ML20052G831
| ML20052G831 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 04/30/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0787, NUREG-0787-S03, NUREG-787, NUREG-787-S3, NUDOCS 8205190042 | |
| Download: ML20052G831 (200) | |
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NUREG-0787 Supplement No. 3 Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3 Docket No. 50-382 Louisiana Power & Light Company l
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation April 1982
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4 NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in N RC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
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Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.
The following documents in the NUREG series are available for purchase from the NRC/GPO Sales Program: formal NRC staff and contractor reports, NRC sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission issuances.
Documents available from the National Technical Information Service inciude NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
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aroceedings are available for purchase from the organization sponsoring the publication cited.
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NUREG-0787 Supplement No. 3 I
Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3 Docket No. 50-382 Louisiana Power & Light Company
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U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation l
April 1982 l
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TABLE OF CONTENTS Page 1.
INTRODUCTION AND GENERAL DISCUSSION..........................
1-1 1.1 Introduction............................................
1-1
- 1. 7. Summa ry o f Outs ta ndi ng I s s ues...........................
1-1 1.8 Co n f i rma to ry I s s u e s.....................................
1-2 1.9 License Conditions......................................
1-2 4.
REACT 0R......................................................
4-1 4.4 Thermal-Hydraulic 0esign................................
4-1 5.
REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS.................
5-1 5.2 Integrity of Reactor Coolant Pressure Boundary..........
5-1
/
5.2.2 Overpressurization Protection....................
5-1 5.3 Reactor Vessel..........................................
5-1 5.3.3 Reactor Vessel Integrity.........................
5-1 5.4 Component and Subsystem Design.........................
5-2 5.4.3 Shutdown Cooling System.........................
5-2 l
7.
INSTRUMENTATION AND CONTROLS............
7-1 I
l 7.4 Systems Required for Safe Shutdown......................
7-1 l
7.4.3 Emergency Shutdown from Outside the Control Room.
7-1 l
9.
AUXILIARY SYSTEMS............................................
9-1 9.1 Fuel Storage Facility...................................
9-1 9.1.4 Fuel Handling System.............................
9-1 i
l 9.4 Heating, Ventilation, and Air Conditioning (HVAC) l Systems.................................................
9-1 i
9.4.1 Control Room Area Ventilation System (Control Room Air Conditioning System)....................
9-1 9.4.2 Spent Fuel Pool Area Ventilation System (Fuel Handling Building Ventilation System)............
9-2 i
l Waterford SSER 3 i
TABLE OF CONTENTS (Continued)
PBA*
9.5 O the r Aux i l i a ry Sys tems.................................
9-2 9.5.1 Fire Protection..................................
9-2 9.5.1.1 Introduction............................
9-2 9.5.1.2 Fire Protection System's Description and Evaluation..........................
9-3 9.5.1.3 Other Items Related to Fire Protection Programs...............................
9-5 9.5.1.4 Fire Protection of Saft Shutdown Capability..............................
9-6 9.5.1.5 Emergency Li ghti ng......................
9-8 9.5.1.6 Fire Protection for Specific Areas......
9-8 9.5.1.7 Administrative Controls and Fire Brigade.................................
9-11 9.5.1.8 Technical Specifications................
9-11 9.5.1.9 Conclusion..............................
9-11 9.5.4 Emergency Diesel Engine Fuel Oil Storage and Transfer System.................................
9-11 9.5.5 Emergency Diesel Engine Cooling Water System....
9-12 9.5.6 Emergency Diesel Engine Air Starting System.....
9-12 9.5.7 Emergency Diesel Engine Lubricating 011 System..
9-12 10.
STEAM AND POWER CONVERSION SYSTEM............................
10-1 10.3 Main Steam Supply System................................
10-1 10.3.1 Main Steam Supply System (Up to and Including the Main Steam Isolation Valves).................
10-1 10.4 Other Features of the Steam and Power Conversion System.
10-1 10.4.9 Auxiliary (Emergency) Feedwater System..........
10-1 13.
CONDUCT OF 0PERATIONS........................................
13-1 13.3 Emergency Preparedness Evaluation.......................
13-1 13.3.1 Introduction....................................
13-1 13.3.2 Eval uation of the Applicant's Plan..............
13-1 13.3.3 Review of State and Local Plan by FEMA..........
13-4 13.3.4 Conclusions......................................
13-4 l
l 15.
ACCIDENT ANALYSIS............................................
15-1 15.2 Normal Operation and Anticipated Transients.............
15-1 Waterford SSER 3 11
TABLE OF CONTENTS (Continued)
Page 15.2.3 Decrease in Reactor Coolant Flow Rate............
15-1 15.2.3.2 Single Reactor Coolant Pump Sheared Shaft...................................
15-1 18.
REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.......
18-1 LIST OF APPENDICES APPENDIX A Continuation of Chronology of Radiological Review.............................................
A-1 APPENDIX B Continuation of Bibliography.......................
B-1 APPENDIX C ACRS Letters.......................................
C-1 APPENDIX D Errata to Safety Evaluation Report.................
D-1 APPENDIX E Principal Contributors.............................
E-1 APPENDIX F FEMA Report........................................
F-1 1
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i Waterford SSER 3 iii
1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction On July 9, 1981, the Nuclear Regulatory Commission (NRC) issued a Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3.
On October 13, 1981, the staff issued Supplement No. I to the SER and on January 13, 1982 the staff issued Supplement No. 2 to the SER.
This supplement updates the SER by providing the staff's evaluation of informa-tion submitted by the applicant since the SER and Supplement Nos. 1 and 2 were issued.
This supplement also includes a copy of the supplemental repcrt by the Advisory Committee on Reactor Safeguards dated March 9, 1982.
1 Each of the following sections of this supplement is numbered the same as the section of the SER that is being updated and the discussions are supplementary to and not in lieu of the discussion in the SER.
Appendix A is a continuation of the chronology of the safety review.
Appendix B is an updated bibliography.
Appendix C is a copy of the ACRS reports.
Appendix D is a list of errata.
Appendix E is a list of principal contributors to this SSER.
Appendix F is the FEMA report.
The Project Manager is Suzanne Black; she may be reached on (301) 492-7119.
1.7 Summary of Outstanding Issues Section 1.7 of the SER and Supplement Nos. 1 and 2 contained a list of out-standing issues.
This supplement addresses the resolution of a number of issues previously identified as open. These are listed below, along with the section of this report wherein their resolution is discussed.
(1) Fire Protection (7.4, 9.5.1) l (2) Emergency Planning (13.3)
(3) Rea-tor coolant pump shaft break (15.2.3.1)
I (4) Thermal-hydraulic design (4.4)
At this time there remain a number of safety issues that have not yet been resolved.
These will be addressed in a future supplement to the SER.
The following is a list of these items.
(1) Fire Protection-(7.4, 7.5, 7.7, 9.5.2)
(2) PSI /ISI (3.9.6, 5.2.4, 6.6)
(3) Environmental Qualification (3.11)
(4) Seismic Qualification (3.10)
(5) Reactor coolant pump shaft break analysis (7.1, 7.2, 7.3, and 7.5)
(6) Thermal-hydraulic design (4.4)
(7) Indemnity requirements (21) i (8) Site hazards (explosions) (2.2)
(9) Licensee qualifications (13.1, 13.2)
(10) Turbine missiles (3.5.1.3, 3.5.3)
Waterford SSER 3 1-1
i (12) TMI Issues Operating procedures (I.C. tasks-long term)
Control room review (I.D.1)
Containment system design (II.E.4.2)
ICC instrumentation (II.F.2) 1.8 Confirmatory Issues Confirmatory issues are those which were essentially resolved to the staff's satisfaction but for which certain confirmatory information had not yet been provided by'ation and has confirmed the preliminary conclusions.
the applicant.
For the following issues, the staff has received that inform (1) Sizing of primary safety valves (5.2.2)
(2) Diesel engine piping (9.5.4.2, 9.5.5, 9.5.6, 9.5.7)
At this time several issues remain for which the staff has not yet received the neccssary confirmatory information.
These issues, which are listed below, will be addressed in a future supplement to the SER.
(1) Testing the ultimate heat sink (2.4)
(2) Piping analyses (3.9.2)
(3) Containment isolation actuation signal (7.3)
(4) Performance of PWR relief and safety valves-(22)
(5) Shutdown initiation using safety grade equipment (5.4.3)
(6) Containment sump vortex text (6.3.3)
(7) Boron dilution events (5.4.3, 15.2.4.4)
(8) Emergency feedwater control (7.3)
(9) Feedwater line break analysis (15.3.2)
(10) Clarification of transient analyses with potential for fuel damage (15.3.1) 1.9 License Conditions In addition to those issues listed in the SER as requiring a license condition to ensure that NRC requirements are met during plant cperation, the staff has identified the following license condition:
(1) LP&L must respond to rapid depressurization capability prior to fuel load or provide a justification for safe operation of the olant in the interim (5.4.3)
Waterford SSER 3 1-2
4 REACTOR 4.4 Thermal-Hydraulic Design In the SER the staff required that the applicant perform the following:
1.
demonstrate the applicability of the CE-1 critical heat flux (CHF) corre-lation and the proposed limit value to the Waterford 3 fuel design; 2.
supply'the necessary information for the Core Protection Calculator (CPC) system by March 1982; and 3.
perform the safety analyses which account for the new fuel design and power distributions in future cycles.
Item (1) was resolved in Sup,,lement No. 1 to the Waterford 3 SER.
This supple-ment addresses the remaining items.
With regard to the CPC system, the staff requested that the applicant provide the following:
1.
a definition of software algorithm and data changes from previously approved CPC systems, giving the reason for changes and a definition of the change; 2.
a description of the conduct and result.s of Phase I and II implementation testing.
The description will include test cases, test results, errors found in testing, and corrective action.
3.
the Software Functional Specifications upon which the Waterford 3 CPC j
systems are based; 4.
the data base constants used in the CPC system algorithms and the assump-tions and methodology used in the data base design; and 5.
evaluation of the CPC system response to design basis transient events by comparison to safety analyses and to other versions of the CPC system soft-ware which have been previously evaluated by the staff.
l In an October 2, 1981 meeting, the staff projected that it would need seven months to complete our review and issue an SER on the CPC modifications. As a result of this meeting, Combustion Engineering (CE) stated in an October 30, 1981 letter (Scherer) that it will submit the algorithm description of the software changes, the test plan for Phase I and II testing, and the summary of the safety analyses by March 31, 1982.
This was submitted on March 31, 1982.
They further stated that the Phase I and II test reports would be submitted by June 30, 1982.
