ML19262A548

Suppl 1 to SER Re Const of Facility

ML19262A548
Person / Time
Site:	05000463, 05000464
Issue date:	06/30/1975
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-75-015, NUREG-75-015-S01, NUREG-75-15, NUREG-75-15-S1, NUDOCS 7911140418
Download: ML19262A548 (78)

Text

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""*"""lj' Philadelphia Electric June 1975 Cornpany Supplement No.1 2203 231 b#

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Available from National Technical Information Service Springfield, Virginia 22161 Pdce: Printed Copy $4.25 ; Microfiche $2.25 2203 232

NUREG-75/015 (Supp.1)

SUPPLEMENT NO. 1 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U. S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF PHILADELPHIA ELECTRIC COMPANY FULTON GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-463 AND 50-464 2203 233

i TABLE OF CONTENTS PAGE

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT.........

1-1 1.1 Introduction....................................

1-1 1.6 Requirements..for Future Technical Information...

1-2 1.8 Summnry of Issues Resulting from Staff Review...

1-2 1.8.1 Resolved Issues.........................

1-3 1.8.2 Outstanding Issues........,.............

1-4 2.0 SITE CHARACTERISTICS..................................

2-1 2.1 Geography and Demography........................

2-1 2.1.2 Site Description........................

2-1 3.0 DESIGN CRITERIA FOR STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS...........................................

3-1 3.9 Mechanical Systems and Components...............

3-1 3.9.1 Dynamic System Analysis and Testing.....

3-1 4.0 REACTOR...............................................

4-1 4.2 Mechanical Design...............................

4-1 4.2.2 Reactor Vessel Internals................

4-1 5.0 REACTOR COOLANT SYSTEM................................

5-1 5.3 Prestressed Concrete Reactor Vessel.............

5-1 5.4 Component and Subsystem Design..................

5-2 5.4.2 Msin Helium Shutoff Valve...............

5-2 5.4.5 Inservice Inspection and Surveillance Programs..............................

5-2 6.0 ENGINEERED SAFETY FEATURES............................

6-1 6.2 Co n t ainmen t Sys t ems.............................

6-1 6.2.1 Containment Functional Design...........

6-1 6.2.1.2 Containment Design Pressure...

6-1 4

dN c

ES 2203 234

11 TABLE OF CONTENTS (CONT'D)

PAGE 6.3 Core Auxiliary Cooling System...................

6-1 6.3.3 CACS Helium Shutoff Valve...............

6-1 7.0 INSTRUMENTATION AND CONTROLS..........................

7-1 7.2 Reactor Trip System.............................. 7-1 7.2.1 Anticipated Transients Without Scram..... 7-1 7.3 Engineered Safety Feature Actuation Systems.....

7-2 7.3.2 Core Auxiliary Cooling System............ '7-2 7.3.2.5 Auxiliary Primary Coolant Shutoff Valve...............

7-2 7.5 Safety-Related Display Instrumentation..........

7-2 7.6 All Other Systems Required for Safety...........

7-3 7.6.2

} bin Loop Shutdown......................

7-3 7.7 Control Systems.................................

7-4 7.7.1 Control Rod System......................

7-4 7.7.6 Reactor Core Instrumentation............

7-6 7.7.7 Turbine-Generator System................

7-8 7.10 Qualification Test Program......................

7-8 9.0 AUXILIARY SYSTEMS....................................

9-1 9.1 Fuel Storage and Handling......................

9-1 9.1.2 Fuel Handling...........................

9-1 13.0 CONDUCT OF OPERATIONS.................................

13-1 13.5 Industrial Security.............................

13-1 14.0 INITIAL TESTS AND OPERATION...........................

14-1 17.0 QUALITY ASSURANCE.....................................

17-1 17.1 General.........................................

17-1 3

2203 235

iii TABLE OF CONTENTS (CONT'D)

PAGE 18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR S AFE GUARDS (ACRS )................................... 18 -1 20.0 FINANCIAL QUALIFICATIONS..............................

20-1 20.1 Introduction....................................

20-1 20.2 Construction Cost Estimates.....................

20-2 20.3 Construction Program and Source of Funds........

20-3 20.3.1 General.................................

20-3 20.3.2 Regulatory Environment..................

20-5 20.3.3 External Financing - Future Plan and Recent History........................

20-6 20.3.4 Internally Generated Funds - Projections and Recent History.................... 20-9 20.4 Conclusion...................................... 20-13 APPENDICES APPENDIX A SUPPLEMENT TO CHRONOLOGY OF RADIOLOGICAL REVIEW APPENDIX B FINANCIAL ANALYSIS APPENDIX C ACRS REPORT APPENDIX D ADDITIONS, LL!ETIONS, AND CORRECTIONS TO THE SAFETY EVALUATION REPORT 2203 236

1-1

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction The Nuclear Regulatory Commission's Safety Evaluation Report (SER) in the matter of the application by Philadelphia riectric Company (PECo) to construct Units 1 and 2 of the Fulton Generating Station was issuet on March 5, 1975.

In that report the staff noted that there were (a) certain staff requirements that would be recommended to the Atomic Safety and Licensing Board as con-struction permit conditions unless PECo made commitments to meet these requirements; and (b) areas where the applicant had not supplied enough information for the staff to complete its review.

The purpose of this supplement is to update the SER by pro-viding the staff's evaluation of additional information received since the issuance of the document and to discuss items recom-mended for consideration by the staff and PECo in the Advisory Committee on Reactor Safeguards (ACRS) report (Appendix C).

In addition, a review of the SER has revealed areas where corrections or further explanations are in order.

Each of the following sections in this supplement is numbered the same as the section of the SER that is being updated.

Appendix A to this supplement is a continuation of the chro-nology of the NRC staff's principal actions with respect to radio-logical matters related to the processing of the application.

Appendix B presents a consolidated financial analysis of PECo.

p Appendix C is the report from the Advisory Committee on Reactor

.i. 3 o(

Safeguards. Appendix D is a listing of additions, deletions and]} }}[ g corrections to the SER.

