifies the pump to 160°F.
The qualification of these pumps will be conducted as part of the environmental qualification program at the Palisades Plant in accordance with 10 CFR 50.49, 11 Environmental Qualification Schedules.
11 This issue is considered to be fully resolved.
2.8.2 Ventilation of Cable-Spreading, Switchgear, and Battery Rooms 10 CFR 50 (GDC 4), as implemented by SRP Section 9.4.1, requires that systems and components important to safety shall 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.
Palisades SEP 2-7
In the topic evaluation, dated February 11, 1982, the staff concluded that the ventilation system that services the cable-spreading, switchgear, and battery rooms is neither safety grade nor is it supplied by emergency power.
The licensee, in a letter dated November 1, 1982, provided the results of tests conducted to determine the effects of the loss of ventilation in the cable-spreading and switchgear rooms.
The staff, in a [[letter::05000255/LER-1983-035-03, /03L-0:on 830505,safety Injection Tank (SIT) C Reached Tech Spec Limit of 198 Inches.Caused by Minor Leakage Past Loop Check Valve & SIT Check Valve or Fill & Drain Valve.Leak Rate Being Monitored|letter dated June 3, 1983]], concluded that:
(1)
The two switchgear rooms, lC and 10, can withstand a loss of ventii'ation for long periods -of time without a significant increase in temperature because there are no appreciable heat sourc~s located within these rooms.
On the basis of *this information, the staff considers this issue resolved.
(2)
The cable-spreading room can withstand a loss of ventilation for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> before the upper design limit of 104°F is exceeded.
In addition, high room temperature is annunciated in the control room.
The operator is instructed by procedure to start fan V-47 when the high room tempera-ture alarm is annunciated.
Fan V-47 is capable of being connected to emergency power sour.ces.
On the basis of this information, the staff considers this issue resolved.
(3)
The inverter cabinets, charging cabinets, and auxiliary feedwater junction boxes J-569 and J-570 require forced air cooling even during periods of normal room ventilation as well as during periods when ventilation is not available.
Jo correct this situation, the licensee has committed to.add fans for this equipment to ensure adequate ventilation for an event causing a loss of room ventilation.
The staff finds this commitment acceptable and considers the issue resolved.
Implementation of 'this modification has been scheduled by the licensee for the 1983 refueling outage.
The redundant battery room exhaust fans were recently connected to separate safety-related motor control centers; therefore, at least one fan would be operational with a loss of offsite power.
The staff, in the evaluation dated June 3, 1983, concluded that this plant modification was an acceptable method to resolve the battery room ventilation issue.
Palisades SEP 2-8
~~~--------~----------------------------*~
VI Ill
- 0.
CD VI SEP Topic No.
Supplement Section No.
II-3.B 2.1 II-3.B.l 2.1 II-3.C 2.1 III-1 2.2 III-5.A 2.3 III-6 2.4 III-7.B 2.5 Table 2.1 Summary of supplement evaluations NUREG-0820 Section No.
Title 4.2 4.3 4.4 4.5 4.9 4.15.2 4.10 4.12 Flooding Potential and Protection Requirements Capability of Operating Plant To Cope With Design-Basis Flooding Conditions Safety-Related Water Supply (Ultimate Heat Sink [UHS])
Classification of Structures, Components, and Systems Effects of Pipe Breaks on Structures, Systems, and Components Inside Containment Operability of Leakage-Detection Systems Seismic Design Considerations Design Codes, Design Criteria, Load Combina-tions, and Reactor Cavity Design Criteria Tech. Spec.
Modifications Required From SEP Review No Yes Corrective Action Requirements None None None None Propose corrective measures for two pipe breaks (charging line and letdown line).
Provide Technical Specification regarding operability of leakage-detection systems to containment Complete evaluation of specific components.
Complete evaluation of
~oecific structural l ddQS.
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SEP Topic No.
VIII-4 IX-3 IX-5 Supplement Section No.
