ML25233A084
| ML25233A084 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 12/23/2025 |
| From: | US Atomic Energy Commission (AEC) |
| To: | |
| References | |
| Download: ML25233A084 (1) | |
Text
{{#Wiki_filter:SUPPLEMENT NO. 2 TO THE SAFETY EVALUATION OF THE BRUNSWICK STEAM ELECTRIC PLANT UNITS 1 AND 2 DOCKET NOS. 50-324 & 50-325 ......... ::\\~tllGY c~:::.... *.********************************** U.S. ATOMIC ENERGY COMMISSION DIRECTORATE OF LICENSING WASHINGTON, D. C. ISSUE DATE: DECEMBER 23, 1974
SUPPLEMENT NO. 2 TO THE SAFETY EVALUATION OF THE BRUNSWICK STEAM ELECTRIC PLANT UNITS 1 AND 2 DOCKET NOS: 50-324 50-325 U. S. ATOMIC ENERGY COMMISSION DIRECTORATE OF LICENSING WASHINGTON, D. C. Issue Date: December 23, 1974
TABLE OF CONTENTS INTRODUCTION PART A - UPDATING OF THE SER 5.2.5 7.3.4 Overpressure Protection Low Pressure Coolant Injection (LPCI) System PART B - UPDATING OF SUPPLEMENT NO. l PART A, Item 5 PART B, Item 6 Appendix A MSLIV Position Switch Environmental Qualification Rod Sequence Control System (RSCS) Chronology 2 3 10 19 19 19 20 INTRODUCTION Since the issuance of the Regulatory staff's Safety Evaluation Report (SER), dated November 1973, and its Supplement No. 1, dated January 31, 1974, for the Brunswick Steam Electric Plant (BSEP), Units l and 2, Amendments 26, 27, 28 and 29 to the Final Safety Analysis Report (FSAR) were submitted by the applicant, the Carolina Power and Light Company. These four amendments to the FSAR have been reviewed by the Regulatory staff and we have determined that the changes described in these amendments do not alter the findings and conclusions we made in the SER and its Supplement No. l. During construction of the Brunswick Plant, as reported in the above cited amendments, the applicant made modifications to two safety related systems. Part A of this supplement provides a description of these modifications and a summary of our review and conclusions regarding their acceptability. In adnition, the applicant furnished information relating to Item 5 of Part A and Item 6 of Part Bin Supplement No. 1 to the SER. Our evaluation of this information is provided in Part B of this supplement. In a letter dated July 1, 1974, to the Directorate of Licensing, the Carolina Power and Light Company submitted a revised Overpressure Protection Report to support a change in the overpressure protection system. Section 5.2.5 Overpressure Protection of Part A provides the safety evaluation which supercedes that evaluation given in the SER and describes the modifications made to the BSEP overpressure protection system since issuance of the SER and its Supplement No. 1. On August 5, 1974, the applicant submitted a modification proposed for the low pressure coolant injection system {LPCIS) of the BSEP, Units l and 2. The modifications to the LPCIS are described and our evaluation of the modified LPCIS are provided in Section 6.3.4 of Part A. Section 6.3.4 supercedes the same section in the SER. In Amendment No. 26 to the FSAR, the applicant discussed its environmental qualification testing of certain position switches used with the MSLIV system. Further, our generic review of the rod sequence control system, which is designed to make the design basis control rod drop accident acceptable, has been completed. Both of these matters are addressed in Part B of this supplement. Based on the results of our review as discussed in this supplement and in Supplement No. l, we reaffirm the conclusions given in Section 22.0 of the SER. An updated chronology covering the period since the issuance of Supplement No. l is presented in Appendix A. PART A - UPDATING OF THE SER As indicated in the following Sections 5.2.5 and 6.3.4, we have reviewed the modifications proposed by the applicant and conclude that the changes to the BSEP, Units l and 2 will provide protection equal to or better than that of the previous system designs. On the basis that the previous design was reviewed and found to be acceptable, we conclude that the design incorporating the proposed modifications is also acceptable. The changes described in this report, which are based on the applicant's letters of July l and August 5, 1974, will be further documented by the applicant in an amendment to the Final Safety Analysis Report (FSAR to keep the FSAR current with the BSEP Units l and 2 design. 5.2.5 Overpressure Protection