1-2 1.6 Requirements for Future Technical Information As outlined in the SER, we have identified prototypical items for which additional information is required. The specific information requirements, including information on research, development and testing, are being discussed with the applicant. The applicant has stated that a report on the identified prototypical items will be submitted in July, 1975. The report will be updated every six months to provide information and test results from the research, development and testing programs. We are discussing with the applicant the overall schedule for submission of the test results to assure that the required information is available in sufficient time for us to complete our review prior to installation of the prototypical items. 1.8 Summary of Issues Resulting From Staff Review As indicated in the SER, we have identified (a) resolved issues which were staff positions with which the applicant disagreed and (b) outstanding issues which required additional information for the staff to complete its review. The following two subsections tabulate these items, reference the sections of this report where each matter is discussed, and indicate their status. As a result of discussions with the ACRS Subcommittee and fur-ther discussions with the applicant, changes to the SER have been made in the summary of issues resulting from our review. These changes are noted in Appendix D. 2203 238

1-3 1.8.1 Resolved Issues Status a. Main loop helium shutoff valves and the The applicant's revised core auxiliary cooling loop shutoff design conforms to valves should have safety grade position the staff position. indicators. In addition, the main loop helium shutoff valves should ha e a direct indicator of valve position for all modes of valve operations. (Supplement Sections 5.4.2, 6.3.3, 7.3.2.5 and 7.6.2) b. The containment design pressure should The applicant's revised be 45.5 psig. (Supplement Section design conforms to 6.2.1.2) the staff position. c. The post-accident and incident The applicant's revised monitoring system should have two design conforms to independent channels for monitoring the staff position. each parameter, with one of the channels recorded, and both satisfying the requirements of IEEE Std 279-1971 with Class IE power. (Supplement Section 7.5) d. The slack cable indication system The applicant's revised should satisfy the same design design conforms to criteria requirements provided for the staff position. the control drive positior "in" indication system and '.he 2203 239

1-4 Resolved Issues Status slack cable indication should be dis-played on the control board within opera-tor view. (Supplement Section 7.7.1) 1.8.2 Outstanding Issues a. The applicant must substantiate that The applicant and the he has the authority to determine all Penn Central Transpor-activities including exclusion or tation Company have removal of personnel and property agreed to establish con-and also controls railroad traffic trol procedures in case within the exclusion area. of an emergency. The (Supplement Section 2.1.2) applicant is filing con-demnation procedures to acquire those portions of land in the exclusian area not owned by PECo. b. Descriptions of the qualification test Commitment made for programs and acceptability requirements a satisfactory resolution. that are to be established for all of the new and unproven components in the electrical, instrumentation, and control systems. The applicant has committed to provide the information. (Supplement Section 7.10) 2203 240 g. c .i .i l s

1-5 Outstanding Issues Status c. Justification for the use of only three Commitment made for wide range logarithmic channels of a satisfactory nuclear instrumentation. (Supplement resolution. Section 7.2) d. Information on the reactor core instru-Commitment made for mentation systems to establish their a satisfactory adequacy to keep the reactor core resolution. within established operational and safety limits and the provision of redundancy in this design. (Supple-ment Section 7.7.6) e. An upiated financial report reflecting Completed the applicont's latest financial qualifications for the staff to per-form their financial evaluation. (Supplement Section 20.0) 2203 241

2-1 2.0 SITE CHARACTERISTICS 2.1 Geography and Demography 2.1.2 Site Description In Amendment 25, the applicant has indicated that discussions were held with officials of the Penn Central Transportation Company for control of rail traffic within the exclusion area during a plant emergency requiring such control. The result of this meeting was an agreement that the Penn Central Transportation Company would fully cooperate with PECo in establishing control procedures. In addition, the applicant will file in the Court of Common Pleas for Lancaster County Declarations of Taking for the condemnation of property for those portions of land in the exclusion area not owned by PECo. If this is accomplished, there is reasonable assurance that the applicant will be able to control all the activities within the exclusion area. 2203 242

3-1 3.0 DESIGN CRITERIA FOR STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS 3.9 Mechanical Systems and Components 3.9.1 Dynamic System Analysis and Testing With regard to flow-induced vibrational testing of reactor internals for Fulton Generating Station Units 1 and 2 the applicant has signified full compliance with Regulatory Guide 1.20, designating Unit 1 as the prototype. We have reviewed the guidance in that document for applicability to HTGR plant applications. We find that full compliance with Regulatory Guide 1.20 with respect to vibration measurement on reactor internals, including all systems e.nd components that are within or form a part of the primary coolant system boundary, can provide an effective vibration testing program acceptable to both the staff and the ACRS. We will review the preoperational vibration test program that will be performed in accordance with Regulatory Guide 1.20 for assurance that it constitutes an acceptable basis for demonstrating the design adequacy of the reactor internals in satisfying the applicable requirements of NRC General Design Criteria 2 and 14. In addition, at the FSAR review stage we will evaluate the preoperational vibration test program proposed by the applicant for verifying the design adequacy of the reactor internals urder loading conditions that will be comparable to those experienced during operation. The combination of predictive analysis, tests, and post-test inspection will provide adequate assurance that the reactor internals can be expected to withstand operation-induced vibrations without loss of structural integrity during their service lifetime. 2203 243

4-1 4.0 REACTOR 4.2 Mechanical Design 4.2.2 Reactor Vessel Internals The PSAR describes the reactor internal structure, and dis-cusses the compressive strength, mechanical properties, and the oxidation and irradiation resistance of structural graphite to be used in the Fulton high temperature gas cooled reactor. We have not conducted an independent confirmation of the adequacy of the actual design of the core structural supports. However, we are in the process of giving a short-term con-tract to an independent technical organization to evaluate and review generically the design, analysis and design criteria of all the graphite structures inside the PCRV. Furthermore, the Office of Nuclear Regulatory Research of NRC has an ongoing program of research at Brookhaven National Laboratory and at Los Alamos National Laboratory in the area cf the possible degradation of the core structural support. This is a long-range program and the results of research will not be available before a decision must be made on issuing a construction permit for the Fulton Generating Station. Therefore, we have established the adequacy of the design of the core support structures on the basis of the infor-mation provided by the applicant. Under operating conditions, the static load capacity of the core support structure exceeds by five times the primary design load. The allowable stresses were selected to result in a more conservative design than required for metallic components by 2203 244

4-2 Section III of the ASME Code. In addition, the graphite is loaded in compression, where the strength is higher and more predictable. The structure is tested to prove the design criteria. The reduction of strength in the core support structure as a result of steam generator leakage is estimated to be minimal in relation to the equilibrium oxidation occurring during normal reactor operation. The graphite weight loss in the support structure for each steam ingress is estimated to be 0.01 percent for a design accident consisting of a 90 lb/sec. leakage. This is compared to a weight loss of 1.1 percent estimated for the core support structure during the reactor lifetime subjected to a water content in the primary coolant of 10 ppm. The latter loss is equivalent to a reduction of strength of less than 30 percent. The effect of graphite impurities o.1 the rate of oxidation is being developed on a generic basis. We concur with the applicant that the properties of the structural graphite are adequate to meet the performance criteria of the Fulton high temperature gas cooled reactor. Our concurrence is based on the information that the compressiv strength exceeds by five times the load requirements, the irradiation damage is negligible at the anticipated fluence, the reduction in strength resulting from oxidants in the coolant will not exceed the designed safety factor, and the structural adequacy has been demonstrated through proof tests on the graphite. 2203 245 b ;. s. c 5, 3