2.6 2.7.1 2.7.2 2.8.l 2.8.2 NUREG-0820 Section Table 2.1 (Continued)
No.
Title 4.26 4.27.1 4.27.2 4.28.1 4.28.2 Electrical Penetrations of Reactor Containment Cooling of CCW Heat Exchangers Flooding of Safety Systems in Intake Structure Ventilation of Auxiliary Feedwater Pump Room Ventilation of Cable-Spreading, Switchgear, and Battery Rooms Tech. Spec.
Modifications Required From SEP Review No No No No Corrective Action Requirements Non.e Revise plant emergency operating procedures to require the alignment of fire water system if only one service water pump available.
(1) Replace 26-in. man-hole cover in intake structure with grated cover.
(2) Provide spray pro-tection for service water pumps.
Qualify feedwater pumps to 160°F in the Environ-mental Qualification of Electrical Equipment Program.
Add fans to provide cooling to inverter cabinets, charger cabi-nets, and auxiliary feedwater junction boxes J-569 and J-570 in the event of loss of room ventilation.
3 STATUS
SUMMARY
OF IMPLEMENTATION ACTIONS In Sections 3 and 4 of the IPSAR (NUREG-0820) and in Section 2 of this supple-
~1 rnent to the IPSAR, Technical Specifications changes and procedural or plant modifications resulting from the Palisades integrated assessment have been identified.
Table 3.1 is a summary of the status for each of the implementa-tion actions.
The licensee is scheduled to complete these actions by the date indicated in the table and upon completion, the Office of Inspection and Enforcement (Region III) will verify the completed action.
'/
. I The following section of this supplement details a request by the licensee to postpone the date of proposed modifications to the main steam isolation valves so that alternative corrective actions could be evaluated.
Table 3.1 also lists other issues for which evaluations still must be completed.
Supplemental evaluations will be prepared as these issues are completed.
3.1 Topic XV-2, Spectrum of Stearn System Piping Failures Inside and Outside Containment (PWR) (NUREG-0820, Section 4.30.1)
As detailed in Section 4.30.l of the IPSAR, the licensee proposed to complete appropriate plant modifications by the end of the next refueling outage that would prevent a single failure of a main steam line isolation valve (MSIV) from causing both steam generators to blow down inside containment.
The staff concluded that the MSIV modification proposed in the integrated assessment was appropriate primarily because (1) decay heat removal without an isolated steam generator is an event that is beyond the analyzed design basis for Palisades and, therefore, decay heat removal under such circumstances cannot be insured and (2) there had been several MSIV failures at Palisades so that the postulated single failure has a fairly high probability.
Decay heat removal through the secondary system is necessary for such events, because the high-pressure safety injection system does not have sufficient capacity to permit decay heat removal through the primary system by feed and bleed.
In a letter dated August 15, 1983, the licensee requested that this modification be deferred for approximately 1 year so that alternative corrective actions might be evaluated further.
In that submittal, the licensee concluded that the MSIV modification might not be cost effective on the basis of the results of recent probabilistic analyses of the accident sequence.
This conclusion is partly due to automating and adding a third pump to the auxiliary feedwater system.
The time extension was requested to confirm critical assumptions in the probabilistic analysis and to study alternative corrective actions to increase the reliability of decay heat removal using the secondary system.
The staff believes that the proposed evaluation of alternatives is desirable because it might result in a more effective corrective action that would provide a broader range of protection (i.e., higher reliability for several events).
In a safety evaluation dated September 14, 1983, the staff stated that continued plant operation in the interim is acceptable because of the low probability of the pipe break and the relative reliability of the existing MSIVs resulting Palisades SEP 3-1
from corrective actions taken by the licensee.
Inservice inspection of the main steam piping conducted during the refueling outages in 1979 and 1983, in accordance with Technical Specification 4.12, showed no evidence of significant deterioration in the critical welds sampled since the plant began commercial operation in 1971.
These inspections have confirmed the low probability of the pipe break.