CE felt that this schedule would allow adequate time for a detailed staff review and provide sufficient time for incorporating technical modifications needed to satisfy the staff and CE.
The staff will complete its review of the Waterford 3 CPC system prior to fuel load.
Waterford SSER 3 4-1
However, the burden of resolving any concerns identified during staff review, in a timely manner, is the responsibility of the applicant.
In response to Open Item (3) the applicant has submitted an amended response to a staff question on future cycle operation and a report describing the effects of spacer grid loss coefficients on the minimum departure from nucleate boiling ratio (DNBR).
The applicant response stated that the physics parameters used in the FSAR analyses are primarily. based on the first-cycle core.
However, the method used to select values for parameters such as Nuclear Powcr Factor and the rod radial power factors is based on one limiting combination of parameters for full power
. operation. This methodology offers considerable flexibility in accommodating future cycle variations in individual parameters since other combinations pro-vide acceptable results.
Similarities in predicted radial power distributions for cycles up to and including Cycle 4 show that no significant differences are expected in power distributions from cycle to cycle.
Based on this response, the staff concludes that there is reasonable assurance of sufficient design thermal margin to permit full power operation during future cycles.
The applicant's submittal also included a report (CE Report Un-numbered) which discussed the effects of the HID-1 spacer grid loss coefficients on minimum DNBR.
The staff has reviewed this report and has requested additional infor-mation to support the conclusion given by the applicant in response to this-open item.
Based on staff review of the information supplied by the applicant, the staff concludes that the latter portion of Open Item (3) (future cycle operation) has been resolved.
However, additional information is required to resolve our concerns regarding the new fuel design.
Summary This SER supplement addresses the following open items:
1.
the applicant's commitment to supply the needed CPC information by March 1982; and 2.
the applicant's response to account for the effects of the new fuel design and power distributions in future cycles on the safety analyses.
The applicant has committed to a schedule of CPC information submittals to be completed by a June 30, 1982 submittal of the Phase I and II test reports.
The staff has agreed to a best effort to complete our review prior to the i
scheduled fuel load.
The staff concludes that the applicant's response-to our concerns on future cycle operation'is acceptable.
However, before our concerns on the new fuel design can be resolved, the staff requires additional informa-tion on the effects of grid design and spacing on the minimum DNBR.
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'Waterford SSER 3 4-2
5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.2 Overpressurization Protection Section 5.2.2 of the SER stated that the SRP Section 5.2.2 requires that the high pressure reactor trip or second safety grade scram signal, whichever occurs later, should be used for sizing the primary system safety valves.
The staff requires the app 1': ant to confirm that this criterion is met in sizing Waterford 3 safety valves.
In Amendment 25 to the FSAR, the applicant has stated that an analysis has been performed to demonstrate that the sizing of the primary system safety valves is adequate if it is assumed that the reactor is tripped on the second safety grade trip signal.
The loss of load event (the design basis event for sizing the primary safety valves) has been reanalyzed with no credit taken for Doppler feedback on core power and it was assumed that the reactor is tripped on low steam generator level, approximately eight seconds after the high pressurizer pressure trip setpoint is reached.
The peak RCS pressure is well below the limit of 110 percent of design pressure.
Based on the above, the staff has concluded that the Waterford 3 design meets the acceptance criteria in SRP (NUREG-75/087) Section 5.2.2 and is acceptable.
5.3 Reactor Vessel 5.3.3 Reactor Vessel Integrity Pressurized Thermal Shock (PTS), as a consequence of certain postulated acci-dent scenarios, is of concern primarily for vessels that have experienced significant degradation of material properties due to irradiation damage in the beltline region during operation.
The staff's Unresolved Safety Issue (USI) A-49 will address this issue for all PWR facilities, but initially it is concerned primarily with operating facilities.
The rate of degradation of material properties is related to the concentration cf trace elements in the vessel materials, especially copper, nickel and phosphorus.
The phenomenon of radiation damage versus accumulated fluence is accounted for in Regulatory Guide 1.99 which is used by the staff to conservatively predict material property degradation until sufficient data from irradiation specimens are accumulated for a particular vessel.
The Waterford 3 reactor vessel has a predicted end-of-life RT at the inside NDT wall of about 100 F.
This was calculated for the limiting material in the belt-line, the intermediate shell plate M-1004-2, which had 0.03 percent copper, 0.005 percent phosphorus, 0.58 percent nickel, and an initial RT f 22 F.
NDT Waterford SSER 3 5-1
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l The end-of-life (EOL) fluence at the inside wall is predicted to be 3.68 x 1018 n/cm2 (E > MeV).
Regulatory Guide 1.99, Rev. I was used to estimate the adjustment of reference temperatures; hence, the value of 100*F EOL. RT is NDT l
believed to be conservative.
The staff believes that the Waterford 3 reactor vessel will not be jeopardized by pressurized thermal shock for 32 effective full power years because of the relatively low E0L. RTNDT (1004).
- However, the staff is continuing to study this issue as Unresolved Safety Issue No. A-49 and, if necessary, may establish new requirements for Waterford's continued operation.
5.4 Component and Subsystem Design 5.4.3 Shutdown Cooling System In a letter to Chairman Palladino dated December 15, 1981, the Advisory Committee for Reactor Safeguards (ACRS) requested that the staff give consideration to the potential for adding valves sized to facilitate rapid depressurization of the CESSAR primary coolant system to allow more direct methods of decay heat removal.
A copy of this letter is included in Appendix C.
The Waterford 3 reactor coolant system (RCS) is designed without power operated relief valves (PORVs) on the pressurizer.
Decay heat removal capability ulti-mately relies on heat removal via steam generators using emergency feedwater and atmospheric steam dump valves.
The auxiliary feedwater system for this plant has been reviewed and found to meet the reliability criterion of staff Technical Position ASB 10-1.
Because of this, the staff has concluded that the addition of a rapid depressurization capability is not required to achieve a " feed and bleed" mode of decay heat removal as a result of loss of auxiliary feedwater.
However, the ACRS letter and the recent steam generator tube rupture event at Cinna have led the staff to reexamine the reliability of steam generator integrity for decay heat removal over the life of the plant.
In particular, the staff is considering the need for a rapid depressurization capability in the event of tube failures in both steam generators.
In addition, the staff is looking at the benefits of providing this capability to afford the operator greater flexibility to respond to unfore-ceen events (e.g., ATWS).
The :,taff has requested that information be provided by CE as to the need for a rapid depressurization capability in the CE System 80 design.
CE has provided an initial response which asserts that the CE System 80 design is adequate without this capability, and suggested that any plant design modifications might more appropriately be directed to providing a rapid depres-surization capability for the secondary system and utilizing additional water sources for feeding the steam generators.
The staff has reviewed CE's response and has requested additional information from CE, as well as from the Waterford and San Onofre applicants.
Responses to the staff's additional requests for l
information have not yet been received.
If the responses are not provided prior to the anticipated feel load date for Waterford 3, the staff has required that the applicant provide a justification for safe operation of the plant in the interim.
The staff has briefed the ACRS Subcommittee on Decay Heat Removal Requirements (on March 16, 1982) and the full ACRS in an Executive Session (on April 2, 1982)
Waterford SSER 3 5-2
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on the status of the staff's evaluation of this matter.
Subsequently, the ACRS issued a letter dated April 5, 1982 which stated that while this evaluation should be conducted expeditiously, its resolution should not now be a condition for operation of CE System 80 plants at full power or of plants having similar features.
In addition, the letter stated that the need for future hardware or procedural changes should be contingent upon the results of this evaluation.
A copy of this letter is included in Appendix C to this SSER.
Should the NRC decide that design or procedural changes are necessary, LP&L will be required to implement them for Waterford 3.
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Waterford SSER 3 5-3
7 INSTRUMENTATION AND CONTROLS 7.4 Systems Required for Safe Shutdown 7.4.3 Emergency Shutdown From Outside the Control Room It is the staff's position that in order to meet General Design Criterion (GDC) 19 SRP Section 7.4, the design should provide redundant safety grade capability to achieve and maintain hot shutdown from a location or locations remote from the control room, assuming no fire damage to any required systems and equipment and assuming no accident has occurred.
Also, the design should provide redun-dant safety grade capability for attaining subsequent coli shutdown through the use of suitable procedures.
Section 7.4.1.5 of the Waterford FSAR and the safe shutdown analysis submitted by letter dated December 21, 1981 describe the local control panel (LCP-43) design and capability.
The design objective of the LCP-43 is to provide a cen-tral point to monitor plant shutdown and provide certain control functions in the event of an evacuation of the control room.
The LCP-43 contains redundant safety grade controls and instrumentation to enable the operator to achieve and maintain the plant in the hot standby condition.
For cooldown from hot standby (cold shutdown) there are redundant controls and instrumentation on the local control panel (LCP-43) and associated suitable procedures.
The procedures stress operator verification that all support equipment are functioning and, if not, then direct the operator to start equipment via LCP-43 or from separate local panels.
The controls and instrumentation for redundant equipment are mounted in separate sections of the local control panel so that no single failure can prevent safe shutdown of the reactor.
All safety-and nonsafety-related channels within the local control panel are physically and electrically separated in accordance with Regulatory Guide 1.75.
The Waterford 3 safe shutdown analysis assumes that offsite power is unavail-able. Therefore, the additional equipment available with offsite power is not relied upon.
The approach ensures that the more conservative analysis enve-lopes the more realistic (offsite power available) case.
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Various inconsistencies exist between the FSAR and safe shutdown analysis tables which list the instrumentation and controls required for safe shutdown l
from outside the control room. Also, during our review, it was discovered that the hot leg reactor coolant temperature indication was not listed as being on the local control panel. Through discussions, the applicant has agreed to update the subject tables in the FSAR and safe shutdown analysis to make them c
consistent.
This is acceptable.
A concern was developed during staff review pertaining to the negation of auto-matic actuation of engineered safety features (ESF) functions.
The applicant was asked to describe compliance to the staff's position which says that the design should be such that the manual transfer of control to the remote location (s)
Waterford SSER 3 7-1
should not disable any automatic actuation of ESF functions while the plant is attaining or maintained in hot shutdown, other than where ESF features are manually placed in service to achieve or maintain hot shutdown. The applicant responded through discussions stating that the Waterford design does not fully comply with the staff's position.
The applicant has agreed to either modify the design to meet our position or provide additional information to justify the existing design.
The resolution of this item will be addressed in a future supplement to the SER.