5-1 5.0 REACTOR COOLANT SYSTEM 5.3 Prestressed Concrete Reactor Vessel The staff has evaluated the need of an independent design confirmation for the PCRV for the Fulton Generating Station and concludes that such confirmation is not considered necessary for the following reasons: An inaustry code exists providing detailed guidance to the a. designer. The staff actively participated in the development of this code which also was reviewed by the staff. b. The PCRV will be tested prior to operation. This test will enable correlation of analysis results with test results and will be reviewed by the staff at the FSAR stage of the review. The analytical techniques utilized for the PCRV have all been c. verified by extensive model tests conducted in this country and abroad. Strains and deflections measured during the model tests have been correlated with those determined by analysis, d. The seismic load, the most difficult load to simulate by structural testing, contributes a small portion to the total PCRV load. Further verification of the PCRV design adequacy will be on the basis of the preoperational structural acceptance test, including comparison of results therefrom with results from analytical calcu-lations. The applicant will supply submittals relating to results of the tests for our evaluation. 03 2zi6 e p' .1 s

5-2 5.4 Component and Subsystem Design 5.4.2 Main Helium Shutoff Valve The SER states that the primary coolant shutoff valve position indication is not satisfactory, and that we will require the appli-cant to provide a system that will indicate actual valve position in all modes of operation (overtravel, motor actuation and secondary release). The applicant in Amendment 25 revised the valve position indicator to provide a system that will indicate valve position in all modes of operation. Our discussion on the electrical and instrumentation fcatures is covered in Section 7.6.2 of this supplement. The revised valve position indication meets the staff's position. 5.4.5 Inservice Inspection and Surveillance Programs The proposed ASME Code, Section XI, Division 2, Draft Cl Edition (dated February 12, 1975) has been approved by the Subgroup on Isservice Inspection and will be issued shortly as a proposed standard for trial use. We have been actively participating in the development of this proposed standard. The staff position on HTGR inservice inspection requirements is as follows: a. The examination and inspection of Class 1, 2, and 3 components should be performed in accordance with the requirements of the proposed standard, Draft Cl Edition (dated February 12, 1975) of Section XI, Division 2, of the ASME Code. For components not covered by the Draft Cl Edition of the proposed 1 2203 247

5-3 standard, we are currently developing positions. When these positions have been developed, we will require the applicant, as necessary, to update his inservice inspection programs to meet our requirements. b. Until such time as better inspection techniques are developed, a visual examination should be performed at each refueling outage on the liner insulation (thermal barriers including SiO blocks), core support blocks, core support posts, cross 2 duct insulation and panels, and PCRV top and bottom plenums. The applicant has agreed that his proposed inservice inspection program will comply with the above staff position. Compliance with the inservice inspection required by the Code and the additional staff position will constitute an acceptable basis for satisfying the requirements of General Design Criterion 32, 10 CFR, Part 50, Appendix A. 2203 248

6-1 6.0 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.1 Containment Functional Design 6.2.1.2 Containment Dets..i Pressure r In the SER, the staff concluded that the minimum containment design pressure for the Fulton Generating Station should be 45.5 psig. In Amendment 27 to the PSAR, the applicant agreed to base the design on an analysis assuming no helium-air mixing resulting in a containment design pressure of 45.5 psig. We find the applicant's containment design accept-able. 6.3 Core Auxiliary Cooling System 6.3.3 CACS Helium Shutoff Valve The SER states that safety grade instrumentation that will give a direct indication of valve position (open or closed) will be required. The applicant in Amendment 25 revised the valve position indicator to be safety grade. Our discussion of the electrical and instrumentation features is covered in Section 7.3.2.5 of this supplement. The revised valve position indication design meets the staff's position. 2203 249

7-1 7.0 INSTRUMENTATION AND CONTROLS 7.2 Reactor Trip System In the SER ve required the applicant to provide justification for including only three channels of nuclear instrumentation for reactor trip input vs. six channels provided in the Fort St. Vrain design. The applicant has since stated "The detailed technical justification for this decision and its impact on instrumentation response during postulated transients and on requirements for trip setpoints is being considered in an analytical study to be completed by September 1, 1975. If, as a result of this study, it is indicated that critical limits are compromised, the use of a Fort St. Vrain type system will be implemented." We maintain that the results of this study will have to be reviewed by us in accordance with the procedures outlined in the SER and found acceptable. The applicant has also agreed to pro-vide a modified design and submit the results of a study related to the modified design for our review and acceptance if the present design cannot be acceptably demonstrated. We conclude that this pro-cedure is acceptable for issuance of a construction permit. 7.2.1 Anticipated Transients Without Scram In the SER we noted that we had reviewed the applicant's submission on anticipated transients without scran and found it incomplete in that the loss of all feedwater and the loss of normal onsite and offsite power transients were not evaluated. 2203 250 L

7-2 In addition, we advised the applicant and his NSSS eupplier, the General Atomic Company, to review the design of the control rod system, the reserve shutdown system along with the instrumentation, control and electrical support systems to assure that common components subject to common mode failures will not defeat the diversity claimed for the design. In response, the applicant has committed to providing a report on his detailed analysis in January, 1976. The report will be reviewed to assure that the design of the control rod system, the reserve shutdown system and their instrumentation control and electrical system meet the staff's position as outlined in the SER. 7.3 Engineered Safety Feature Actuation Systems 7.3.2 Core Auxiliary Cooling System 7.3.2.5 Auxiliary Primary Coolant Shutoff Valv_e_ In the SER we required the applicant to provide a complete description of this valve system by clearly defining the monitoring system for each flapper plate position indication in order to facilitate our review. In addition, we required that the instru-ment system satisfy the same criteria required foi the engineered safety feature system. We have reviewed the applicant's response and conclude that the system meets the Commission's requirements and is acceptable. 7.5 Safety Related Display Instrumentation In the SER we required the applicant to modify the design to meet the following $taff position regarding the post-accident and s incident monitoring system. The post-accident and incident 03 2E>1

7-3 monitoring system should provide two independent channels for monitoring each parameter defined in Table 7.5.1-1 of the PSAR with one channel being recorded, that these channels satisfy the requirements of IEEE Std. 279-1971, and be supplied _by separat.e and indepcadent Class IE power. The applicant hae provided modified design criteria which we find meet the above position. We therefore conclude that the design criteria for this system are acceptable. 7.6 All Other Systems Required for Safety 7.6.2 Main Loop Shutdown In the SER we required that the primary coolant shutoff valve position indication (full open and closed) for all modes of operation (i.e., with or without motor operation) be provided in the design and that these instrument systems satisfy the same criteria required for a safety-related system. We have reviewed the applicant's response describing the modified design and criteria and conclude that the design satisfies the Commission's requirements and is acceptable for the issuance of a construction permit. We note, however, that the applicant's reference to periodic testing addresses only part-stroke closure of the valves during reactor operations. The staff concludes that since these valves are designed to fully close when required, the Technical Specifications should include provisions to permit periodic testing for full stroke closure of these valves during reactor 9' TS 2203 252