The causes of MSIV failures experienced in the past have been identified and corrected, and MSIV operability is verified during each refueling outage by the valve closure time test required by Technical Specification 4.8.
The staff recommends that the MSIV tests in the current (1983) outage be conducted as close to normal operating temperature as practical to enhance the confirmation of valve reliability.
On this basis, the staff concludes that the time extension requested by the licensee to continue the evaluation of this issue is acceptable.
The results of the evaluation are to be submitted by September 1, 1984.
-If, as a result of this further evaluation, the licensee concludes that the MSIV modification is necessary, that modification is to be made during the refueling outage currently scheduled for 1986.
Palisades SEP 3-2
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III-2**
III-3. C III-4.A III-5. A III-6
.III-7.A Table 3.1 Section No.*
Title 4.6.1 Safety Injection and Refueling Water (SiRW) and Condensate Storage Tanks 4.7.1 4.8.1 2.3(S) 2.4(S) 3.3.1 Coolin~ Waie~ System Structurei Inspection Safety Injection and Refueling Water (SIRW) and Condensate Storage Tanks Effects of Pipe Breaks on Structures, Systems, and Compcinerits Inside Containment Seismic Design Considerations Seismic Design Considerations 4.11.1 Tendon Force Acceptance Criteria See footnotes at end of table.
Integrated assessment summary Corrective Action Requirements Develop procedures to achieve cold shutdown using alternate sources of water and safety-grade equipment if preferred sources of water are not available.
Perform inspection as described in IPSAR Section 4.7.1.
Same as in Section 4.6.1.
Licensee Completion Date 1983 refueling***
outage Start routine inspections in 1982 1983 refueling outage Propose corrective measures for two 12/15/83 postulated pipe break locations (charging line and letdown line).
Complete seismic evaluation*of 4/1/84 specific co~ponents, Upgrade anchorage and support all safety-related equipment.
Brace diesel oil storage tank.
Develop acceptance crite~ia for each tendon that vary with time.
Complete 6 months before next inservice inspection (ISI)
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SEP Section Topic No.
No.*
Title III-7.A 4.11.3 III-7.B 2.5(S)
III-7.C 4.13 V-5 2.3(S)
V-10. B 4.16.1 4.16.2 V-11.A 4.17 Inspect Tendon-End Anchorages Design Codes, Design Criteria, Load Combina-tions, and Reactor Cavity Design Criteria Delamination of Pre-stressed Concrete Containment Structures Effects of Pipe Breaks on Structures, Systems, and Components Inside Containment.
Overpressurization Protec-tion of Shutdown tooling System Use of Safety-Grade Systems for Safe Shutdown Requirements for Isolation of High-and Low-Pressure Systems See footnotes at end of table.
Corrective Action Requirements Inspect end anchorages under load if new significant cracks are noted during the tendon inservice inspection.
Complete evaluation of specific components.
Perform one-time delamination inspection of containment dome.
Provide Technical Specifications for operability of leakage-detection systems.
Place pressurization system in service before the shutdown cooling system.
Develop procedures to achieve cold shutdown using safety-grade systems in the event nonsafety-grade systems are unavailable.
Require verification of check valve closure before criticality after each use of low-pressure safety injection system for shutdown cooling.
Licensee Completion Date Begin at next ISI (1986)
Complete 12/15/83 Complete 1983 refueling outage Complete
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No.*
Title VI-4 4.20.3 Manual Isolation Valve 4.20.4 Threaded Pipe Connection 3.3.5 VI-6 4.21 VII-LA 4.23.1 VII-3 4.24.1 Containment Isolation Systems Containment Leak Testing Inadequate Isolation Removal of Nonessential Loads as an Alternative to GDC 17.
See footnotes at end of table.
Table 3.1 (Continued)
Corrective Action Requirements Change manual isolation valve on penetration 44 to power-operated valve.
Modify threaded pipe connection on penetration 19 to meet containment standards.