Based on staff review of the FSAR and safe shutdown analysis and pending the applicant's satisfactory resolution of the item discussed above, the staff concludes that the Waterford design provides sufficient redundant safety-grade controls and instrumentation on the local control panel (LCP-43) to allow the operator to achieve and maintain hot standby.
Also, there is sufficient redun-dant safety grade capability for obtaining cold shutdown through the use of suitable procedures.
This meets cur GDC 19 requirements as interpreted by SRP Section 7.4.
However, the applicant must update the FSAR and safe shutdown analysis to include redundant safety grade reactor coolant system hot leg temperature indication on the local control panel (LCP-43) and to correct any inconsistencies which exist.
See Section 9.5.1.4 of this SSER for a discussion regarding the fire protection for the safe shutdown capability as it pertains to the requirements of Appendix R.
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Waterford SSER 3 7-2
9 AUXILIARY SYSTEMS 9.1 Fuel Storage Facility 9.1. 4 Fuel Handling System In Amendment 25, the applicant revised the cask crane electrical interlock system. The crane is provided with bridge and trolley track type limit switches which provide a warning to the crane operator that the crane has travelled beyond its proper position.
Mechanical backups (bridge bumpers and trolley chocks) to the limit switches ensure that the cask cannot be transported over the spent fuel pool or the bulkhead gate.
Thus, the requirements of General Design Criterion 61, " Fuel Storage and Handling and Radioactivity Control," are satisfied, and the above modification is therefore acceptable.
9.4 Heating, Ventilation, and Air Conditioning (HVAC) Systems 9.4.1 Control Room Area Ventilation System (Control Room Air Conditioning System)
In Amendments 20 and 25 to the FSAR, the applicant revised the emergency operating modes for the control room air conditioning system as follows:
The control room air conditioning system is designed to automatically maintain the control room and associated areas within the environmental limits required for operation of plant controls and uninterrupted safe occupancy of required manned areas during all operating modes including LOCA conditions.
The system is designed to maintain the control room in either an isolation (full recir-culation) mode or under positive pressure.
Redundant radiation and toxic gas detectors are located near the normal system outside air intake.
Receipt of a high radiation signal from the normal outside air intake detectors or a safety injection actuation signal (SIAS) automatically closes (isolates) the normal 4
outside air intake and exhaust, stops the normal exhaust fans, opens the recir-culation dampers, and starts the emergency filtration units, thus initiating the positive pressure emergency operating mode.
The control room air conditioning system recirculates the air with a portion passing through the emergency and charcoal filters for cleanup.
The operator may open either of the two separate emergency outside air intakes (the one with the lowest concentration of radio-activity) to provide additional air for pressurization of the control room.
In the pressurized mode, all outside air is also passed through the emergency and charcoal filters.
Receipt of a toxic gas signal automatically initiates the isolation (full recirculation) emergency operating mode.
Under this condition, the same automatic actions indicated above for the positive pressure emergency mode will occur except that the emergency filtration units are not started.
The control room air conditioning system recirculates 100% of the air and no outside air is provided.
In the event that a toxic gas signal occurs after a SIAS or high radiation signal, any open outside air intakes would be automatically closed.
Waterford SSER 3 9-1
l The design described above conforms to the requirements of General Design Cri-terion 19, " Control Room," and the guidelines of Regulatory Guide 1.95,
" Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," with respect to assuring environmental limits _ for proper operation of plant controls and safe occupancy of the control room under all normal and accident conditions including LOCA conditions, and is, therefore, acceptable.
9.4.2 Spent Fuel Pool Area Ventilation System (Fuel Handling Building Ventilation System)
In Amendment 22 to the FSAR, the applicant revised the normal operating logic design for the fuel handling building ventilation system as follows:
During normal operation, the fuel handling building air handling unit is j
started and one of the two redundant full-capacity normal exhaust fans is l
started.
The operation of these fans is interlocked [the exhaust fan cannot i
run unless the supply fan (air handling unit) is operating] to assure proper pressurization of the fuel handling building areas from uncontaminated to poten-tially contaminated zones.
This is an acceptable method for providing normal fuel handling building ventilation.
Operation of the system in emergency con-ditions is not affected by this change and the evaluation in the SER is not affected.
l 9.5 Other Auxiliary Systems 9.5.1 Fire Protection 9.5.1.1 Introduction The staff has reviewed the Waterford 3 program reevaluation and fire hazards i
analysis submitted by the applicant by letter dated July 1, 1977, including
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i FSAR Amendments 1 through 24 for conformance to Appendix A to BTP ASB 9.5-1 and Appendix R to 10 CFR Part 50.
These requirements have been included in NUREG-0800 section 9.5-1 as BTP CMEB 9.5-1 therefore, the review is also in conformance with BTP CMEB 9.5-1.
As part of the review, the staff visited the plant site to examine the relationship of safety related components, systems i
and structures in specific plant areas to combustible materials and to fire j
detection and suppression systems.
The overall objective of the review was to ensure that in the event of a fire, personnel and the plant equipment would be adequate to safely shut down the reactor, to maintain the plant in a safe shut-i down condition, and to minimize the release of radioactivity to the environment.
The review included an evaluation of the automatic and manually operated water and gas fire suppression systems, the fire detection systems, fire barriers, fire doors and dampers, fire protection administrative controls, and the fire brigade size and training.
The staff consultant, Gage Babcock, Inc. participated in the preparation of -
this SSER.
P l-Waterford SSER 3 9-2 i.
- 9. 5.1. 2 Fire Protection Systems Description and Evaluation (1) Water Supply Systems:
The water supply system consists of three fire pumps separately connected to a buried, 10-inch pipe loop around the plant.
The fire pumps are rated at 2000 gpm at 100 psig head. One pump is electric motor driven and two are diesel engine driven.
The fire pumps are located in a separate fire pump house and are separated by 3-hour fire-rated walls.
Automatic sprinklers are provided to protect the fire pump building.
The fire pumps and their controllers are UL listed.
Their design and installation conforms to the requirements of NFPA 20 " Standard for the Installation of Centrifugal Fire Pumps."
A separate jockey pump, rated at 200 gpm at 100 psig, maintains the yard fire main pressure at 100 psig.
If the fire main pressure drops to 90 psig, the electric motor-driven fire pump will automatically start.
The first diesel engine-driven fire pump will start automatically if t.he pressure drops to 85 psig, and the other diesel pump will start auto-matica11y if the pressure falls to 80 psig.
Separate audible and visual alarms are provided in the control room for each pump to monitor pump operation, prime mover availability, power failure, and failure of a fire pump to start.
All valves in the fire protection water supply system that control automatic suppression systems are electrically supervised with alarms in the control room. All other valves in the water supply system are locked in their normal position with strict key control procedures.
The pumps take suction from two 260,000 gallon vertical ground-level water storage tanks, arranged so that all fire pumps can take suction from either or both tanks.
However, as presently designed, either a break in the pump suction header or in the pump discharge lines, including the cross-connaction between the two discharge lines, could cause the loss of all three fire pumps.
The applicant has committed to install additional valves in the fire pump suction header and discharge lines to preclude a single break from shutting down all three fire pumps.
As initally designed, the fire water system supplied lubricating water for the circulating water air evacuation pumps and the circulating water pumps.
The applicant has committed to disconnect the fire water system as a source of lubricating water so that the fire water supply system provides water only for the fire protection system.
The greatest water demand for the fixed fire suppression systems is 1200 gpm I
and, coupled with 750 gpm for hose streams creates a total water demand 1950 gpm.
The fire pumps can deliver the required water supply with one pump out of service.
Additionally, the 260,000 gallon water storage tanks are adequate to supply the required flow for a 2-hour duration.
Yard hydrants are connected to the yard fire main at intervals of approxi-mately 250 ft.
Each two hydrants are provided with a hose house located l
between the hydrants, and each hydrant lateral is provided with an isolation valve to facilitate hydrant maintenance and repairs without shutting down part of the fire water supply system.
Waterford SSER 3 9-3
Based on this evaluation, the staff concludes that the water supply system is now adequate, meets the guidelines of Section C.2 of Appendix A to BTP ASB 9.5-1, and is, therefore, acceptable.
(2) Sprinkler and Standpipe Systems:
The automatic sprinkler systems and standpipe risers are connected to common interior water supply headers.
The interior headers are fed from each end through separate supply connections to the looped yard system with appropriate valves so that sections can be isolated to perform maintenance or to prevent a single break from impairing the entire distribution system.
In addition, header and divisional valve arrangement is such that no single failure can impair both primary and backup fire protection systems protecting a single fire area.
The water supply valves to the suppression systems are electrically supervised with alarms in the control room.
In additioa, the sprinkler systems have waterflow alarms which alarm in the control room.
The automatic sprinkler systems, e.g., wet pipe sprinkler systems, pre-action sprinkler systems, on-off multi 'ycle sprinkler system, and water spray system will be designed to the recommendations of National Fire Protection Association (NFPA) Standards No. 13, " Standard for the Installation of Sprinkler System and No.15, " Standard for Water Spray Fixed Systems."
The areas that have been equipped with automatic water suppression systems are as listed in the applicant's FSAR Table V-1, and supplemented in its letter dated December 21, 1981, Table 9.5A-6.
Manual hose stations are located throughout the plant to ensure that an effective hose stream can be directed to any safety-related arei in the plant except for the diesel generator day tank rooms, in the Reactor Auxiliary Building, and zone RAB 35.
By letter dated December 21, 1981, the applicant committed to install additional standpipe hose stations so that an effective hose stream can be directed to all parts of these areas with a maximum of 100 ft. of hose at a station.
The standpipe stations to be added will be supplied from a different section of the supply header than the automatic water suppression system in each area.
The standpipes are consistent with the requirements of NFPA 14, " Standard for the Installation of Standpipe and Hose Systems." However, the 2-inch diameter standpipes do not meet the guidelines of Appendix A which call for minimum 4-and 2-1/2-inch diameter pipe sizes for multiple and single hose station supplies, respectively.
The applicant performed hydraulic calculations to determine the flow and pressures available through the standpipe system.
These calculations indicated that 200 gpm is available at 65 psi in the standpipe system except at two remote areas.
The supply to these areas can provide 200 gpm at 54 psi and the staff finds that acceptable.
Based on this evaluation, the staff concludes that, with the indicated modifications the sprinkler and standpipe systems are adequate, meet the guidelines of Appendix A to BTP ASB 9.5-1, and are, therefore, acceptable.
(3) Gaseous Fire Suppression Systems: A halon total flooding system is used as the primary extinguishing agent in the nonsafety-related computer room Waterford SSER 3 9-4
underfloor spaces.