7-4 operation unless otherwise justified. We will include appropriate testing requirements in the Technical Specifications during the operating license stage of review. 7.7 Control Systems 7.7.1 Control Rod System In the SER we required the applicant to provide control rod slacked cable indication for each control rod pair to be displayed on the control room control board and to satisfy the design criteria requirements provided for the control rod drive position "in" indication system. The applicant's modified design now includes two channels that are redundant to assure that integrity between the actual control rod position and the rod position transmitter will be maintained. One channel consists of a slack cable trans-mitter which provides input to the control rod system for interlocking and automatic inhibiting of control rod motion, and to the DAP system for continuous display and alarming of slack cable and manual slack cable bypass. The redundant channel is physically and electrically separate from the other. It consists of a separate slack cable transmitter which provides an input to the slack cable indicating pilot light displayed in the control All electrical equipment will be of high reliability and room. accuracy. Conformance to IEEE Std 317-1972, 323-1974 and 336-1971 is included. It will also be tested environmentally, seismically and operationally. The equipment in the DAP system control room 22LD3 253

7-5 control board is not included in this testing program. The capability for functionally testing the operability of the slack cable transmitters is provided by jogging the control rod drive during reactor operation and by reactor trip tests during shut-dcwn periods. The applicant states that the calibration of the slack cable transmitters and associated circuitry will be accomplished every eight years during the scheduled maintenance for the entire rod drive mechanism. We have reviewed the modified design and found that the eight-year interval between calibration of the transmitters will have to be justified by demonstrating the reliability or by including features in the design to permit annual calibration capability. The remaining pertion of the design satisfies the Commission's requirements and the staff requires that the above stated justification or design modifications be provided and included in the FSAR submittal. We conclude this to be acceptable for the issuance of a construction permit. The control rod drive position indicating system has been modified and the two redundant channels now conform to the codes and standards listed in Section 7.1.2.1 of the PSAR with one exception. This exception concerns the buffered output signal from each channel that provides input to the DAP system for continuous monitoring of the two rod position signals from each drive me-hanism and an alarm identifying deviation between the two transmitters in the same mechanism. This modified design 2203 254

7-6 includes a single rod withdrawal interlock system to preclude rewind on the cable drum in the mechanism on a rod inset t command if the "in" limit switch fails to stop drum rotation. The interlock is also part of the PPS and its redundant channels receive input from the rod position indication system that is arranged in a one of two logic. In our review of '51s modification, we find that the changes in addition to the slack cable system satisfy the Commission's requirements and are acceptable for assuring that a control rod cannot be withdrawn when an insert signal is actuated. This includes the capability of detecting when the integrity between the actual control rods and their associated rod position transmitters is lost. As stated in the SER, we still require a post-construction permit review and approval before the final design commences. 7.7.6 Reactor Core Instrumentation In the SER we required the applicant to supplement the information presented in the PSAR clearly describing the preliminary designs for (a) the instrumentation systems required for determining radial peaking, regional local power tilts, axial power shapes, and (b) the instrumentation from nuclear excore and incore systems, core plenum inlet and regional outlet temperature systems, regional orifice valve 1.osition systems and control rod position systems. In addition we required that redundancy be included in the design assuring conformance to the single failure criteria. 2203 255 -9x s

7-7 We have reviewed the revised preliminary description and design criteria provided by the applicant and conclude that they satisfy the requirements stated in Section 4.4 of the PSAR and are therefore acceptable for issuance of a construction permit. We note, however, the following deficiencies which we require to be included in the final system design. a. Provide deviation alarms for signals comparing power to flow region derived from rod position indication to core region flow rate derived from orifice position. b. Provide deviation alarms for signals comparing power to flow derived from PPS reactor trip instrumentation to steam generator inlet temperature. c. Supplement the description of the DAP system diagnostics and self checking to (a) include surveillance functions to closely monitor the DAP's capabil ty to insure compliance with LCO's and (b) to include Point Demand Function pro-viding capability to display or print a summary of all pointe currently deleted from processing. d. Provide equipment in the DAP system to ensure high overall system availability. In addition, the equipment should provide practical immunity to transient hardware failure; i.e., the capability to restart equipment following electrical transients. 2203 256 .c \\ f ' '.

7-8 These items result from a need for consistency throughout the system design as well as between the Summit and Fulton designs and have been discussed with the applicant. Since these items affect the applicant's original design criteria, we will review these required provisions in detail during the OL review stage. 7.7.7 Turbine-Generator System As indicated in the SER, we will require a post-construction permit review and approval of the twin-turbine feature in the steam and power conversion system. This is a consequence of the unique nature of the design and the fact that the design of the systems is preliminary. We will review this system to assure that systems important to safety are not adversely affected. 7.10 Qualification Test Program In the SER we required the applicant to provide the infor-mation in regard to the qualification test program for isolation devices, start-up detectors, moisture detectors, axial flow compressors and the complete electronics of the reactor trip system, plus other equipment identified in Sections 7.3.2.1, 7.3.2.3 and 7.7.1 of the SER. The applicant has submitted a report entitled " Preliminary Description of Representative Qualification Test Programs for HTGR Controls and Electrical Equipment" (GA-A13392) dated March 31, 1975. This report, currently being reviewed by the 2203 257

7-9 staff, does not completely satisfy all the requirements of IEEE 323-1974, is general in nature and incomplete in its present form as a valid reference document. However, the information presented and the commitments made provide adequate assurance that the staff's concerns can be resolved on a timely basis. We therefore conclude that the present report is adequate for the issuance of a construction permit, conditioned on the satisfactory resolution o'f these items during a post-construction review but before fabrication commences, as required in the SER. 2203 258

9-1 9.0 AUXILIARY SYSTEMS 9.1 Fuel Storage and Handling 9.1.2 Fuel Handling In the initial submittal of the PSAR both the fuel handling machine and the fuel transfer cask contained cooling systems in their design. In Amendment 7 the applicant changed the design of these components by not including any cooling systems. This resulted as a consequence of additional investigations performed by CAC which indicated that the fuel did nct exceed the 1000*F, the limiting fuel temperature with air. The staff requires more investigation of this design change to verify these new investi-gations. We will require that the applicant provide the results from his additional investigations for our review and comments prior to fabrication of these components. 2203 259

13-1 13.0 CONDUCT OF OPERATIONS 13.5 Industrial Security We have reviewed the plant designs and layout for potential enhancement of physic.al security and for the feasibility of implementing current NRC regulations that would be applicable to protection of the fuel at this facility and find that they do in fact contain features that would make it possible to comply with applicable requirements of 10 CFR Part 73 of NRC regulations. The applicant will submit a detailed Security Plan during the operating license review stage as proprietary information pursuant to 10 CFR 2.790, that will be revi aed and evaluated by the staff for compliance with regulatory requirements. 2203 240

14-1 14.0 INITIAL TESTS AND OPERATIM A detailed review 't the proposed initial test program which will include preoperation and start-up testing will be performed during the review of the FSAR and thus will be based on the final design of the plant and its components. The objective of this review will be to establish the adequacy of this test program prior to a decision on the issuance of an operating license. 2203 26l

17-1 l','. 0 QUALITY ASSURANCE 17.1 General The quality of the construction of all safety-related equipment will be inspected and certified to a quality assurance program which, as outlined in the SER, complies with 10 CFR 50, Appendix B. 2203 262