Place block valve in front of threaded caps or weld cap on the following penetrations:
13, 17, 17a, 21, 2la, 28, 29, 48, and 73.
Verify airlock door seal integrity within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of each opening or the first of a series *of openings, whenever containment integrity is required.
Licensee Completion Date 1983 refueling outage 1983 refueling outage Complete Technical Specifica-tion revisions submitted on 3/3/82 Install. qualified isolation devices 1983 refueling outage on steam generator A and B pressure channels and on the reactor coolant flow.channel going to the plant computer.
Dev~lop procedure to remove nones-1983 refueling outage sential Joads from the battery if the immediate sources of offsite and onsite power are *not available to ensure a 6-hour battery **
capabil Hy.
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SEP Topic No.
VII-3 VIII-2 Section No.*
Title 4.24.2 Component Cooling Water Surge Tank Level and Indication 3.3.2 3.3.3 Systems Required for Safe Shutdown Onsite Emergency Power Systems VIII-3.A 4.25 Station Battery Capacity Test Requirements VIII-3. B 3. 3. 4 DC Power System Bus Voltage Monitoring and Annunciation IX-3 2.7.l(S) Cooling of CCW Heat Exchangers 2.7.2(S)
Flooding of Safety Systems in Intake Structure See footnotes at end of table.
Corrective Action Requirements Install another level sensor to the component cooling water surge tank and its indicator in the control room.
Replace battery system with one of larger capacity.
Licensee Completion Date 1983 refueling outage Complete Supply separate an~unciators for Complete (1) control switch not in automatic,r (2) overspeed, (3) overcrank, and (4) loss of de control.
Conduct battery tests* in accordance 1983 refueling outage with IPSAR Section 4.25.
Instal.l three new alarm inputs:
(1) 125-V de tie breaker open (both buses), (2) public address system inverter loss of voltage (bus 010 only), and (3) battery undervoltage (both batteries).
Complete Revise plant emergency operating Complete procedures to require the alignment of fire water system if only one service water pump is available.
Replace 26-ih. manhole cover in intake structure with grated cover and provide spray protection for service water pumps.
1983 refueling outage
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IX-5 Section No.*
2.8.l(S)
Table 3.1 (Continued)
Title Ventilation of Auxiliary Feedwater Pump Room Corrective Action Requirements Qualify feedwater pumps to 160°F in the Environmental Qualification of Electrical Equipment Program.
Licensee Completion Date In accordance with 10 CFR 50.49 2.8.2(S) Ventilation of Cable-Spreading, Switchgear, and Battery.Rooms Add fans to provide cooling to 1983 refueling outage IX-6 4.29 XV-2 3.l(S) 4.30.2 Fire Protection Spectrum of Steam System Piping Failures Inside and Outside Containment Failure of Main Feedwater
'Isolation inverter cabinets, charger cabinets, and auxiliary feedwater junction boxes J-569 and J-570 in the event of loss of room ventilation.
Will be reviewed generically outside the context of SEP.
Perform cost~benefit analysis on 9/1/84 main steam line isolation valve/
main steam system failure leading to blowdown of two steam generators.
Being reviewed by licensee under IE Bulletin 80-04, 11Analysis of a PWR Main Steam Line Break With Continued Feedwater Addition.
11
- IPSAR section number; supplement section number if followed by (S).
- To be coordinated with TMI Action Plan Item I~C.l.
- When completion date refers to a refueling outage, the item will be completed before restart of the facility.
I I
4 REFERENCES Code of Federal Regulations, Title 10, 11 Energy, 11 U.S. Government Printing Office, Washington, D.C. (includes General Design Criteria).
Consumers Power Company, 11 Final Safety Analysis Report, 11 September 1, 1970.
Letter, June 15, 1981, from R.A. Vincent (CPCo) to D.M. Crutchfield (NRC),
Subject:
Palisades Plant - SEP Topic VIII-4, Electrical Penetrations of Reactor Containment
---, July 31, 1981, from D. M. Crutchfield (NRC) to D. P. Hoffman (CPCo),
Subject:
SEP Topic VIII-4, Electrical Penetrations of Reactor Containment.