Background
In its letter dated July l, 1974, the Carolina Power and Light Company submitted a revised Overpressure Protection Report to support a proposed design change in the Brunswick Steam Electric Plant, (BSEP), Units l and 2 overpressure protection system. The proposed change in the overpressure protection system is to replace the two Dresser type safety valves with Target Rock safety/relief valves which will be piped so as to discharge into the torus. In addition, the number of safety/ relief valves for Unit 2 (the first unit to be placed in operation) will be reduced by one valve. The proposed design for the safety/relief valves, eleven valves for each plant, will have a total capacity greater than the original design for Unit l and approximately the same as the original design for Unit 2. The transients resulting from both the turbine trip without bypass and the main steam line isolation valve (MSLIV) closure were considered. The MSLIV closure transient was determined to be the most severe transient. The analysis of this transient assumed a delay time of 0.4 seconds for initiating the valve opening. The proposed overpressure protection system will satisfy the requirements of Article 9, Section III of the ASME Pressure Vessel Code. Evaluation The applicant's evaluation of the sizing of the overpressure protection system was performed using the same methods, transients, and requirements which were previously reviewed and found acceptable. Based on the previously approved and accepted overpressure protection system design the proposed system, which meets the same requirements, is found to be acceptable. Table 5.2.5 provides a comparison of the original and the proposed overpressure protection system for the BSEP, Units 1 and 2. TABLE 5.2.5 OVERPRESSURE PROTECTION SYSTEM COMPARISON A. OPERATING CONDITIONS (SAME FOR UNITS 1 & 2) USED IN ORIGINAL & PROPOSED DESIGN Operating Power Vessel Dome Pressure Steam Flow Design Basis Transient Direct Reactor Scram Vessel Code Pressure Limit 2535 MWt 1020 psi g 10.96 x 106 lbs/hr Main Steam Line Isolation Valve Closure Failed 1375 psig TABLE 5.2.5 (Con'd.) B. FSAR OVERPRESSURE PROTECTION SYSTEM No. Safety/Relief Valves No. Safety Valves Type of Safety/Relief Valves Type of Safety Valve Delay Time (Opening) No. of Safety/Relief Valves 3 3 3* Unit l 9 2 Target Rock Dresser Maxiflow 0.4 seconds Set Pressure ipsig) 1080 1090 1100 Unit 2 10 2 Target Rock Dresser Maxiflow 0.4 seconcfs ASME Rated Capacity@ 103% of set pressure (lbs/hr) per valve 788,000 796,000 803,000
- Unit 2 has four safety/relief valves at this set pressure No. of Safety Valves 2
Number of Safety/Relief Valves Required for MSLIV 1240 Closure Transient with Flux 7 Scram (No Safety Valves Open) Peak Pressure from Transient at bottom of Pressure Vessel 1336 psig Margin to ASME Pressure Limit (1375 psig) 39 psi Margin to ASME Pressure Limit (1375 psig) for Peak Pressure Transient for MSLIV Closure with High Pressure Scram and One Valve Failed 25 psi 642,000 7 1336 psig 39 psi 25 psi TABLE 5.2.5 (Con'd.) C. PROPOSED OVERPRESSURE PROTECTION SYSTEM Unit 1 No. Safety/Relief Valves 11 No. Safety Valves 0 Type of Safety/Relief Valves Target Rock Delay Time (Opening) 0.4 seconds No. of Safety/Relief Set Pressure Valves (psig) 4 1080 4 1090 3 1100 Number of Safety/Relief Valves Required for MSLIV Closure Transient with 5 Flux Scram Peak Pressure from Transient 1345 psig Margin to ASME Pressure Limit (1375) 30 psi Margin to ASME Pressure Limit (1375 psig) for Peak Pressure Transient for MSLIV Closure with High Pressure Scram and One Valve Failed 95 psi Unit 2 11 0 Target Rock 0.4 seconds ASME Rated Capacity@ 103% of set pressure (lbs/hr) per valve 830,000 838,000 845,000 5 1345 psig 30 psi 95 psi The proposed overpressure protection system does not change the automatic depressurization system (ADS) described in Section 6.3.2 of the FSAR. The transient analysis performed in