The system is designed to produce a 5 to 7 percent halon concentration with a soaking time of 15 to 20 minutes, and is designed to the reouirements of NFPA 12A, " Standard on Halogenated Fire Extinguishing Agent Systems - Halon 1301." The system is activated by cross-zoned fonization detectors.
Based on the evaluation, the staff concludes that the halon fire suppression system meets the guidelines of Appendix A to BTP ASB 9.5-1 and is, therefore, acceptable.
(4) Fire Detection Systems:
The fire detection systems consist of ionization detectors, associated electrical circuitry, electrical power supplies, and the fire annunciation panels.
The systems provide both a local audible alarm and audible and visual alarms in the control room.
Such detection systems will be installed in all areas of the plant containing safe shutdown related system components.
This includes the control room, the new and spent fuel pool areas, and areas of cable concentration.
The areas which have been provided with fire detection systems are as listed in the applicant's FSAR, Table V-1, and supplemented in its letter dated December 21, 1981, Table 9.5A-6.
The fire detection systems are designed according to NFPA 720, " Standard for the Installation, Maintenance, and Use of Proprietary Protective Signaling Systems." By letter dated December 21, 1981, the applicant has committed to design the detection systems in safety-related areas of the plant in accordance with the requirements for Class A systems in NFPA 720.
Based on our review, the staff concludes that the fire detection systems and the applicant's comitments meet the guidelines of Appendix A to BTP ASB 9.5-1 and are, therefore, acceptable.
9.5.1.3 Other Items Related to Fire Protection Programs (1) Fire Barriers and Fire Barrier Penetrations: Walls and floor / ceiling assemblies separating fire areas and zones consist of 2-or 3-hour fire rated construction.
In cases when the fire rating is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the staff has evaluated the fuel loading and fire protection provided and finds the 2-hour fire rating to be acceptable.
Also, the applicant has committed by letter December 21, 1981 to provide 3-hour fire-rated penetration seals at all penetrations of fire rated barriers.
By letter dated December 21, 1981, the applicant has committed to provide test reports to demonstrate the fire resistance rating of the penetration seals.
Based on staff review and the applicant's commitments, the staff concludes that the fire barriers and fire barrier penetration seals meet the guidelines of Section 0.1.(j) of Appendix A to BTP ASB 9.5-1.
(2) Fire Doors and Dampers: The applicant indicated that all doorways and access hatches in fire area boundaries are protected by 3-hour rated fire door assemblies.
However, during the site visit, the staff found several doors that were not labeled fire doors.
By letter dated December 21, 1981, the applicant identified all doors previously identified as 3-hour rated i
which are not labeled fire doors.
The applicant has on file at the plant documentation of the rating of the doors.
This is acceptable to the staff.
Waterford SSER 3 9-5
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Originally, the applicant had not provided automatic fire dampers at most ventilation penetration of fire-rated' walls and floors.
The applicant has since installed 3-hour fire dampers in duct penetrations thrcugh rated fire barriers.
The fire dampers provided are listed in the< applicants 21,1 981, Table 9.5A-7.
The staff finds that the letter dated December fire doors and dampers conform 'to Section 0.1 (j) of the guidelines of.
Appendix A to flip ASR 9.S-1, and are, therefore, acceptable.
9.5.1.4 Fire Protection of Safe S$.tdown Capability s
(1) Safe Shutdown Capability:
By l'etter dated December 21, 1981, the applicant provided responses to staff W ncerns involving fire protection for the safe shutdown capability in accordance with the requirements of i
Appendix R.
The applicant's safe shutdown analysis states that systems needed for hot shutdown and cold shutdown consist of redundmt trains and that one of the redundant trains needed for safe shutfown would be free of fire damage by providing separation, fire barriers, rerair (for equipment required for cold shutdown only) and/or alternative shutdown capability.
For hot shutdown, at least one train of the following shutdown systems would be available:
(1) The Emergency Feedwater System and (2) the Main Steam
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i Atmospheric Dump Valves.
For cooldown/ cold shutdag at least one train of the following shutdown systems would be available:
(1) Auxiliary-Spray Valves, (2) Charging System, and (3) Low Pressure Safety Injection System.
The safe shutdown analysis considered component s, cabling, and support equipment for systems identified above which are needed to achieve shutdown.
The applicant performed a cable separatha analysis utilizing a computerized cable and conduit list for all rooms'of the plant housing safe shutdown equipment to ensure, that at least orie train of this-equipment is available ti the event of a fi k in any of these rooms.
Safe shutdown equipment cabfing was ideatified athi traced through each fire area from the component to the power source.
The computer program then identified where redundant trains of safe shutdown systems are located within the sane liie area.
A manual separation verification in which safe shutdown system cabling was traced on general.'ar r)ngement drawings was also performedrin addition to the above com'puter method.
Corrective measures were taksn as necessary to assure proper separation.
The staff has reviewed the applicant's method and audited several arrangement drawings to verity correct appl.ication of the 5.ethodology.
The staff concludes that the applicant hd.provided"an acceptable means of defbonstrating that separation exists between redundant safe shutdown systems trains.
The applicant's analysis indicated that alternative shutdown measures were required for the control room, cable vault area (incl 0 ding the cable spreading room and cable penetration areas), and auxiliary / remote shutdown (local control) panel. area in order to assure the availability of the safe
-shutdown system as these areas contain,'rrury than one division of safe shutdown system cabling.
In the event that fire disables'the control room l
.'or; cable vault area, the 1.ocal control' panel (LCP-43), located in a i
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separate fire protected area in the reactor auxiliary building, provides an, ti l
alternative to providing fire protection separation (refer to Section 7.4.3 of this SSER).
The control functions and indications provided at the-1 Waterford SSER 3
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9-6 4
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A" local panel which are necessary for safe shutdown are electrically isolated or otherwise separate and independent from the control room.
Refer to the following section for further discussion on the alternative shutdown ca.pabil i ty.
Based on the above, the staff concludes th'at the fire protection for shutdown capability of Waterford 3 complies with the requirements of Section III.G of: Appendix R and is, therefore, acceptable.
(2) Alternative Shutdown Capability:
Section1[4.1.5oftheFSARdescribes the auxiliary / remote shutdown (local control) panel's (LCP-43) design and capability.
The design objective of LCP-43-is to provide a central point-to monitor plant shutdown and provide certain control functions in event of an' evacuation of the control room due to a fire disabling the control room or cable vault area.
Emergency procedures delineating the shutdown sequence utilizing LCP-43 are provided.
Plant operators initially verify that automatic reactor trip and turbine trip has occurred andithat the emergency feedwater system and diesel generators have automatically started.
The emergency feedwater flow control valves'and atmospheric dump valves will then be manually modulated at LCP-43 to begin plant cooldown.
The operators will also verify that all support equipment necessary for 2
achieving safe shutdown has come on automatically, 'or will turn on this equipment at LCP-43 or separate local panels.
Primary system depressurization is accomplished by normal shrinkage as cooldown progresses.
Depressurization and boration is controlled from LCP-43, utilizing the auxiliary spray valves and charging pumps.
The low pressure safety injection system is initiated and controlled from LCP-43 when the cut in temperature is achieved.
The design of' the remote shutdown panel complies with the performance goals outlined in the requiremants of.Section III.L of Appendix R.
Reactivity control is accomplished by-a manual scram before the operator leaves the control room and boron'sddition control from LCP-43 via the chemical and volume control system (charging pumpsf as indicated above.
The following direct reading of process variables is provided at the LCP-43:
1.
Emergency feedwater flow; 2.
Steam generator wide range level; 3.
Steam generator pressure; 4.
Condensate storage pool level; 5.
Reactor coolant system hot leg temperature; 6.
Reactor coolant system cold leg temperature; 7.
Pressurizer pressure;-
' 8.
Pressurizer level; 9.
Neutron flux (sourcc range).
' The above LCP-43 indications are either electrically isolated from the control room through transfer switches or are provided with power from cables routed separately from the control room and cable vault area to assure their availability in a fire in these areas.
m' ax s
Waterford SSER'3 9-7
Based on the above, the staff concludes that the local control panel
-(LCP-43) shutdown capability complies with the requirements of Section III.L of Appendix R and is, therefore, acceptable.
9.5.1.5 Emergency Lighting i
The emergency lighting system for the plant originally consisted of hard-wired a.c. lighting in areas where safety functions are performed, and 8-hour battery power supplied emergency lighting in access routes to these areas and for emergency evacuation.
Eight-hour battery pack emergency lights are required for areas of the plant r:ccessary for safe shutdown.
The applicant committed by letter dated November 10, 1981 to install self-contained 8-hour battery pack emergency lighting in all areas of the plant which must be manned to bring the plant to a safe cold shutdown and in access and egress routes to and from these fire areas.
The staff concludes that, with the indicated modifications, the emergency lighting meets the criteria of Appendix A, Section 0.5.a to BTP ASB 9.5-1 and, also, the provisions of Section III.J of Appendix R to 10 CFR Part 50 and is, therefore, acceptable.
9.5.1.6 Fire Protection for Specific Areas (1) Control Room: The control room complex is separated from other areas of the plant by 2-hour rated fire-resistant construction.
The staff has evaluated the fuel loading and protection provided and finds that the 2-hour fire rating is acceptoole.
Doors between the control room and reactor auxiliary building are 3-hour fire doors.. Originally the applicant was going to install a window at the entrance to the control room; however, this has been replaced by a 2-hour rated fire barrier.
Ducts which penetrate the walls of the control room will be provided with 3-hour fire dampers.
Fire detection, fire extinguishers, automatic sprinklers, four standpipe hose stations within approximately 75 feet of the fire area, and emergency lighting are provided for the control room complex.
The applicant committed by letter dated December 21, 1981 to install a control supplement at the remote control panel to ensure complete electrical independence from the control room after activation of the transfer switches.
The staff concludes that, with the indicated modifications, the fire protection for the control room meets the guidelines of Appendix A, Section 0.2, to BTP ASB 9.5-1 and is, therefore, acceptable.
(2) Cable Spreading Rooms:
The cable spreading room is separated from the rest of the plant by 2-hour fire rated walls and floor-ceiling assemblies.
The staff has evaluated the fuel loading and fire protection provided and finds that the 2-hour fire rating is acceptable.
All piping and tray L
penetrations are sealed as described in Section 9.5.1.3.
Ionization type detectors are provided to activate pre-action sprinkler system serving the entire area and alarm locally and in the control room.
Waterford SSER 3 9-8 l
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The automatic pre-action sprinkler system can also be activated manually.
There is one hose station inside the fire area and two hose stations within 75 feet of the' fire area.