---, December 4, 1981, from D. M. Crutchfield (NRC) to D. P. Hoffman (CPCo),
Subject:
Palisades - SEP Topic III-5.A, Effects of Pipe Break on Structures, Systems and Components Inside Containment.
---,December 18, 1981, from D. M. Crutchfield.(NRC) to D. P. Hoffman (CPCo)
Subject:
SEP Safety Topics III-6, Seismic Design Consideration, and III-11, Component Integrity.
---, December 28, 1981, from D. M. Crutchfield (NRC) to D. P. Hoffman (CPCo),
Subject:
SEP Topic III-1, Quality Group Classification of Components and Systems.
---, February 11, 1982, from T. V. Wambach (NRC) to D. P. Hoffman (CPCo),
Subject:
Forwarding Draft Evaluation Report on SEP Topic IX-5, Ventilation Systems. -
---, February 12, 1982, from T. V. Wambach (NRC) to D.. P. Hoffman (CPCo),
Subject:
Systematic Evaluation Program III-7.B, Design Codes, Design Criteria and Load Combinations.
---,February 19, 1982~ from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
Palisades - SEP Topics II-3.A, Hydrologic Description; II-3.B, Flooding Potential and Protection Requirements; II-3.B.l, Capability of Operating Plants To Cope With Design-Basis Flooding Conditions; and II-3.C, Safety-Related Water Supply (Ultimate Heat Sink).
---, February 22, 1982, from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
SEP Topic IX-3, Station Service and Cooling Water System.
---, March 17, 1982, from R. A. Vincent (CPCo) to D. M. Crutchfield (NRC),
Subject:
SEP Topic IX-3, Station Service and Cooling Water Systems.
---, March 23, 1982, from R. A. Vincent (CPCo) to *D. M. Crutchfield (NRC),
Subject:
SEP Topics II-3.B, II~3.B.1, II-3.C, Hydrologic Considerations..
Palisades SEP 4-1
---, August 12, 1982, from R. A. Vincent (CPCo) to D. M. Crutchfield (NRC),
Subject:
SEP Topic III-1, Classification of Structures, Components and Systems (Seismic and Quality).
---, August 16, 1982, from R. A. Vincent (CPCo) to D. M. Crutchfield (NRC),
Subject:
SEP Topic III-5.A, Effects of Pipe Break on Structures, Systems and Components Inside Cqntainment~.
---, August 25, 1982, from D. J. VandeWalle (CPCo) to D. M. Crutchfield (NRC),
Subject:
SEP Topic IX-3; Station Service and Cooling Water Systems.
---, October 7, 1982, from T. V. Wambach (NRC) to_D. J. VandeWalle (CPCo),.
Subject:
Palisades - SEP Topics II-3,A, Hydrologic Description; II-3.B,.
Flooding Potential and Protection Requirements; II-3.B.l, Capability of Operating Plants To Cope With Design Basis Flooding Conditions; and II-3.C, Safety Related Water Supply (Ultimate Heat Sink).
---, October 8, 1982, from D. J. VandeWalle (CPCo) to D. M. Crutchfield (NRC),
Subject:
Palisades Plant - SEP Topic III-7.B, Design Codes, Design Criteria and Load Combinations.
---, October 12, 1982, from K. A. Toner (CPCo) to D. M. Crutchfield (NRC);
Subject:
SEP Topic VIII-4, Ele~trical Penetrations of Reactor Containment.
---, October 13, 1982, from K. A. Toner (CPCo) to D. M. Crutchf{eld (NRC),
Subject:
Palisades Plant - SEP Topic III-6, Seismic Design.Considerations.