accordance with the requirements of the ASME Boiler and Pressure Vessel Code, Section III, 1968 using the proposed overpressure protection design characteristics satisfies the maximum pressure limit of 1375 psig with a margin 30 psi, with only 5 of the 11 safety/relief valves operable. The replacement of the two safety valves, which were not piped to the torus, with safety/relief valves which discharge to the torus, elimi-nates the possible discharge of steam into the drywell during any inadvertent discharging of the safety valves. The applicant has determined that the increased size and capacity of the eleven safety/relief valves will not require any change in the size or means of support of the discharge piping to the torus. The peak pressure resulting from the MSLIV closure transient and a high flux scram (indirect scram) is 40 psi lower for the proposed system than for the original system for the seven safety/relief valves and zero safety valve case. For the same MSLIV closure transient with only a high pressure scram (indirect scram) the proposed system will provide an additional margin of 70 psi over that provided by the original design. Conclusion Based on our review of the Summary Technical Report of Reactor Vessel Overpressure Protection, dated May 1974, and submitted to the Directorate of Licensing by letter on July 5, 1974, we find that the proposed change in the overpressure protection system for the BSEP, Units 1 and 2 satisfies the requirements of the Article 9, Section III of the ASME Pressure Vessel Code, in the area of vessel overpressure protection design of nuclear vessels and we conclude that the proposed design is acceptable. The applicant plans to amend the FSAR to include the description and analyses for the proposed change. 6.3.4 Low Pressure Coolant Injection (LPCI) System
Background
In its August 5, 1974, submittal of the analyses and ctata required by Appendix K of 10 CFR Part 50, the applicant proposed a design chanqe to the mode of operation of the low pressure coolant injection system (LPCIS) from that design which was reviewed and accepted in the Regulatory staff's SER and Supplement No. 1. The applicant has provirled the analyses and data to support the proposed change in the LPCIS. We have reviewed this information as it is affected by the proposed modification. Based upon a comparison of the proposed design change with the original design of the LPCIS, the proposed design will be equal to or better than the original design when judged by all the criteria and performance require-ments used to determine that the original design was acceptable. Whether the proposed design satisfies the criteria and functional requirements of Appendix K of 10 CFR Part 50 is currently under review and the results of this review will be reported in a later supplement to the SER. The proposed design changes to the LPCIS will be documented in the FSAR in an amendment. Evaluation The applicant has proposed design changes to the LPCIS mode of operation which eliminates the loop selection logic required to detect the broken loop involved in a circulation pipe rupture accident. The proposed LPCIS will use components of the original design and will meet all the criteria and functional requirements used to conclude that the original LPCIS was acceptable. Figure l provides a line diagram of the original LPCIS and Figures 2 anct 3 show the proposed modified LPCIS for a break in the suction and discharge side of the recirculation loop, respectively. Table 6.3.4 was prepared by the applicant to show the comparison of the original and modified LPCIS design. Using the model and methods described in Appendix K of 10 CFR Part 50 the applicant has calculated that the peak clad temperature will be reduced by 210°F for the design basis accident, a recirculation suction line break, using the modified LPCIS and assuming that the worst single failure occurs. The model and methods used to satisfy the Interim Acceptance Criteria results in a reduction of approximately 150°F in the peak clad temperature for the same design basis accident, using the modified LPCIS and assuming the worst single failure.