18-1 18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS) The Fulton Subcommittee of the Advisory Committee on Reactor Safeguards met with representatives of PECo and the staff to consider the Fulton application on January 7, 1975; February 20, 1975; and March 20, 1975. The full ACRS Committee meeting with PECo and the staff to consider the Fulton application was held on April 3, 1975. At the full Committee meeting of the ACRS on April 3, 1975, the ACRS noted a number of items that should be considered by the staff and PECo. These are as follows: An appropriate testing program to confirm design and a. operating features. (Discussed in Supplement Section 14.0) b. Progress in the research, development and testing of the relatively new safety-related components and systems. (Discussed in Supplement Section 1.6) The unique feature in which steam produced by each reactor c. will drive two turbines and whether this design feature will adversely affect important safety systems. (Discussed in Supplement Section 7.7.7) d. The Ltisfactory environmental qualification of vital instru-ments is under generic review by the staff to insure uniform requirements for all nuclear plants. (Discussed in Supplement Section 7.10) 2203 263 he 3:

18-2 Satisfactory resolution of questions on anticipated transients e. without scram. (Dir:assed in Supplement Section 7.2.1) f. Independent confirmation of the adequacy of the designs for the prestressed concret reactor vessel, the core structural posts, and the other vital structures. (Discussed in Supplement Sections 4.2.2 and 5.3) g. Reconfirmation of the adequacy of the acceptance criteria for the graphite which will be used for structural components. (Discussed in Supplement Section 4.2.2) h. Quality assurance for the constru : tion of safety-related equipment. (Discussed in Suppler ent Section 17.1) j. The development of a satisfactor ' inservice inspection program which includes a well-cc aceived and acceptable vibration testing program. (Di scussed in Supplement Sections 3.9.1 and 5.4.5) k. Review of the plant design and layout for potential enhance-ment of physical security. (Dircussed in Supplement Section 13.5) 1. Substantial information concerning performance of vital materials and components such as behavior of fuel, graphite moderator and structural members, insulation, liner, instru-mentation, valves, circulators, steam generators, and the prestressed concrete reactor vessel should be developed during power ascension and operation of the Fort St. Vrain 2203 264

18-3 reactor. This information is presently being reported and transferred to the Fulton project in the following ways: 1. Through monthly and semi-annual Operations Reports published by the Public Service Company of Colorado (PSCo). The fourteenth monthly report was issued for February, 1975. The second semi-annual report was issued in January, 1975. ii. Through ytarterly General Atomic Company Reports on the Fort St. Vrain Surveillance and Testing Program (USAEC Contract AT(04-3)-167. An example of these reports is GA-A13209 (UC-77) which is for the period from July 1, 1974, through September 30, 1974. iii. Through topical reports published by PSCo such as Fort St. Vrain Hot Functional Test Initial Summa v Report dated February 21, 1975. iv. Through Abnormal Occurrence Reports on Facility Operating License No. DPR-34, Docket No. 50-267. The ACRS believes that these items can be resolved by the staff and the applicant during construction. They therefore concluded that subject to the satisfactory resolution of these items the Fulton Generating Station can be constructed with reason-able assurnnce that it can be operated without undue risk to the health and safety of the public. The ACRS report which summarizes its review of the Fulton Generating Station application is included as Appendix C to this 2203 2,65 supplement to the SER.

20-1 20.0 FINANCIAL QUALIFICATIONS 20.1 Introduction Philadelphia Electric Compeny has applied for construction permits for the Fulton Generating Station, Units 1 and 2 to be located in Lancaster County, Pennsylvania. The applicant is a public utility engaged in supplying electric service in the city of Philadelphia and five surrounding counties in southeastern Pennsylvania, gas service in the five counties, and steam heat service in the central Philadelphia business district. As of December 31, 1974, it was serving a total of 1.5 million electric, gas and steam customers. The NRC regulations relating to financial data and informa-tion required to establish financial qualifications for applicants for facility ::::ttuction permits appear in 10 CFR Part 50, Paragraph 50.33(f) and Appendix C. In accordance with these regulations the applicant sabmitted financial information with its applicacion as well as several amendments thereto in response to requests by the staff for additional financial information. Our analysis regarding the financial qualifications of PECo focuses on the reasonableness of the company's projected system-wide financing plan for the period of construction of the subject facility and the assumptions underlying the plan. 2203 266 .I 'y hh

20-2 PECo's operating revenues increased from $789.9 million for the 12 months ended March 31,1974, to $1,102.5 million for the 12 months ended March 31, 1975. The net income increased from $124.2 million to $128.6 million over the same period. Invested capital as of March 31, 1975, amounted to $3,262.1 million and consisted of 51.9% long-term debt, 14.9% preferred stock, and 33.2% common equity. PECo's first mortgage bonds are rated "A", or upper medium grade by both Moody's and Standard and Poor's. The debentures are rated "Baa" by Moody's and "BBB" by Standard and Poor's, both medium grade ratings. 20.2 Construction Cost Estimates The applicant has submitted construction cost estimates for the two units itemized as follows: (dollars in millions) Unit 1 Unit 2 Total Nuclear production plant costs $1,365.9 $1,052.7 $2,418.6 Transmission, distribution and general plant costs 10.8 7.5 18.3 Nuclear fuel inventory cost for first core = 71.4 77.4 148,8 TOTAL $1,448.1 $1_,137.6 $2,585.7 The estimated cost of the nuclear production plant has been compared with costs estimated by the Energy Research and Development Administration's CONCEPT costing model. The CONCEPT costing model estimated the cost of the nuclear production plant to be $2,126.0 million. The applicant's estimate of $2,418.6 is 2203 267

20-3 13.8% above CONCEPT. The applicant has explained its higher estimate based on a number of factors including the construction of additional equipment and auxiliary facilities beyond what is minimally necessary for a plant of this type and size. Accordingly, we find the applicant's estimate reasonable. 20.3 Construction Program and Source of Funds 20.3.1 General In support of its application the company has submitted a Sources of Funds Statement or projected financing plan (see Appendix B) for its overall construction program for the years 1975 through 1986, the latest estimated year for completion of Unit 2. Included in the statement are the results of actual financing experience in the years 1973 atA 1974. The plan indi-cates that the construction expenditures (including the subject nuclear plant) will be financed in PECo's usual manner; i.e., through the use of retained earnings, depreciation, deferred taxes and other accruals; and through the sale of debt and equity securities in the form of common and preferred stock and long-term debt. PECo's planned gross construction expenditures (including Allowance for Funds Used During Construction - AFDC) for the five year period from 1974 through 1978 reflect the announcement made during 1974 of a $660 million (or 20%) cutback from $3.3 billion to $2.6 billion or to a reduced average of $526 million per year. 2203 268