The exposed cable in the room has been coated with a flame retardant to minimize fire propagation.
The applicant, by letter dated December 21, 1981, connitted to provide alternate shutdown around the cable spreading room.
Ingress and egress are through three Class A fire doors.
The staff concludes that, with the commitments on alternate shutdown, the fire protection for the cable spreading room meets the requirements of Appendix R to 10 CFR Part 50.
(3) Emergency Diesel Generator Rooms:
The diesel generators and day tanks are each located in different fire areas separated by 3-hour fire-rated barriers.
Each diesel generator room is protected by an automatic pre-action sprinkler system and alarm to the control room.
Portable fire extinguishers are provided inside the area and there are four hose stations within 75 feet of these areas.
The applicant committed by letter dated December 21, 1981 to provide watertight curbs at all entrances to the diesel generator rooms.
The staff was concerned that B train circuits essential for safe shutdown would be damaged by a fire in the A diesel generator room.
By letter dated December 21, 1981, the applicant agreed to relocate all B train circuits outside of the A diesel generator room.
Based on the evaluation and the indicated modifications, the staff concludes that the fire protection for the diesel generator rooms meets the guidelines of Appendix A, Section D.9 to BTP ASB 9.5-1 and is, therefore, acceptable.
(4) Diesel Generator Day Tank Areas:
The diesel generator day tank areas are separated from each other and from the balance of the plant by 3-hour fire-rated barriers.
Each day tank room is protected by an automatic sprinkler system and alarm to the control room.
Originally, ingress and egress for the "A" day tank area was through the "B" day tank area.
If a fire were to occur in the "A" day tank area, the fire brigade would have to enter through the "B" day tank area, to work the "A" day tank area exposing the "B" day tank area to that fire.
The staff found this unacceptable.
By letter dated December 21, 1981, the applicant committed to seal the opening between the two day tank areas, and to provide a separate entrance to the "A" day tank area. This modifi-cation provides a separate entrance to eacn day tank area.
The staff finds this arrangement acceptable.
Based on this evaluation and the indicated modifications, the staff concludes that the fire protection for the diesel generator day tank areas meets the guidelines of Appendix A, Section D.10 to BTP ASB 9.5-1 and is, therefore, acceptable.
(5) Switchgear Room:
The switchgear room is separated from the balance of the plant by 3-hour fire barriers.
The switchgear room contains both safety division circuits.
It was the staff's concern that a single fire could Waterford SSER 3 9-9
damage A, A/B, and B circuits.
The applicant committed by letter dated December 21, 1981 to provide complete 1-hour fire-rated separation of both safety division switchgear areas with complete area sprinkler suppression system in each area with 1-hour barrier on all redundant circuits in each i
area. The staff finds that these modifications provide fire protection for this area to the requirements III.G.2.c of Appendix R and are, therefore, acceptable.
Based on this evaluation and the commitment, the staff concludes that the fire protection for the switchgear room meets the guidelines of Appendix A, Section D.5 to BTP 9.5-1 and is, therefore, acceptable.
(6) Containment: The reactor containment area is separated from adjacent areas by 3-hour fire-rated barriers.
Fire protection features include hose stations, ionization smoke detectors, and fire extinguishers.
The staff finds this acceptable. An automatic multicycle sprinkler fire suppression system has been provided for the reactor coolant pumps.
By letter dated December 21, 1981, the applicant committed to provide an oil collection system for the reactor coolant pumps which meets the requirements of Appendix R,Section III.0 to 10 CFR Part 50.
The staff finds this acceptable.
Based on this evaluation, the staff concludes that the fire protection for the containment area meets the guidelines of Appendix A, Section F.1 to BTP 9.5-1 and is, therefore, acceptable.
(7) Battery Rooms:
The plant battery rooms are separated from each other and from the balance of the plant by three-hour fire rated barriers.
The ventilation system is designed to maintain the hydrogen level below 2 percent.
Ionization smoke detection systems are provided in each battery room.
Each battery room is provided with loss of ventilation alarm which annunciates in the control room.
Based on this evaluation, the staf f concludes that the fire protection for the battery rooms conforms to the guidelines of Appendix A, Section F.7 to BTP 9.5-1 and is, therefore, acceptable.
(8) Hydrogen Piping:
There is no bulk storage of flammable gases inside structures housing safety-related systems.
All such storage is maintained outdoors, away from safety equipment.
Although the applicant has stated that no hydrogen piping is routed in areas containing safe shutdown equipment or cabling, hydrogen is routed in areas containing safety-related equipment.
By letter dated February 3, 1982, the applicant stated that the only hydrogen line in safety-related areas is equipped with an excess flow check valve.
The staff finds that the hydrogen piping meets the guidelines of BTP ASB 9.5-1, Section D.2, and is, therefo"e, acceptable.
(9) Other Plant Areas:
The applicant's Fire Hazards Analysis addresses other plant areas not specifically discussed in this report.
The staff finds that the fire protection for these areas is in accordance with the guidelines of Appendix A to BTP ASB 9.5-1, and is, therefore, acceptable.
Waterford SSER 3 9-10
9.5.1.7 Administrative Controls and Fire Brigade Appendix R to 10 CFR Part 50, Sections III.H, III.I, and III.K contain NRC requirements.concerning administrative coritrols and fire brigade, which con-sists of fire protection organizations, fire brigade training, controls over combustibles and ignition source, prefire plans and procedures for fighting fires, and quality assurance.
By letter dated November 10, 1981, the applicant committed to the technical requirements of Appendix R to 10 CFR Part 50.
The staff concludes that, with the applicant's commitment, the administrative controls and fire brigade will conform to the requirements of Appendix R to 10 CFR Part 50, Sections III.H, III.I, and III.K, and are, therefore, acceptable.
9.5.1.8 Technical Specifications By letter dated December 15, 1981, the applicant committed to follow NRC's standard technical specifications.
The staff finds this acceptable.
9.5.1.9 Conclusion On February 19, 1981, the Commission approved a rule concerning fire protection, which was issued as Appendix R to 10 CFR Part 50.
The technical requirements set forth in Appendix R, as well as the criteria of BTP 9.5-1, have been used as guidelines in the staff's fire protection evaluation set forth above.
By letter dated November 10, 1981, the applicant committed to meet the technical requirements of Appendix R to 10 CFR Part 50.
This commitment, along with specific commitments described in this SSER, were evaluated by the staff.
Subsequently, the staff concluded that the fire protection program is in conformance with the guidelines of Appendix A to BTP 9.5-1, the requirements of Appendix R and GDC 3 to 10 CFR Part 50, and therefore is acceptable.
Specific verification of the commitments made by LP&L will be performed by the NRC staff prior to issuance of an operating license.
9.5.4 Emergency Diesel Fuel Oil Storage and Transfer System 9.5.4.2 Emergency Diesel Engine Fuel Oil Storage and Transfer System In Section 9.5.4.2 of the SER, dated July, 1981, the following statements were made:
"The diesel oil storage tank fill and vent lines and f'uel oil drain lines 7EG1-43 and 7EG1-44 are seismically supported.
This piping and the associated components are designed, manufactured, and will be inspected in accordance with the guidelines and requirements of ANSI Standard B31.1, " Code for Pressure Piping," ANSI N45.2, " Quality Assurance Program Requirements for Nuclear Facilities," and 10 CFR 50 Appendix B."
Based on the above statements the design of fill, vent, and drain lines had been found to be acceptable.
Waterford SSER 3 9-11
I In a letter dated December 16, 1981, the applicant stated that this piping was designed and constructed to ANSI B31.1 but that neither ANSI N45.2 nor 10 CFR 50 Appendix B quality assurance requirements were imposed.
The staff informed the applicant that the design was unacceptable and that the piping would have to be inspected in accordance with the standards and/or replaced with ASME Section III Class 3 piping.
In a letter dated January 21, 1982, the applicant proposed to modify the design so that in the event of damage to the fill and vent lines, the fuel oil storage tank can be refilled and vented.
The applicant proposed to add a seismic Cate-gory I, ASME Section III, Class 3 emergency fill connection to the cross-tie connecting Train A diesel oil transfer pump suction to Train B diesel oil transfer pump suction.
Since this connection will be located inside the diesel oil tank storage room, the applicant proposed to use hoses to connect the emergency fill connection with the tank truck.
The hose routing through the building was provided.
Venting of the tank would be accomplished by opening the manhole cover on top of the storage tank.
The staff finds the design acceptable, provided that the following condition be included in the Technical Specifications: When the emergency fill connection is being used to fill the storage tank, a fire watch will be posted along the hose routing.
The December 16, 1981 and January 21, 1982 letters stated that drain lines 7EG1-43 and 7EG1-44 were strictly for housekeeping purposes and were nonpressurized gravity feed lines from the diesel engine injector seal drip pans to the diesel oil storage tank.
The lines are safety Class 3 from the diesel engine up to and including the isolation valve adjacent to the diesel.
The balance of the lines are designed to ANSI B31.1.
The maximum expected flow in these lines is one half gallon per hour of engine operation.
The applicant stated that this amount of flow is not expected to cause any fire hazard if the oil leaks out of the drain lines and that adequate room drainage is provided.
l In addition, during engine operation, the diesel engine room will be manned, so that any leak that develops can be quickly repaired or isolated.
The staff concurs with this statement and therefore finds the above drain lines acceptable.
Based on our review, the staff concludes that the emergency diesel engine diesel oil storage and transfer system as modified meets the requirements of General Design Criteria 2, 4, 5, and 17, meets the guidance of the cited Regulatory Guides and Standard Review Plan 9.5.4, it can perform its design safety function, and meets the recommendations of NUREG/CR-0660 and industry codes and standards, and is therefore acceptable.
9.5.4.2, 9.5.5, 9.5.6, and 9.5.7 Emergency Diesel Engine Fuel Oil, Cooling Water, Air Starting and Lubrication Systems The applicant in Amendmant 21 to the FSAR provided the standards to which the engine-mounted auxiliary systems (fuel oil, cooling water, air starting, and lubrication) piping and associr.ted components were designed.
This engine-munted piping and the associated components, such as valves, fabricated headers, fabricated special fittings, and the like are designed, manufactured, and inspected in accordance with the guidelines and requirements of ANSI
. Standard B31.1, " Code for Pressure Piping," ANSI N45.2, " Quality Assurance Program Requirements for Nuclear Facilities," and 10 CFR 50 Appendix B.
The Waterford SSER 3 9-12
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engine-mounted auxiliary syste, piping and associated components are inten-i l
tionally overdesigned (subjected to low working stresses) for the application, l
and thereby result in high operational reliability.