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-L\\(_ DG #2 'if? ])/SCI-IP.RO~ A .S/flJTCff' I r-,, ___ J 1/U'{,:r,~;.'J/IM -**-t=J f I GU R £_ __ ]:_- /10_}) If} Ej) ___ SYS T £ 11 t/0JJE OF OPERIJTION JJLIR/NG SUCTION LIN£ BRERJ<. Worst Single Failure: Diesel Generator Fai111rn or Ll>CI Injection* Valve Fails to Open. OG /13 OG f,3 OG P.4 p Auto logic Auto logic 'Batt 2/\\ C.f.O.SSTIE Batt 2~ 1 ~ FIC-LIRF _ 3 -_1107)/ Fl ED _ SYSTEM OG #4 ~ OG /12 Auto ksi c for /f/J. ])JSC!lf! P.G£ Si/LITOrr /1()])£ IJF OPERRT!ON ])li/?ING-P.!SC:fi/J.f(Qf LIN£ BR£fiK HC'~ST Sl..:CilC FAILURE: LPCl lliJECTlOf~ y,;L\\'1,: lN UNSR.OKCN LOOP FAILS 10. Of'EH OR LOSS OF Pir:SEL GtNDu\\ ~OR ASSOCI/\\TED \\ilTH UNB~OKlfll LOOP TABLE 6.3.4 ECCS PUMP CONFIGURATION (WORST CASE}* SUCTION SIDE BREAK, MODIFIED SYSTEM No Failures Injection Valve Failure Diesel Failure Battery Failure DISCHARGE SIDE BREAK, MODIFIED SYSTEM No Failures Injection Valve Failure Diesel Failure Battery Failure SUCTION SIDE BREAK, UNMODIFIED SYSTEM No Failures Diesel Failure Injection Valve Failure Battery Failure DISCHARGE SIDE BREAK, UNMODIFIED SYSTEM No Failures Diesel Failure Injection Valve Failure Battery Failure PUMPS AVAILABLE 2 Core Spray+ 4 LPCI's; 2 In Each Loop 2 eore Spray+ 2 LPCI's In One Loop 2 Core Spray+ 2 LPCI's In One Loop 2 Core Spray+ 2 LPCI's In One Loop PUMPS AVAILABLE 2 Core Spray+ 2 LPCI's In One Loop 2 Core Spray 2 Core Spray 2 Core Spray PUMPS AVAILABLE 2 Core Spray+ 4 LPCI's l Core Spray+ 3 LPCI's 2 Core Spray l Core Spray: 2 LPCI's PUMPS AVAILABLE 2 Core Spray+ 4 LPCI's l Core Spray+ 3 LPCI's 2 Core Spray l Core Spray+ 2 LPCI's
- Note:
This table presents the least effective pump combinations resulting from the worst single failure. For example: a) for the suction line break 2 CS+ 2 LPCI's in one loop is less effective than 2 CS+ 2 LPCI's; 1 in each loop or 1 CS and 3 LPCI's, b} for the discharge break 2 CS is less efficient than 1 CS and 1 LPCI. Both systems satisfy the criteria and functional requirements upon which we evaluated and found that the original LPCIS was acceptable. The proposed LPCIS will consist of two loops with two residual heat removal (RHR) pumps in each loop. On receipt of a LOCA initiation signal, all four RHR pumps are started and the discharge valves in each of the primary recirculation loops are signaled to close; however, closure of these discharge valves will not start until the reactor vessel pressure is less than 350 psig and the LPCIS injection valves are signaled to open. The design basis accident which is used to establish the ECCS acceptability is the double ended break of the primary coolant recircula-tion pipe with the worst single failure, which is failure of the LPCIS injection valve to open. This design basis accident is the same for both the original and proposed LPCIS. The proposed LPCIS will require only the following modifications:
- 1.
The power to the LPCIS and recirculation discharge valve will be changed from two buses each powered by either of two sources, to two buses supplied by independent sources. This ensures that failure of one bus cannot disable both sets of LPCI and recirculation discharge valves.
- 2.
The recirculation loop selection logic will be eliminated. The accident initiation signals will be rewired to direct both LPCI injection valves to open upon detection of accident conditions. Redundant wiring will be added to provide single failure protection.
- 3.
Both recirculation loop discharge valves (downstream of pumps) will be signaled to close when reactor pressure decreases to less than 350 psig.
- 4.
The control switch for the cross-tie valve between the two LPCIS headers will be changed from key-locked open to key-locked closed and an annunciator will be added to indicate the open position.
- 5.