20-4 This cutback involves a two-year delay in the completion dates of Units 1 & 2 to 1984 and 1986, respectively. The construction stretchout allows the company a longer period <f time in which, to secure the requisite funds. We regard this cutback as a prudent financial decision in light of existing economic and financial market conditions. Included in the reduced program are planned gross construction expenditures for 1975 and 1976 of $410 million and $492 million, respectively. These planned expenditures are reasonably conser-vative whau compared with PEco's actual gross property additions for the period 1970 through 1974. The comparable amount for this five-year period was $2.1 billion or an average of $420 million per year. For the years 1977 through 1986, when completion is planned for Unit 2, PECo plans average annual gross construction expendi-tures of $753 million. These annual expenditures, while significantly larger than the near-term plans, are not unreason-able when one considers the effect on construction costs of future inflation and if one assumes a reasonable regulatory environment which allows the company a return on investment sufficient to attract new capital and a resumption of growth in demand for electricity occurring in conjunction with a general economic recovery from t9e current recession. 2203 269 .e [j0', 1

20-5 20.3.2 Regulatory Environment In its electric rate order of March 25, 1975, the Pennsylvania Public Utilities Commission (PPUC) determined that the fair value of PECo's property used and useful in public service at June 30, 1974, was $2,355.0 million. In this order the PPUC granted $105 million or 77% of the three-part electric rate increase totaling $136 million requested by the company on January 31, 1974. The PPUC permitted the first part ($24 million) to go into effect on April 1, 1974, subject to possible refund. The second and third parts totaling $112 million became effective on January 1, 1975, also subject to possible refund pending the final rate order. The company is refunding an estimated J6.2 million in excess revenue collected from January 1, 1975, through April 9, 1975, and has placed the finally approved, increased rates into effect on April 10, 1975. In the above proceeding the PPUC allowed an 11.0% rate of return on the common equity component of fair value. In assessing this allowed rate of return we have considered the significant fact that Pennsylvania uses fair value to measure the rate base rather than original cost, as is used by many states and the Federal Power Cocmission. That is, if the applicant's rate increase were determined using an original cost rate base, a higher rate of return would be required to achieve the same increase in revenue. Furthermore, the 11.0% allowed rate of return is equatable with a return on book common equity of about 2203 270 I ix, o c.,

20-6 13.5 to 14.0%. The earned rates of return on average book common equity were 9.75%, 8.93%, and 8.68% for the years ended December 31, 1973, Decemger 31, 1974, and for the twelve months

• .t -

ended March 31, 1975, respectively. In its assumptions underlying the Sources of Funds statement for its construction program, the company includes rates of return on average book coumon equity of 9.1% in 1975, 11.7% in 1976, and an average of 13.1% for the years 1977 through 1986. The projections for 1975 and 1976 appear reasonable based on the company's actual rates of return in the three prior periods, coupled with the rate of return allowed in the most recent rate proceeding. The assumed rates of return for the years beyond 1976, although somewhat higher than the near-term projections, are not unreasonable when compared with the recent rate order and rates of return currently being allowed by various state commissions. When considering such projections, we have necessarily assumed that the regulatory authorities having jurisdiction over the company's rates will periodically adjust rates when necessary to allow the company to fully recover its cost of capital. This includes a return on common equity that will permit the company to successfully compete for construction funds in the securities market. 20.3.3 Extern,ql Financing - Future Plan and Recent History In developing its financing plan for the years 1975 through 1986, the planned year of completion of Unit 2, PECo projects that 2203 271

20-7 53% of its total construction expenditures will be provided by external funds and that its capital structure will be maintained at approximately 50% long-term debt, 13% preferred stock, and 37% common equity. The external sources include the sale of common and preferred stock and long-term debt. The company sold $149.3 million and $11.2 million of common stock in 1973 and 1974, respectively. Its projected external financing plan for 1975 and 1976 includes the sale of $140.0 million and $100.0 million, respectively, of common stock. In April of 1975, PECo sold approximately $50 million of common stock, or 36% of the total 1.975 proj ection. The 1976 projection for common stock sales reflects a 29% reduction from the 1975 level. Common stock sales for the period 1977 through 1986 are projected to average $75.0 million per year. We conclude that these projections are reason-able based on the allowed and projected rates of return on common equity discussed in Section 20.3.2 of this supplement. The company sold $75 million of preferred stock shares in 1973 and $75 million again in 1974. It projects no preferred sales in 1975, but $50 million in 1976. For the years 1977 through 1986, preferred stock sales are projected to average $45 eillion per year. Based on the company's recent history of successful preferred stock sales and on the previous assumption of rational regulation we conclude that the company's projections of preferred stock financing are reasonable. 2203 272

20-8 In 1973, PECo sold $100 million in first mortgage bonds. In 1974, the company sold $250 million in first mortgage bonds and a $125 million 5-year bank note. It projects 1975 long-term debt sales of $290 million. Through April 1975, the company has com-pleted a $65 million sale of first mortgage bonds and a $100 million sale of debentures. Thus, it has completed 57% of its 1975 projected long-term debt financing in the first 4 months of the year. The company has announced the proposed sale of an additional $80 million of first mortgage bonds in August 1975, in continuation of the overall financing plan. Long-term debt issuance in 1976 is projected at $200 million, or a 31% reduction from the 1975 plan. Long-term debt issuances for the years 1977 through 1986 are projected _s average $247.5 million per year. The company's first mortgage bonds are issued under an indenture which requires that additional bonds may not be issued unless earnings before income taxes and interest are at least 2.0 times the pro forma annual interest requirements on all such bonds outstanding and applied for. Such interest coverage for the 12 months ended March 31, 1975, was 2.26 before giving effect to the April 1975 issuance of $65 million first mortgage bonds and was 2.10 after being pro formed for the new issue. However, this issue was not subject to the interest coverage test under the indenture since it was issued against a like amount of maturing bonds which had been paid for or for which payment had been 2203 273 n .$ \\.

20-9 deposited with the trustee. PECo's debentures can be issued without regard to any such interest coverage test. As an assumption underlying its proposed mortgage bond issuances, PECo projects that mortgage interest coverage will recover to approximately 2.5 at the end of 1975 and to 3.1 at the end of 1976. We find these projected coverages to be within a range of reasonableness based on the company's latest electric rate increase and its annualized effect on earnings as discussed in Section 20.3.4 of this supplement. 20.3.4 Internally Generated Funds - Projections and Recent History The company projects gross construction expenditures of $7,190.3 million for the years 1975 through 1986, the estimated year of completion of Unit 2. Of this total, the company projects that it will raise $3,359.8 million (or 47% of the total) from Internal sources. These internal sources include retained earnings, caferred taxes, investment tax credit adjustments, depreciation and amortization, nuclear fuel charged to operations, and other accruals. Actual internally generated funds as a percentage of construction expenditures for the five-year period 1970 through 1974 was 8.9%. The company projects tiat this percentage will be 24% in 1975, 30% in 1976, 27% in 1977, and that it will rise steadily thereafter, resulting in the 47% average over the 1975 - 1986 period. While this average is higher than the recent historical experience of the company noted g abo it is not unreasonable when considered in light of major p 2203 274