---, O~tober is, 1982~ from_K; A. Toner (CPCo) t6 D. M. Crutchfield (NRC),
Subject:
SEP Topi~ IX-3, Station Service and Cooling Water Systems.*
---, November 1, 1982, from K. A. Toner (CPCo) to D. M. Crutthfield (NRC),
Subject:
SEP Topic IX-5, Ventilation Systems - Submittal of Switchgear.
Room, Cable Spreading Room and Auxiliary FeedwaterPump Room Test Results.
---, December 9, 1982, from K. A. Toner. (CPCo) to D. M. Crutchfield (NRC),
Subject:
SEP Topic III-5.A, Effects of Pipe Break on Structures) Systems~
and Components Inside Containment.
---, January 6, 1983, from K. A. Toner (CPCo) to D. M. Crutchfield (NRC),
Subject:
.SEP Topic IX-3, Station Service and Cooling Water Systems, Summary of Service Water System Performance Adequa~y Analysis.
---, January 18, l983, from K. A. Toner (CPCo) to D. M. Crutchfield (NRC),
Subject:
Palisades Plant - SEP Topic III-6, Seismic Design Considerations, Response to Commitment Concerning Electrical Components Evaluations.
---, February 10, 1983, from K. A. Toner (CPCo) to D. M.. Crutchfield (NRC),
Subject:
SEP Topic I~-3, Station Service and Cooling Water Systems.
---, February 11, 1983, from K. A. Toner. (CPCo) to D. M. Crutchfield (NRC),
Subject:
SEP Topic VIII-4, Electrical Penetrations ~f Reactor Containment -
Status Update of Program To Evaluate the Adequacy of Penetration Protection From Overload and Short-Circuit Conditions.
, May 31, 1983, from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
- Integrated Pl aht Safety Assessment Report Section 4. 27, Station Service and Cooling Water Systems.
-~~) June 3~ 1983, from T, V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
Integrated Plant Safety Assessment Report Section 4.28, Venti-lation Systems.
---, June 10, 1983, from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
Integrated Plant Safety Assessment Report SecUon 4.26, Electrical Penetrations of Reactor Containment.
---, June 15, 1983, from K. A. Toner (CPCo) to D.. M. Crutchfield (NRC),
Subject:
Palisades Plant - Systematic Evaluation Program Topic fII-6, Seismic Design Considerations, Subtopic:
Seismic Evaluation of Electrical Raceway Systems.
---, June 22, 1983, from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
IPSAR Sections 4.9 and 4.15, Pipe Breach Inside Containment and Reactor Coolant Pressure Boundary Leakage Detection Requirements.
---, June 27, 1983, from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
Integrated Plant Safety Assessment Report Section 4.5, Classi-fication of Structures, Systems and Components. -
---, August 15, 1983, from K. A. Toner (CPCo) to D. M. Crutchfield (NRC),
Subject:
Palisades Plant - SEP Topic XV-2, Spectrum of Steam Piping Failures Inside and Outside Containment (PWR) - Request for Time Extension To Resolve Concern of Two Steam Generator Slowdowns Inside Containment.
---, September 6, 1983, from D. M. Crutchfield (NRC) to D. J. VandeWalle (CPCo),
Subject:
Palisades Plant - IPSAR Section 4.10, Seismic Design Considerations.
---, September 14, 1983, from D. G. Eisenhut (NRC) to D. J. VandeWalle (CPCo),
Subject:
Evaluation of Main Steam Line Break - Palisades Plant.
---, September 23, 1983, from K. A. Toner (CPCo) to D. M. Crutchfield (NRC),
Subject:
Palisades Plant - SEP Topic III-7.8, Design Codes, Design Loads and Load Combinations - Evaluation of Steel Embedment.
---, November 1, 1983, from D. M.
C~utchfield (NRC) to D. J. VandeWalle (CPCo),
Subject:
Integrated Plant Safety Assessment Report (IPSAR) Section 4.12, Design Codes, Design Criteria and Lead Combinations for the Palisades Plant.
U.S. Nuclear Regulatory Commission, NUREG-0800 (formerly NUREG-75/087),
11Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Pl ants, 11 July 1981.