The DC control logic which provides initiation of the signal to start the RHR pumps will be serviced such that each of two logic channels will start all four RHR pumps. The applicant has indicated that the above modifications will be made in accordance with the applicable codes found acceptable by the Regulatory staff for the original LPCIS design, including the latest approved quality assurance requirements. Our review of the proposed LPCIS design modifications indicates that the functional performance requirements and criteria established for the acceptance and approval of the original LPCIS are satisfied. Conclusion Based on our evaluation of the proposed design modification to the LPCIS for the BSEP Units l and 2, we conclude that design change in the mode of operation satisfies all the requirements and criteria which were used to determine the acceptability of the original LPCIS. We are currently reviewing the LPCIS to determine if any additional requirements on this system are required to satisfy the 10 CFR Part 50 Appendix K criteria; we will issue a supplement to the SER when our review is completed. We will require that power to the operator of the cross-tie valve between the two LPCIS loops be disconnected during normal power operation to assure that the valve remains closed under accident conditions. PART B - UPDATING OF SUPPLEMENT NO. 1 PART A, Item 5: MSLIV Position Switch Environmental Qualification We indicated in Supplement No. 1 to the SER (Item 5 of of Part A) that the applicant had agreed to install environmentally qualified MSLIV position switches on Units 1 and 2. In Amendment 26, on page M.7.21-4, the applicant indicated that two MSLIV position switches of the types installed in Units 1 and 2 were tested in an autoclave at sustained pressures of 103 to 16 psig and temperatures of 340°F to 250°F and successfully functioned throughout these ranges of environmental conditions. Based on these test results we conclude that the MSLIV position switches installed in Units 1 and 2 are acceptable. PART B, Item 6: Rod Sequence Control System (RSCS) Since the issuance of Supplement No. 1 to the SER, we have completed our review of the installation drawings for the RSCS used by the BWR-4 class of reactors and which is installed in BSEP Units 1 and 2. The details of this review, which was perfonned on a generic basis, are contained in the staff document, "Review of Rod Sequence Control Systems With Group Notch (RSCS/BWR-4) 11 attached to an internal memorandum from V. Stello to V. Moore, dated June 6, 1974. Based on our review and evaluation we conclude that the RSCS for Units 1 and 2 are acceptable. Verification of the proper installation of the RSCS for Unit 2 will be made by the preoperation system test P0-15, which will be completed prior to the initial fuel loading of Unit 2.
February 4, 1974 February 13, 1974 February 19, 1974 February 21, 1974 March 4, 1974 March 4, 1974 March 4, 1974 March 12, 1974 March 27, 1974 APPENDIX A CHRONOlOGY Reaulatory Review of Carolina Power an Light Company, Brunswick Steam Electric Plant, Units l & 2 Since January 1974 CP&L submits "Evaluation of Alternative Cooling System". CP&L letter referring to AEC/DL letter, dated January 22, 1974 concerning Quality Assurance. AEC/DL letter concerning new regulations regarding byproduct, source, and special nuclear materials for the operating license of Brunswick l & 2. CP&L submits their comments on the Final Environmental Statement. Notice of Reconstitution of Board issued by AS&LBP. Because of schedule conflicts Mr. Glenn 0. Bright is unable to continue his duties as a member of the Board of Brunswick. Mr. Gustave A. Linenberger, AS&LBP is appointed a member of the Board of Brunswick, replacing Mr. Bright. CP&L letter transmitting Addendum B to Design Report No. 15. CP&L requests that this transmittal be withheld from public disclosure as proprietary infonnation. Design Report 15 and Addendum A to that report were withheld from public disclosure. CP&L letter submits the details of the test program and CP&L's schedule for completion and replacement of any non-qualified switches of the BSEP MSLIV. Notice and Order for Prehearing Conference - issued by the AS&LB. A further prehearing conference will be held at 10:00 a.m. on Friday, March 22, 1974 at the United States Tax Court, Courtroom No. l, Room 2132, 1111 Constitution Avenue, N. W., Washington, D. C. 20044. Memo to R. C. DeYoung, A/D for LWR Group l, Thr.u: W. R. Butler, BC/LWR 1-2 from R. Powell, Project Manager for the Brunswick Electric Plant. This memo justifies the withholding of Addendum B to Design Report No. 15 from public disclosure.
March 27, 1974 Apri 1 2, 1974 April 17, 1974 April 17, 1974 May 10, 1974 May 22, 1974 May 23, 1974 May 31, 1974 June 3, 1974 June 18, 1974 June 28, 1974 CP&L BRUN'SWICK AEC/DL letter advising that Addendum B to nesign Report No. 15 has been withheld from public dis-closure as proprietary information pursuant to Section 2.790(b) of 10 CFR Part 2 in accordance with CP&L's letter, dated March 4, 1974, requesting this withholding of information from public disclosure. CP&L transmits additional information regarding a standard licensing provision for by-product, source and special nuclear materials in the operating licenses for Brunswick Units 1 & 2. Notice and Order for Further Prehearing Conference issued by the AS&LB. The prehearing conference will be held at 10:00 a.m. on Tuesday, April 23, 1974, in the U. S. Tax Court, Courtroom No. l, Room 2132, 1111 Constitution Avenue, N. W., Washington, D. C. Sunmary of site meeting held on March 25, 26 and 27, 1974, prepared by R. Powell. AEC/DL letter advising that the review of CP&L's submittal of February 13, 1974 concerning the authority and.organizational freedom of Quality Assurance (QA) perspnnel performing work on the design and construction of the Brunswick Steam Electric Plant has been approved. CP&L advises that the submittal of the revised fuel densification analysis originally due on May 20, 1974, will be delayed. CP&L submits an Operator Retraining Program. CP&L submits change in end-of-cycle scram reactivity curve for reload BWR cores. CP&L submits Application for Extension of Construction Permit for Unit 2 of the Brunswick Station. CP&L amends and supplements information forwarded June 3, 1974, regarding the Brunswick Unit 2 construction extention. AEC/DL letter advising that the construction completion date for the Brunswick Unit 2 has been extended from June 30, 1974, to January 31, 1975.