The design of the engine-mounted auxiliary system piping and components to the cited design philosophy and standards is considered equivalent to a system designed to ASME Section III, Class 3 requirements with regard to system functional operability and inservice reliability.
Based on our review, the staff concludes that the engine-mounted piping and components of emergency diesel engine auxiliary systems (fuel oil, cooling water, air starting, and lubrication) meet the requirements of General Design Criteria 2, 4, 5, and 17, meet the guidance of the cited Regulatory Guides and Standard Review Plans, can perform their design safety function, and meet the recommendations of NUREG/CR-0660 and industry codes and standards, and are therefore acceptable.
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l Waterford SSER 3 9-13
10 STEAM AND POWER CONVERSION SYSTEM 10.3 Main Steam Supply System 10.3.1 Main Steam Supply System (Up to and Including the Main Steam Isolation Valves)
In the SER the staff stated that the main steam isolation valves (MSIVs) close on receipt of a main steam isolation signal (MSIS) or containment isolation actuation signal.
In Amendment 25 to the FSAR, the applicant indicated that the MSIVs close only on an MSIS. An MSIS is generated on low steam pressure or high containment pressure. The staff concludes that this is acceptable, as it provides adequate assurance of main steam isolation when required.
Refer to Section 7.3 of the SER for further discussion of the MSIS.
10.4 Other Features of the Steam and Power Conversion System 10.4.9 Emergency Feedwater System The applicant has informed the staff of an error in the SER discussion concerning the NUREG-0635 Additional Short Term Recommendation dealing with redundant low level indication and alarms for the emergency feedwater system primary water supply (the condensate storage pool).
In the SER the staff indicated that the condensate storage pool level instrumentation consisted of two " low-low" level annunciator windows powered by two separate power supplies which were backed by a common battery-backed power source.
The design actually consists of redun-dant level transmitters (level monitoring instrumentation loops) at the condensate storage pool powered from redundant Class 1E power sources.
Each transmitter provides redundant control room level indication and alarm (two separate annun-clator windows) of " low-low" level in the pool.
The annunciators are powered by an independent power channel which is backed by a battery-backed power source.
The " low-low" level alarm setpoint allows at least 30 minutes for operator action assuming that the largest capacity emergency feedwater pump is operating.
The staff concludes that the above response satisfies NUREG-0635's recommendation and is therefore acceptable.
Waterford SSER 3 10-1
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13 CONDUCT OF OPERATIONS 13.3 Emergency Preparedness Evaluation 13.3.1 Introduction NRC's evaluation of the state of emergency preparedness associated with Waterford 3 involves review of the applicant,'s onsite emergency preparedness plus review of the Federal Emergency Management Agency (FEMA) findings and-determinations pertaining to State and local emergency preparedness.
The Louisiana Power and Light Company (applicant or LP&L) filed with the NRC a comprehensive revision to the Waterford 3 Emergency Plan (Plan) by letter dated February 22, 1982.
The Plan, which previously was contained in Section 13.3 of the FSAR, has been revised, reformatted, and issued as a stand-alone document.
Previously, the staff had reviewed the earlier version of the Plan and issued its evaluation in the SER (NUREG-0787, dated July, 1981) and Supplement No. 1, 4
i dcted October, 1981.
The revised Plan has been reviewed against the 16 planning standards in Section 50.47 of 10 CFR Part 50, the requirements of Appendix E to 10 CFR Part 50, and the specific criteria of NUREG-0654/ FEMA-REP-1, Revision 1 i
entitled " Criteria of Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,"
November, 1980, which has been endorsed as Regulatory Guide 1.101 (Rev. 2).
Based on this review, the staff identified several areas for which additional applicant commitments and/or staff review are required to ensure that the l
criteria of NUREG-0654 are met.
Section 13.3.2 of this. report includes (1) a discussion on those areas for which acceptable commitments have been made, (2) a description of items currently under staff review, (3) the status of items previously identified in SSER No.- 1, and (4) the status of the Alert and l
Notification System.
t j
Section 13.3.3 contains a discussion on FEMA's findings and determinations on the adequacy of State and !ccal plans.
Section 13.3.4 provides the staff's conclusions.
4 13.3.2 Evaluation of the Applicant's Plan i.
13.3.2.1 Emergency Preparedness Items For Which Acceptable Commitments Have Been Made Based on the review of the Plan, as described in Section 13.3.1 above, the staff determined that additional commitments were required regarding the following items:
1 1.
Clarification of the interface between utility, Federal, State, and 1ccal l
support groups.
I J
Waterford SSER 3 13-1 i
-. ~. - -.. _, - _ _
1
2.
Assignment of onshift personnel to two functional positions in the onsite emergency organization.
(
3.
Completion of agreement letters with local hospitals and Ebasco.
4.
Identification of primary and backup means for emergency communication.
5.
Provision for relocating news media personnel in the event the primary news center becomes uninhabitable.
6.
Description of the postaccident sampling system.
7.
Description of offsite emergency notification system (see Section 13.3.2.4 for further discussion).
8.
Clarification of the duties and respons bilities of the Emergency Planning i
Coordinator.
9.
Description of onsite contamination control measures with regard to food supplies.
10.
Description of the training program for corporate personnel who are assigned to the Corporate Control Center.
11.
Clarification of the training program for the individuals responsible for the overall planning effort.
The applicant addressed each of these items in correspondence dated April 7, 1982, and has committed to revisions to the Plan to reflect the resolution of each of the above items.
This area is confirmatory in nature and the revised plan will be reviewed in each of the areas listed above to ensure conformance with the applicable criteria of NUREG-0654. A future supplement to this report will provide the staff's conclusions.
13.3.2.2 Emergency Preparedness Items Under Review The following items, previously addressed in the SER, are being reviewed by the staff as a result of recent changes to the Plan.
13.3.2.2.1 Emergency Classification System The emergency classification system previously described in Section 13.3.3.1 of the FSAR has been revised.
The revised classification system is currently under review by the staff.
A supplement to this report will provide the staff's conclusions as to its acceptability.
13.3.2.2.2 Emergency Facilities The description of the Control Room, TSC, E0F, and OSC previously described in Section 13.3.6.1 of the FSAR has been completely revised.
This description will be reviewed against the criteria contained in NUREG-0696 entitled j
Waterford SSER 3 13-2
l i
" Functional Criteria for Emergency Response Facilities," February 1981.
A supplement to this report will provide the staff's conclusion as to the acceptability of the emergency response facilities.
13.3.2.3 Items Previously Identified In SSER No. 1 This section addresses those items identified in SSER No.1 for which resolution was required.
Item 1.
Existing agreement letters must be updated with regard to access control of the 10 inile EPZ.
I Discussion The applicant has not yet clarified, by reference to State and local plans or updated letters of agreement, the access control of waterway and railways in the 10-mile EPZ by the U.S. Coast Guard and Missouri Pacific Railroad, et al.,
respectively, as previously committed to by the applicant in correspondence dated October 7, 1981.
This matter is confirmatory in nature and the applicant has agreed to incorporate updated agreement letters or to make appropriate 4
reference to State and local plans containing such letters.
Item 5.
Evacuation time estimates must satisfy the criteria of NUREG-0654.
Discussion Supplement 1 to the SER identified the Waterford 3 evacuation time estimate study as an unresolved item.
The applicant provided the additional information as Revision 1 to Appendix B of the Plan, Evacuation Time Estimate, dated February 1982. The staff has reviewed the revised evacuation time estimate study and found the methodology acceptable.
Based on our review of their Plan as discussed above, the staff finds that the applicant has provided an acceptable response to this item.
13.3.2.4 Alert and Notification System The alert and notification system was initially discussed in Section 13.3.2.5 of the SER. The applicant submitted on August 10, 1981, its conceptual design for the alert and notification system for the 10-mile EPZ.
The submittal contained a schedule for system delivery on March 1, 1982, and for testing during the period June 1-30, 1982.
I On February 24, 1982, the applicant submitted a report to the NRC entitled,
" Verification of the Siren Alert System for Waterford 3 Nuclear Power Station" (February 1982).
A copy of this report was also sent to FEMA Region VI for review and comment. The applicant's letter to FEMA contained a proposed system completion date of May 1982.
The subject report, prepared by Acoustic i
Technology, Inc. (ATI) under contract from the applicant, provides a general description of the " composite" warning system.
In the report, ATI identified a potential problem that may exist in industrial activities where high noise areas may require additional alert coverage.
There are 30 industrial plants Waterford SSER 3 13-3
identified b.'
jeicant in the 10-mile EPZ, and it is not clear which of these, if any affected by the potential problem indicated above.
In a letter doced April 7, 1982, the applicant committed to revising Section 7.5.4 of the Emergency Plan to describe and commit to an Alert and Notification System which meets the guidance criteria of Appendix 3 to NUREG-0654.
The adequacy of this system will be evaluated during actual system test, probably during joint exercises of the facility with offsite authorities and will be upgraded if necessary in response to any identified deficiencies.
13.3.3 Review of State and Local Plans by FEMA The State of Louisiana Peacetime Radiological Response Plan (Annex J, Appendix 7 to Louisiana Preparedness Plan for Emergency Operations),
Revision 3, dated September,1981, and Attachment 1, Parish Plans for Waterford 3, have been rev owed by FEMA.
Interim findings by FEMA as to the adequacy of State and local plans are provided in Appendix F to this supplement.
Subsequent to the review of State and local plans described above, the State of Louisiana revised their Plan and submitted Revision 4 dated February 1982.
Following a review of Revision 4 to the State Plan and conduct of a joint exercise, a supplement to this report will provide final findings as to the adequacy of State and local plans.
It should be noted that the Waterford 3 joint exercise is scheduled for August 25, 1982, and not April 26, 1982 as indicated by FEMA's memorandum.
13.3.4 Conclusions Based on our review against Regulatory Guide 1.101 (Rev. 2), which endorses (by reference) the criteria in " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," NUREG-0654/ FEMA-REP-1, Revision 1, November, 1980, the staff concludes that upon satisfactory completion of those items requiring resolution identified in Section 13.3.2 and those items committed to by the applicant, the Waterford 3 Steam Electric Station Emergency Plan will provide an adequate planning basis for an acceptable state of emergency preparedness and will meet the requirements of 10 CFR Part 50, and Appendix E thereto.
After receiving the additional findings and determinations made by FEMA on State and local emergency response plans, and after reviewing the revision (s) to the applicant's Plan, a supplement to this report will provide the staff's overall conclusion on the status of emergency preparedness for Waterford 3 and related emergency planning zones.