---, NUREG-0820, 11 Integrated Plant Safety Assessment, Systematic Evaluation Program, Palisades Plant, 11 Final Report, October 1982.
---, NUREG-0821, 11 Integrated Plant Safety Assessment, Systematic Evaluation Program, R. E. Ginna Nuclear Power Plant, 11 Final Report, December 1982.
Palisades SEP
---NUREG-0822, 11 Integrated Plant Safety Assessment, Systematic Evaluation Program, Oyster Creek Nuclear Generating Station, 11 Final Report, January 1983.
---, NUREG-0823, 11 Integrated Plant Safety Assessment, Systematic Evaluation Program, Dresden Nuclear Power Station Unit 2, 11 Final Report, February 1983.
---, NUREG-0824, 11 Integrated Plant Safety Assessment, Systematic Evaluation Program, Millstone Nuclear Power Station Unit 1, 11 Final Report, February 1983.
---, NUREG-0825, 11 Integrated Plant Safety Assessment, Systematic Evaluation Program, Yankee Nuclear Power Station, 11 Draft Report, February 1983.
---, NUREG/CR-0098, 11 Development of Criteria for Seismic Review of Selected Nuclear Power Plants, 11 by N. M. Newmark and W. J. Hall, May 1978.
---, Regulatory Guide (RG) 1.26, 11Quality Group Classifications and Standards for Water-, Steam-and Radioactive-Waste-Containing Components of Nuclear Power Plants.
11 RG 1.46, 11 Protection Against Pipe Whip Inside Containment.
11 RG 1.59, 11 Design Basis Floods for Nuclear Power Plants.
11
, RG 1.63, 11 Electric Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants.
11 INDUSTRY STANDARD Institute of Electrical and Electronics Engineers (IEEE), Std 317-1972, 11 Electrical Penetration Assemblies in Containment Structures for Nuclear Pow~r Generating Stations (ANSI N45.3-1973):
Palisades SEP 4-4
Palisades SEP APPENDIX A REFERENCES TO CORRESPONDENCE FOR EACH TOPIC EVALUATED
IP SAR Supplement Section No.
2.1 2.2 2.3 2.4 2.5
- 2. 6 '
2.7 2.8 3.1 Palisades SEP Date Reference 10/7/82 Letter from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
Palisades - SEP Topics II-3.A, Hydrologic Description; II-3.B, Flooding Potential and Protection Requirements; II-3.B.1, Capability of Operating Plants To Cope With Design Basis Flooding Conditions; and II-3.C, Safety Related Water Supply (Ultimate Heat Sink).
6/27/83 Letter from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
Integrated Plant Safety Assessment Report Section 4.5, Classification of Structures, Systems and Components.
6/22/83 Letter from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
IPSAR Sections 4.9 and 4.15, Pipe Break Inside Containment and Reactor Coolant Pressure Boundary Leakage Detection Requirements.
9/6/83 Letter from D. M. Crutchfield (NRC) to D. J.
VandeWalle (CPCo),
Subject:
Palisades Plant - IPSAR Section 4.10, Seismic Design Considerations.
11/1/83 Letter from D. M. Crutchfield (NRC) to D. J. VandeWalle (CPCo),
Subject:
Integrated Plant Safety Assessment Report (IPSAR) Section 4.12, Design Codes, Design Criteria and Load Combinations for the Palisades Plant.
- 6/10/83 Letter from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
Integrated Plant Safety Assessment Report Section 4.26, Electrical Penetrations of Reactor Containment.
5/31/83 Letter from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
Integrated Pl ant Safety Assessment Report Section 4.27, Station Service and Cooling Water Systems.
6/3/83 Letter from T. V. Wambach (NRC) to D. J. VandeWalle (CPCo),
Subject:
Integrated Plant Safety Assessment Report Section 4.28; Ventilation Systems.