July l, 1974 July 2, 1974 August 1, 1974 August 1, 1974 August 5, 1974 August 20, 1974 August 23, 1974 September 16, 1974 CP&L BRLJNSWICK CP&L submits Summary Technical Report of the Reactor Overpressure Protection. This report provides sufficient information and documentation to demonstrate compliance with the ASME Pressure Vessel Code. AEC/DL letter advising that the PRT system proposed by CP&L is now under review and has not as yet been found acceptable for any BWR and reco1m1ending that any modifications not be proceeded with until the system has been approved. Order issued by the AS&LB. This order states that an evidentiary hearinq will be held on the construction permits issued for.the Brunswick 1 & 2 plants to determine whether they should be continued, modified, terminated or appropriately conditioned to protect environmental values. Notice and Order for Evidentiary Hearing issued by AS&LB. This hearing will be held on environmental matters on August 19, 1974, at 10:00 a.m., Federal Building, Water & Princess Streets, Wilmington, North Carolina 28401. CP&L submits reports concerning Loss of Coolant Accident & Emergency Core Cooling Systems Low Pressure Coolant Injection System Modifications for Performance Improvement. CP&L letter advising that Unit 2 will be ready for fuel loading on October 15, 1974, and that the AOG system and the CAC system will not be ready or completed but they are not required until completion of the start-up, low power and power range test period. GE letter requesting an extension of installation of decouplers on the recirculation pumps for Brunswick Station. CP&L letter concerning ATWS - results of analyses will be submitted to the AEC by October 1, 1974.
September 17, 1974 September 23, 1974 October 1, 1974 October 11, 1974 October 23, 1974 November 1, 1974 November 1, 1974 November l, 1974 November 5, 1974 November 12, 1974 November 15, 1974 CP&L BRUNSWICK CP&L submits Amendment No. 26 to the FSAR. This amend~ent includes a complete set of revised Technical Specifications, written resolutions to items discussed at a DL site visit in March 1974 and FSAR changes resulting from the modification to the pressure relief system. Notice of Hearing issued by AS&LB. Evidentiary hearing scheduled for September 26, 1974, in the Hearing Room, Landau Bldg., Woodmont Ave., Bethesda. CP&L reply to AEC/DL letter of October 11, 1973, regarding ATWS. CP&L adopts NED0-20626 as their response. CP&L submits revised pages to the Brunswick Steam Electric Plant Industrial Security Plan. CP&L requests that this document be withheld from public disclosure as proprietary. DL & RO representatives meet with CP&L VP and officers in Bethesda, Md. to discuss the Operational QA Program. AEC/DL letter forwarding comments on the revised Technical Specifications submitted by CP&L on October 1, 1974. CP&L submits Amendment No. 27 to the FSAR. This amendment consists of revised pages concerning the QA Program. Summary of Meeting with CP&L to discuss the Brunswick Operational Quality Assurance Program prepared by R. Powell. CP&L letter discussing preoperational testing. Representatives of CP&L & AEC discuss the detailed status of the preoperational tests. This meeting was held in AEC offices, Bethesda, Md. CP&L submits Amendment No. 28 to the FSAR. This amendment consists of revised pages to the FSAR.
November 22, 1974 December 2, l 97 4 CP&L BR~ICK CP&L letter submitted to provide additional details of the Unit 2 preoperational systems test program and schedule. CP&L submits Amendment 29 to the FSAR. This amendment provides operational reference to the Continuing QA Operational Program provided in Section 12 of the PSAR.}}