Waterford SSER 3 13-4
15 ACCIDENT ANALYSIS 15.2 Normal Operation and Anticipated Transients 15.2.3 Decrease in Reactor Coolant Flow Rate 15.2.3.2 Single Reactor Coolant Pump Sheared Shaft In Section 15.2.3.2 of the SER the staff stated that the analysis of this event has not been provided in the Waterford 3 FSAR.
However, the applicant has committed to provide a safety grade protection system for this postulated event and to submit an analysis of this event.
In Amendment 25 to the FSAR, the applicant provided the results of its analysis for the subject accident.
The rapid reduction in reactor coolant flow caused by the RCP sheared shaft results in an increase in core average coolant tem-perature, a corresponding reduction in the margin to DNB, and an increase in RCS pressure.
Due to the rapidly decreasing flow, a trip on low steam genera-tor AP is generated at 1.26 seconds.
The CEAs begin to drop into the core at 1.76 seconds. Also, at 1.76 seconds the turbine / generator is tripped which is assumed to result in a loss of offsite power and subsequent coastdown of the remaining three RCPs.
The event results in a transient minimum DNBR of 0.747 at 3.4 seconds of the transient.
The percentage of fuel pins which are calcu-lated to experience DNB is 7.5 parcent.
The calculational method presented in CENPD-183 (this topical report has been approved by the staff) was used to cal-culate the fuel pins which experience DNB.
For the purpose of radiological release calculations, all fuel rods that experience DNB are assumed to fail.
The results of the analysis showed a 2-hour thyroid dose at the exclusion area boundary is less than 30.0 rem. The maximum RCS pressure is less than 110 percent of the RCS design pressure.
The staff has evaluated the applicant's analysis and has concluded that these results are acceptable because they meet the staff acceptance criteria for this event.
For the above event combination plus a postulated single active failure of a safety-related component, the staff acceptance criteria is that the offsite doses at the exclusion boundary should be within the 10 CFR Part 100 guideline values.
The applicant has not provided the results of an analysis for this event with a single failure.
However, the staff could not identify any credible single active failure of a safety-related component which leads to a more severe result.
Based on the above, the staff has concluded that the results of the applicant's analysis for a single RCP sheared shaft accident meet the staff acceptance cri-teria and, therefore, are acceptable.
Waterford SSER 3 15-1
l l
l 18 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS On January 13, 1982 the staff issued Supplement 2 to the Waterford SER which included our evaluation of LP&L's management organization, staffing, and training.
The ACRS, during its 263rd meeting on March 4-6, 1982, reviewed these subjects and issued a report dated March 9, 1982. A copy of this report is included in Appendix C to this SSER.
1 i
Waterford SSER 3 18-1
l APPENDIX A Continuation of Chronology of Radiological Review December 17, 1981 Letter from applicant transmitting " Response to NUREG-0588 -
Introduction, Methodology and Master Equipment List,"
Revision 0 and " Response to NUREG-0588 - Qualification Data Evaluation Forms," Vol. 1 and 2 December 18, 1981 Letter from applicant providing monthly staffing report December 18, 1981 Letter from applicant requesting approval of location of Emergency Operations Facility December 22, 1981 Generic Letter 81 Qualifications of Reactor Operators - License Examinations December 22, 1981 Meeting with applicant to hear its plans for Emergency Operations Facility December 31, 1981 Letter from appicant concerning preparation of plant procedures December 31, 1981 Letter from applicant forwarding " Evaluation of Pressurized Shock Effects Due to Small Break LOCAs With Loss of Feedwater for C-E NSSS," CEN-189 January 6, 1982 Li^ter from applicant requesting approval of location of Emergency Operations Facility, proposed as a result of December 22 meeting January 7, 1982 Meeting with applicant to hear its plans for Emergency Operations Facility January 13, 1982 Issuance of Supplement No. 2 to Safety Evaluation Report January 19, 1982 Letter from applicant requesting extension of effective date for compliance with rule on protection of unclassi-fled safeguards information January 19, 1982 Letter to applicant transmitting request for additional information January 21, 1982 Meeting with applicant to hear applicant's presentation of information related to site hazards January 21, 1982 Meeting with applicant to hear applicant's presentation on risk of turbine missiles Waterford SSER 3 A-1
January 21, 1982 Submittal of Amendment No. 25, consisting of information on reactor coolant pump shaft break analysis, sizing of
. primary safety valves, in adequate core cooling instrumenta-tion and miscellaneous information January 25, 1982 Letter from applicant transmitting submittal schedule for_
open items January 25, 1982 Letter from applicant transmitting monthly report on staffing January 26, 1982 Letter from applicant transmitting " Method of Coolability Analysis for Deformed Grids in Peripheral Assemblies,"
proprietary and nonproprietary versions
. January 26, 1982 Letter from applicant transmitting " Response to Question on Effect of Spacer Grid Design on DNBR Prediction," proprietary and nonproprietary versions February 3, 1982 Letter from applicant concerning hydrogen lines in safety-related areas February 4, 1982 Letter to applicant advising that Emergency Operations Facility proposed January 6 is acceptable February 8, 1982 Generic Letter 82 Nuclear Power Plant Staff Working Hours February 10, 1982 Meeting with applicant to discuss flood protection of ultimate heat sink February 16, 1982 Letter from applicant announcing delay in fuel loading until January 16, 1982 February 79, 1982 Letter to applicant encouraging it to join CE Owners Group investigation of feed and bleed, and forwarding related information.
February 24, 1982 Letter from applicant transmitting " Verification of the Siren Alert System for Waterford 3 Nuclear Power Station" February 25, 1982 Submittal of Amendment No. 26, including information on turbine missiles and miscellaneous changes 4
j February-26, 1982 Meeting with applicant to discuss its submittal on environ-j mental qualificat;on of electrical equipment l
March 2, 1982 Board Notification 82 Information Items on Pressurized
[
Thermal Shock and Feed and Bleed Capability l
March 3, 1982 ACRS Subcommittee meeting with staff and applicant March 4.-1982 ACRS meeting with staff and applicant Waterford SSER 3 A-2
l' l
March 9, 1982 Letter from Advisory Committee on Reactor Safeguards March 16, 1982 Meeting with applicant to discuss preservice inspection /
inservice inspection program March 19, 1982 Letter to applicant transmitting Branch Technical Position regarding shutdown capability from a location outside the main control room.
March 26, 1982 Letter to applicant concerning Human Factors Engineering Branch control room review.
March 27, 1982 Letter to applicant concerning depressurization and decay heat removal.
l i
I i
l Waterford SSER 3 A-3
APPENDIX B Bibliography
- USNRC REGULATORY GUIDES 1.95 Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release 1.99 Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials COMBUSTION ENGINEERING REPORT CENPD-183, "C-E Methods for Loss of Flow Analysis" LETTERS A. E. Scherer (CE) to H. R. Denton (NRC), " Reactor Power Cutback System Review Schedule," October 30, 1981.
CE Letter, C-CE-7459, Enclosure 1, " Response to Questions on Effect of Spacer Grid Design on DNBR Prediction," January 1982 D. G. Eisenhut (NRC) to A. E. Scherer (CE), February 8,1982.
A. E. Scherer (CE) to R. L. Tedesco (NRC), March 4,1982.
R. L. Tedesco (NRC) to A. E. Scherer (CE), March 26, 1982.
R. L. Tedesco (NRC) to L. V. Maurin (LP&L), March 27, 1982.
L. V. Maurin (LP&L) to R. L. Tedesco, April 7, 1982.
USNRC REPORTS NUREG-0635, " Generic Assessment of Small-Break Loss-of-Coolant Accidents in Combustion Engineering Designed Operating Plants," January 1980.
NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants--LWR Edition, " July 1981.
(Formerly NUREG-75/087)
- Available for inspection and copying for a fee in the NRC Public Document Room, 1717 H St., NW., Washington, D.C.
Waterford SSER 3 B-1
i APPENDIX C
/
o,,
UNITED STATES NUCLEAR REGULATORY COMMISSION o
,I ADVISORY COMhMTTEE ON REACTOR SAFEGUARDS o
WASHINGTON, D. C. 20555 March 9, 1982 Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
REPORT ON THE WATERFORD STEAM ELECTRIC STATION UNIT 3
Dear Dr. Palladino:
During its 263rd meeting, March 4-6, 1982, the Advisory Committee on Reactor Safeguards continued its review of the application of Louisiana Power and Light Company (Applicant) for a license to operate the Water-ford Steam Electric Station Unit 3 (Waterford-3).
This project was considered at a Subcommittee meeting on March 3, 1982 in Washington, D.C. and at a previous full Committee meeting on August 6-8, 1981.
During the August meeting, the Committee prepared an interim report to you dated August 11, 1981.
In its review the Committee had the benefit of discussions with the Applicant and the NRC Staff.
The Committee also had the benefit of the documents listed.
In its interim report the Committee expressed concern about the organiza-tional readiness of the Applicant to operate the plant and about the adequacy of the Applicant's training program.
The report made several specific suggestions, and we indicated that we would report to you further on the adequacy of staffing and management.
During the meetings on March 3 and 4,1982, the NRC Staff reported its conclusion that the Applicant's organization, staff, and management will i
be adequate tn operate Waterford-3 in a safe manner by the time of fuel loading, currently scheduled for January 1983.
The Applicant described i
i efforts over the past six months to strengthen the Waterford-3 organiza-tion and training program.
These efforts include important changes in the corporate structure to provide increased dedication of management to the task of completing and operating Waterford-3, changes in the operat-ing organization to pemit improved focus on direct operational and technical support functions, substantial progress toward completion of staffing, the fomation of a comprehensive training program, and estab-1 lishment of a strong Safety Review Committee.
In addition, the Applicant described the integration of the Waterford-3 and contract personnel into an effective startup organization.
I C-1
Honorable Nunzio J. Palladino March 9, 1982 The Committee believes that the Applicant has effectively responded to the concerns regarding organization and management expressed in our August 11, 1981 report.
We believe that with continued dedication of Louisiana Power and Light Company management, satisfactory completion of staffing and the planned program for training, and due consideration to other matters noted in our August 11, 1981 report, there is reasonable assurance the Waterford Steam Electric Station Unit 3 can be operated without undue risk to the health and safety of the public.
Sincerely,
\\.
P. Shewmon Chairman References 1.
Louisiana Power and Light Company, "Waterford Steam Electric Station Unit No.
3, Final Safety Analysis Report," with Amendment's 1-25.
2.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Operation of Waterford Steam Electric Station, Unit No. 3,"
NUREG-0787, dated July 1981 with Supplement 1, dated October 1981 and Supplement 2, dated January 1982.