9/14/83 Letter from D. G. Eisenhut (NRC) to D. J. VandeWalle (CPCo),
Subject:
Evaluation of Main Steam Line Break - Palisades Plant.
A-1
APPENDIX B NRC STAFF CONTRIBUTORS AND CONSULTANTS Palisades SEP
This supplement to Safety E*valuation Report is a product of the NRC staff and its consultants.
The NRC staff members listed below were principal contribu-tors to this report. A list of consultants follows the list of staff members.
NRC STAFF Name M. Boyle S. Brown P. Y. Chen T. Cheng D. Chery C. Grimes E. McKenna T. Michaels D. Persinko R. Scholl T. Wambach CONSULTANTS Name D. Barrett T. Stilwell Palisades SEP Title Integrated Project Manager Integrated Project Manager Senior Mechanical Engineer Senior Structural Engineer Integrated Project Manager Section Leader Integrated Project Manager Senior Project Manager Integrated Project Manager Senior Project Manager Senior Project Manager Company Franklin Research Center Franklin Research Center B-1 Branch Systematic Evaluation Program Systematic Evaluation Program Systematic Evaluation Program
- systematic Evaluation Program Systematic.Evaluation Program Systematic Evaluation Program Systematic Evaluation Program Systematic Evaluation Program Systematic Evaluation Program Systematic Evaluation Program Operating Reactors Branch #5 IPSAR Supplement Section Report Date 2.5 September 1983 2.5 September 1983
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION
- 1. REPORT NUMBER (Assigned by DDCJ 111-811 BIBLIOGRAPHIC DATA SHEET NUREG-0820, Supplement No. 1
- 4. TITLE AND SUBTITLE (Add.Volume No., if appropriate)
- 2. (Leave blank)
Integrated Plant Safety Assessment Report, Systematic Evaluation Program, Palisades Plant, Consumers Power
- 3. RECIPIENT'S ACCESSION NO.
Comnanv-Docket No. 50-255
- 7. AUTHOR{S)
- 5. DATE REPORT COMPLETED MONTH I YEAR November 1983
- 9. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code)
DATE REPORT ISSUED Division of.Licensing MONTH I YEAR Office of Nuclear Reactor Regulation November 1983
- u. s. Nuclear Regulatory Commission
- 6. (Leave blank J Washington, D.C.
20555
- 8. (Leave blank I
- 12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (/ncludeZ1j:; Code)
- 10. PROJECT/TASK/WORK UNIT NO.
Same as #9 above.
- 11. FIN NO.
- 13. TYPE OF REPORT I PERIOD COVERED (Inclusive dates)*
Technical Evaluation
- 15. SUPPLEMENTARY NOTES 114. (Leave blank I n-*-"-~.;~~ +,... n"rlro+ 1\\1" a:;n_?a:;a:;
- 16. ABSTRACT (200 words or less)
The Nuclear Regulatory Commission (NRC) has published its Supplement No. 1 to the Final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0820) under the scope of the Systematic Evaluation Program (SEP), for Consumers Power Company's Palisades Plant located in Covert, Van Buren County, Michigan.
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 refi~ed engineering evaluations or the continuation of ongoing evaluations after the Final IPSAR for the Palisades Plant was issued.
The review has provided for (l)* an assessment of the significance of differences between current technical positions on.selected safety issues and those that existed when the Palisades 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 supplements will form part of the bases for considering the conversion of the license.
- 17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS Systematic Evaluation Program l 7b. IDENTIFIERS/OPEN-ENDED TERMS
- 18. AVAILABILITY STATEMENT
- 19. SECURITY CLASS (This report}
- 21. NO. OF PAGES Unclassified Unlimited
- 20. tfiC~ITY CWiS dTh1s page}
- 22. PAI.CE nc ass1 ie s
NRC FORM 335 111*811
UNITED STATES NUCLEAR REGULATORY COMMISSION
. WASHINGTON. D.C. 20555 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE. $300 FIRST CLASS MAIL POSTAGE & FEESPAIO USN RC WASH
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