C-2
44 lo UNITED STATES g
NUCLEAR REGULATORY COMMISSION n
,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS t,
/g WASHINGTON, 0. C. 20555 December 15, 1981 Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
ACRS REPORT ON FINAL DESIGN APPROVAL FOR CGiBUSTION ENGINEERING, INC.
STANDARD NUCLEAR STEAM SUPPLY SYSTEM (STANDARD REFERENCE SYSTEM 80)
Dear Dr. Palladino:
During its 260th meeting, December 10-12, 1981, the Advisory Committee on Reactor Safeguards reviewed the application of Combustion Engineering, Inc.
for final design approval for its Standard Reference System 80 described in CESSAR.
A Subcommittee meeting was held with representatives of the Appli-cant and the NRC Staff in Windsor, Connecticut on November 19, 1981.
The Committee also had the benefit of the documents listed.
The Committee's report on the preliminary design approval for this standard nuclear steam supply system (NSSS) was provided in a letter to the NRC Chaiman dated September 17, 1975.
The System 80 design consists of a reactor system with a design rated core output of 3800 MWt and includes the reactor coolant system, reactor protec-tion system, engineered safety features actuation system, chemical and vol-ume control system, shutdown cooling system, safety injection system, and fuel handling system. The System 80 design provides safety-related inter-face requirements information essential to the design of the balance of pl ant, Combustion Engineering provides, at the option of the user, certain other nonstandard safety-related systems and services which are outside the scope of the System 80 design.
Such systems will need to be dealt with in each user's Safety Analysis Report. The regulations governing the review of standard plant designs under the " reference system" option described in the Federal Register (42 FR 34395 and 43 FR 38954) are contained in para-graph 2.110 of 10 CFR Part 2 and Appendix 0 to 10 CFR Part 50.
CESSAR provides infomation required to ensure that the balance of plant is designed to protect the System 80 from site-related hazards. It envelops all plant sites approved to date for Combustion Engineering nuclear steam supply systems.
When the System 80 design is applied, the related site must be evaluated to establish its acceptability within the System 80 envelope.
For multiple reactor units at a single site, the reference design requires that each important safety-related item be separately provided for each.
reactor unit.
The first plant using the System 80 design will be Palo Verde Nuclear Generating Station, Units 1, 2, and 3, of which Unit 1 is scheduled to load fuel during November 1982.
C-3
Honorable Nunzio J. Palladino December 15, 1981 Because the utility-applicant is responsible for instituting the quality assurance programs necessary to assure that all safety-related requirements have been met, the NRC must review these matters with the utility-applicants on a case-by-case basis. The ACRS believes that Combustion Engineering should be required to evaluate the adequacy of the implementation of inter-face requirements, including such items as the influence of plant control system performance and reliability on NSSS integrity and function.
In recent years, the availability of reliable shutdown heat removal capa-l bility for a wide range of transients has been recognized to be of great importance to safety. The System 80 design does not include capability for rapid, direct d6 pressurization of the primary system or for any method of heat removal immediately after shutdown which does not require use of the steam generators.
In the present design, the steam generators must be op-erated for heat removal after shutdown when,the primary system is at high pressure and ter.perature. This places extra importance on the reliability of the auxiliary feedwater system used in connection with System 80 steam generators and extra requirements on the integrity of the steam generators.
The ACRS believes that special attention should be given to these matters in connection with any plant employing the System 80 design.
The Committee also believes that it may be useful to give consideration to the potential for adding valves of a size to facilitate rapid depressurization of the System 80 primary coolant system to allow more direct methods of decay heat removal. The Committee wishes to review this matter further witit the i
cooperation of Combustion Engineering and the NRC Staff.
System 80 employs some new design features for the steam generators, the core outlet flow region, control rod guidance and shrouding, and the core support structure.
These appear to be acceptable, but, because they are new features, they should be monitored during early operation to determine if they perfom as expected.
l A number of items have been identified as Outstanding Issues and Confirma-tory Issues.
T' se include some TMI-2 Action Plan requirements.
Progress on these matters is satisfactory, and we believe these issues can be re-solved in an acceptable manner.
The Committee wishes to be kept infomed.
The manner of applying preliminary and final design approvals of the type proposed for System 80 will not be completely defined until System 80 has l
been used for several licensing actions at both the construction permit and operating license stages. The Committee believes that standard designs'such as System 80 can be useful in assuring acceptably safe plants. However, a 1
policy to establish when and how changes will be permitted to new or pre-viously licensed plants is needed.
C-4
Honorable Nunzio J. Palladino December 15, 1981 The Committee believes that, subject to the above comments and approval of the balance-of-plant designs, the System 80 design can be incorporated into nuclear power plants that can be operated without undue risk to the health and safety of the public.
Sincerely, J. Carson Mark Chaiman
References:
1.
Combustion Engineering, Inc., " System 80 CESSAR FSAR," with Amendments 1 through 5.
2.
U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Re-lated, to the Final Design of'the Standard Nuclear Steam Supply Reference System CESSAR SysteC80," NUREG-0852, dated November 1981.
C-5
o UNITED STATES g
/
NUCLEAR REGULATORY COMMISSION n
,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 8,
$g April 5, 1982 Mr. Willian J. Dircks Executive Director for Operations U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Dircks:
Subject:
RELIABILITY OF THE SHUTDOWN HEAT REMOVAL SYSTEM ON THE SYSTEM 80 DESIGN The ACRS in its December 15, 1981 report to Chairman Palladino on the Combustion Engineering, Inc. Standard Reference System 80 commented on the reliability of the decay heat removal system. These comments addressed the lack of a capability for rapid, direct depr'essurization of the primary sys-tem to allow feed and bleed operations and the reliance placed upon the secondary system for heat removal capability.
The ACRS Subcommittee on Decay Heat Removal Systems met with representatives of Combustion Engineer-ing, Inc. and the NRC Staff on March 16, 1982 to discuss these issues.
The ACRS discussed these issues further during its 264th meeting, April 1-2, 1982.
Representatives of Combustion Engineering have defended their design, stat-ing that:
1.
The System 80 NSSS will be coupled with highly reliable emergency feedwater systems (EFWS) by addition of an interface reguirement to 10-5 4
that the EFWS have an unavailability in the range of 10 per demand.
2.
The Systen 80 NSSS is capable of achieving cold shutdown conditions using only safety grade systems even without offsite power and with an added single failure.
3.
The System 80 steam generator design includes many features that will assure adequate tube integrity, minimizing concerns associated with operating reactors.
4.
Even if all auxiliary feedwater supply were somehow lost, the secon-dary side of the steam generators could be depressurized to allow use of low head pum:; which might be aligned to provide water to the steam generators frvm a number of sources.
5.
Probabilistic analyses have not shown that installing PORVs will result in a significant improvement in safety. The added costs are not justified.
C-6
l j
William J. Dircks April 5, 1982 i
Combustion Engineering has proposed that the issues associated with the Committee's comments on the System 80 design be resolved in a continu-1 ing dialogue among the ACRS, the NRC Staff, and Combustion Engineering.
It is the NRC Staff's intention to address these issues on an expedi-tious schedule with all applicants requesting licenses for Combustion Engineering NSSS designs which do not have capability for rapid depres-surization independent of the steam generator. We concur with this ap-proach and wish to be kept informed.
The Combustion Engineering response to the Comittee's comments on the System 80 design emphasizes the expected very high reliability of the feedwater systems and the integrity of the steam generators. We believe that these are necessary goals but note that past operating experience indicates that these goals are difficult to achieve.
We believe that for this reason Combustion Engineering and the.NRC Staff should consider fur-ther the addition of valves of a size to facilitate rapid depressuriza-tion of the System 80 primary coolant system as stated in the Committee's December 15, 1981 letter on the System 80 design.
We believe that a plan for addressing this issue should be fomulated in the near future. We wish to be kept informed and to discuss this further with Combustion Engineering and the NRC Staff.
We believe that, while this evaluation should be conducted expeditiously, i
its resolution should not now be a condition for operation of System 80 plants at full power, or of plants having similar features. The need for future hardware or procedural changes should be contingent upon results of this evaluation.
Sincerely,
\\.
P. Shewmon Chaiman C-7
APPENDIX D ERRATA TO SAFETY EVALUATION REPORT Section 1.11, Page 1-10 Last paragraph in Section (2) should read exhaust cycle instead of recirculation mode.
Section 2.3.2, Page 2-14 Predominant onsite wind direction should read south-southeast, rather than southeast.
Section 2.5.2, Page 2-26 Second line, evaulated should read evaluated.
Section 2.5.2.3, Page 2-27 Second line, Wateford should read Waterford.
Section 2.5.4.1, Page 2-30 Subsection (3), 64 locations should read 74 locations.
Section 2.4.2.2, Page 2-17 Second paragraph, 27 ft MSL should read 24 ft MSL.
Section 2.4.3, Page 2-19 Water withdrawal rate of 1,003,274 gal / day should be gal / minute.
Section 4.4.1, Page 4-16 Paragraph 1, line 6, 4 x 4 and 5 x 5 should read 5 x 5.
Paragraph 3, line 12, th should read the.
Section 6.2.2, Page 6-12 Last paragraph, GDC 38, 29, 40 and 50 should read G0C 38, 39, 40 and 50.
5,ection 6.2.4, Page 6-14 Second paragraph, delete the end of the last sentence after the word radiation.
Waterford SSER 3 0-1
3 Section 6.2.5, Page 6-17 Last paragraph, next to last sentence, delete the end of the sentence after the phrase zinc paint corrosion rates.
Section 8.4.6, Page 8-18 Last paragraph, 120-V dc power should read 125-V dc power.
Section 9.5.2.2, Page 9-32 Last line of this section, lighting should be communication.
Section 22.2, Page 22-59 Next to last paragraph, change 0-200 psia to -5 to 195 psig in two places.
Appendix H, Page H-1 Principal Contributiors should read Principal Contributors.
Supplement No. 2 Section 1.8, page 1-2, confirmatory item number 8 should be deleted.
Waterford SSER 3 D-2
j 1
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- s.,
4 APPENDIX E m
Principal Contributors to SSER No. 3 P. Sears Chemical Engineering - Fire Protection J. Wermeil Auxiliary Systems-
.s C. Liang Reactor Systems' N
B. Elliot Materials Engineering -
J. Holonich C
rformance 3
1
,antatJon_and Control R. Stevens R. Barnes u s-Babcock, Inc.
- Fires?rotection D.'Perrotti Emergency Planning s R. Giardina Power Systems l
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