20-10 increases in depreciation and nuclear f uel charged to operations during the period, which are a result of significantly higher net plant in service expected during that period. PECo projects the total of depreciation, amortization, and nuclear fuel charged to operations for the period 1975 through 1986 to be $2,778.0 million, or 83% of internally generated funds during the period (39% of total construction expenditures). The company anticipates that the follow.cg major project additions will go into service d Jing Units 1 and 2 construction period: Eddystone No. 4; Salem Nuclear Units 1 and 2 (42.6% interest owned by PECo); Limerick Nucicar Units 1 and 2; and Summit Nuclear Units 1 and 2 (15% interest owned by PECo). The balance of internally generated fends is expected to come from an improved earnings performance. Retained earnings are projected to supply $1,088.0 billion, or 15% of the company's total construction budget for the years 1975 through 1986. During the past 12 to 15 months, the company has been affected by the same economic and financial hardships that have confronted the entire electric utility industry. These have included rapid increases in all costs of operation, sky-rocketing fuel costs, due in part to the Arab oil situation, and the significantly increased cost of raising capital funds. In addition, the phenomenon of regulatory lag in providing rate increases has prevented the company from being able to fully recover these increased casts on a timely basis. Consequently, earnings per average share of common stock declined from 9 \\

20-11 for the year ended December 31, 1973, to $1.76 for the 12 months ended March 31, 1975. However, cash earnings available for common stock (defined as net income after preferred dividends, plus depreciation, deferred income taxes, investment tax credit adjustments, and minus the allowance for funds used during construction) increased from $2.37 to $2.81 per average common share over the same period. The cash earnings available for common stock have thus provided adequate coverage throughout the period for the company's common stock dividend which has been maintained at the annual rate of $1.64 per share. In analyzing the company's earnings in relation to the most recent electric rate increase discussed above, it is noted that the 1974 earnings ($129.1 million net income, or $1.81 per average common share) include the effect of a 9-month portion of only the first interim part ($24 million) of the three-part requested increase. The carnings for the 12 months ended March 31, 1975, ($128.6 million net income, or $1.76 per average common share) include the effect of 12 months of the $24 million interim increase but only 3 months of the additional $112 million interim rate increases. These interim increases were not suspended until April 9, 1975, at which time the final total increase of $105 million was placed into effect. Thus the full annualized impact of the total increase will not be reflected in earnings until the new rates have been in effect for a year, or until December 31, 1975. In light of this analysis and in the presence of a 2203 276

20-12 reasonable regulatory environment, it can be assumed that the company will continue to adequately cover its current common dividend and increase it moderately over time. Two additional measures of the reasonableness of the near-term internal financing projections are a comparison of the 1975 and 1976 plans with the actual financing experience of 1973 and 1974, and an aralysis of the first quarter actual progress of the 1975 financing projections. A similar analysis of external financing was discussed earlier in this section. In 1973 and 1974, the company generated $16.4 million and $8.3 million, respectively, of retained earnings. It projects that $10.6 million and $35.4 million of retained earnings will te generated in 1975 and 1976, respectively. Significantly, the company has already generated $3.6 million in retained earnings in the first quarter of 1975, or 34% of the year's projection. In light of the annualized effect of the $105 million rate increase discussed above, it appears that the company can achieve its 1975 retained earnings projection. The 1976 retained earnings projection of $35.4 million results from a projected 33% increase in net income over 1975 and a 17% increase in total dividends over 1975. Based on the latest electric rate increase and on the company's assumption of an additional 10% electric rate increase in 1976, we believe that the 1976 projected retained earnings are within a zone of reasonableness. 2203 277 ce

20-13 Projected increases in net income and retained earnings for the years beyond 1976 are based on a number of assumptions made These by rhe company in developing its projected financing plan. assumptions include: (a) an annual percentage growth rate in KWH sales of approximately 6 percent; (b) periodic rate increases for electric, gas, and steam services; and (c) growth rates in operating expenses and interest charges that will be fully covered Based on by the combination of sales growth and rate increases. the previously discussed assumptions of a reasonable regulatory environment which provides rate increases sufficient to cover the cost of capital and on a resumption of growth in demand for electricity, we conclude that the company's long-range projections of retained earnings are within the zone of reasonableness. 20.4 conclusion Based on the preceding analysis, we conclude that Philadelphia Electric Company is financially qualified to design and construct Fulton Generating Station Units 1 & 2. This conclusion is based on our determination that PECo's projected financing plan is within the zone of reasonablenes= and thus providcs reasonable assurance of obtaining the funds necessary to complete the design and construction activities over the life of the construction permits. 2203 278

APPENDIX A SUPPLEMENT TO CHRONOLOGY OF THE RADIOLOGICAL REVIEW FOR FULTON GENERATING STATION, UNITS 1 AND 2 March 5, 1975 Safety Evaluation Report issued March 7, 1975 Submittal of Amendment No. 24 to the PSAR, consisting of additional information relating to the foundations on Which containment, control building and reactor service building are located; clarification of containment liner temperature analysis under DBDA conditions; clarification of PCRV support subcompartment pressure analysis; and elimination of use of propane on site. March 20, 1975 ACRS Subcommittee meeting March 25, 1975 Submittal of Amendment No. 25 to the PSAR, consisting of information on resolved and outstanding items identified in SER April 3, 1975 ACRS meeting April 8, 1975 ACRS Report issued April 15, 1975 Submittal of Amendment No. 26 to the application, consisting of additional financial information April 17, 1975 Submittal of Amendment No. 27 to the PSAR, consisting of additional information on resolved and outstanding items identified in SER April 24, 1975 Letter to applicant requesting additional financial information May 15, 1975 Submittal of Amendment No. 28 to the application consisting of additional financial information May 30, 1975 Applicant submits Amendment No. 29 to the PSAR, consisting of additional information on core performance instrumentation; post accident monitoring; and shutoff valve position indication systems 2203 279

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APPENDIX C ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 April 8, 1975 Honorable William A. Anders Chairman U.S. Nuclear Regulatory Commission Wash'ington, D. C. 20555

Subject:

PEPORT ON FULTON GENERATING STATION UNITS 1 AND 2

Dear Hr. Anders:

At its 180th meeing, April 3-5, 1975, the Advisory Connittee on Reactor Safeguards completed its review of the application of the Philadelphia Electric Company for a permit to construct the Fulton Generating Station, Units 1 and 2. The Committee reported previously on the Conceptual Desigt. for a Large High Temperature Gas Cooled Reactor (HTGR) in its letter of November 12, 1969; that design was a prototype for the Fulton Generating Station. Subconnittee nectings were held in Lancaster, Pennsylvania on January 7,1975, in connection with a site visit, and in Des Plaines, Illinois on March 20, 1975. In addition, a General Atomic Company Subcommittee meeting was held in Denver, Colorado on January 30-31, 1975. In its review, the Cocnittee had the benefit of discussions with representatives and consultants of the Philadelphia Electric Company, the General Atomic Company, and the Muclear Regulatorf Commission (NRC) Staff. The Committee also had the benefit of the docu-ments listed. The Fulton Generating Station will be located in Fulton and Druncre Townships on the cast bank of Conowingo Pond, approximately 17 ciles south of Lancaster, Pennsylvania. The Peach Botton Atonic Power Station, which is also owned and operated by the Applicant, is located opposite the site on the west bank of Conowingo Pond. The Applicant has prcposed an exclusion area with a radius of 2500 ft., and a low population zone with a radius of 1.5 miles. The nearest population center is Lancaster, approxinately 17 miles north of the site, with a 1970 census population of about 58,000. The 1970 population within 50 miles of the site was about 4.4 million, and is expected to increase to about 9.7 million by 2020. The Fulton Generating Station consists of two nuclear units, each using a General Atomic High Tenperature Gas-Cooled Reactor (UTCR) having a rated power level of 1160 MW(e). All safety systems were analyzed and designed for 3150 MU(t). The reactors for the Fulton Station are very similar to those for the Sumnit Station except that they have six loops rather than four for Sunnit. 21203 281

Honorable William A. Anders April 8, 1975 Since this plant will be the prototype for future six-loop HTGks, an appropriate testing program to confirm design and operating features will be required. The Conmittee wishes to be kept inforned of progress in research and developaant and testing of components critical to safety such as primary circulators, primary valves, core auxiliary cooling systems, and insulation, and in verification of prestressed concrete reactor vessel design and steam generator performance. The Fulton Station will incorporate a feature unique to nuclear power plants in the United States; the steam produced by each reactor will be used to drive two turbines, of approximately 600 MU(c) rating each, arranged in parallel. A review of this feature to assure that systems important to safety are not affected adversely will be conducted by the NRC Staff during construction. The Committee wishes to be kept informed. The similaritics between the Fulton and Summit Stations have been taken into account in the Committee's review. Uith no significant exceptions, the Committee's concerns are generic to both the Fulton and Sumnit Stations and to all large HTGRs. This is reficcted in the following connents which are essentially the sane as those made in the Committee's report of March 12, 1975 on the Summit Pouer Station. The Committee recognizes that the v lton Generating Station represents u a new design so that many of the proposed systems and components are relatively untested at this time. This aspect is apparent in the NRC Staff Safety Evaluation Report (SER) where several items are unresolved or resolution is to be deferred until the post-construction pernit period. The Concittee urges the resolution of these outstanding items well before equipment is installed. A s> ;nificant number of outstanding items remain in the field of nuclear instrumentation, moisture monitors and various electrical systems. Particular attention should be given to the environmental qualification of vital instruments prior to installation. These items should be resolved to the the satisfaction of the NRC Staff. The Committee wishes to be kept informed. Further information is being developed by the Appl _icant and his contractors with regard to the subject of anticipated transients without scram. This matter should be resolved in a manner satisfactory to the NRC Staff and the ACRS. The NRC Staff is gaining an independent capability for accident analysis of HTGRs. The Conaittee believes this is an appropriate step. The Committee recommends that the NRC Staff also assure that appropriate independent confirmation of the adequacy of actual design exists for the PCRV, core structural supports, and other vital structures for this prototype reactor. m M 2203 282

Honorable William A. Anders April 8, 1975 Substantial infornation concerning performance of vital materials and components such as fuel, graphite moderator and structural neebers, insulation, liner, instrumentation, valves, circulators, steam ;;cnerators, and PCRV should be developed during power ascension and operation of the Fort St. Vrain Reactor. The NRC M.aff should reconfirm the adequacy of perfornance criteria for graphite used in structural couponents, including such factors as permissible level of impuritics, acchanical behavior, acceptabic flav sizes, and dimensional changes due to neutron irradiation. The Committee reiterates its interest in construction to high quality standards and in the development of wcll-conceived surveillance and inspection prograns for vital components. Current progress on the ASME Section XI Division 2 Code for Inservice Inspection is an acceptable beginning. Continued effort is required to develop inspection criteria for vital components such as insulation, graphite structures, circulators and steam generators. Sinilar prograns are required for the PCRV tendons. These programs should cover both the integrity ot' vit: 1 components and their operational reliability. A necessary aspect of the surveillance testing of this prototype plant is a ucll conceived vibration testing program acceptable to both Staff and ACRS. 'I i ( The Committee recocuends tNat the NRC Staff nnd the Applicant review the plant designs and layout for potential enhancetsnt of physical security, particularly the protection of the fuel. The ACnS believes it advisable to review the various outstanding items cited in this report and the SER in approximately 12-18 months. The Advisory Conmittee on Reactor Safeguards believes that the above items can be resolved by the Applicant and the MRC Staff during con-struction. Subject to the satisfactory resolution of these itens the Committee believes that the Fulton Generating Station can he constructed with reasonable assurance that it can be operated without undue risk to the health and safety of the public. Sincerely yours, Uilliam Kerr Chairman References attached. 2203 283

Honorable William A. Anders April 8, 1975 References 1. Philadelphia Electric Company Application and Preliminary Safety Analysis Report (PSAR) (Volumes 1-6), for Fulton Generating Station Units 1 and 2. 2. Amendments 1-25 to the Fulton Cenerating Station Units 1 and 2 PSAR. 3. Proprietary report entitled " Detailed Discussion of Materials Used for Compressional and Shear Wave Velocity Measurements and How Shear Waves are Identified from the Wave Train" (ated November 16, 1973. 4. AEC Licensing Staff, Advanced draft of Chapter 2 of the Safety Evaluation Report, Issued January 2, 1975. 5. NRC Licensing Staff, Safety Evaluation Report (NUREC-75/015), Issued March 1975. )_}0b

D-1 APPENDIX D ADDITIONS, DEI.ETIONS, AND CORRECTIONS TO THE SAFETY EVALUATION REPORT Page Line Comments 1-9 2 Change " hemispherical" to "torispherical". 1-20 5 Change to read: "(1) the steam / water dump system piping from the dump valves to and including the dump tank was changed from Quality Group D to Quality Group C". 1-20 9 Delete "and cleanup". 1-23 15 Change to read: "(d) The slack cable indication system should satisfy the same design criteria requirements provided for the control drive position "in" indication system and the slack cable indication should be displayed on the control board within operator view." 1-24 8 Add a period after the word " systems" and delete the remainder of the sentence. 1-24 11 Delete the comma between 7.3 and 7.7; insert "and" between 7.3 and 7.7. 1-24 12 Delete. 1-24 19 Delete item e. and insert as item h. under Post-Construction Permit Issues. 1-24 23 Change f._to e. 1-26 Add item 1: " Substantiation of the design changes of the fuel handling machine and the fuel transfer cask. (Section 9.1.2)" 1-30 9 Change "105.4" to "100.5". 2-2 20 Change to read: "The applicant has made preliminary arrange-ments with the Bureau of Radiological Health of the Pennsylvania Department of Environmental Resources and with the Bureau of Waterways of the Pennsylvania Fish Commission to control...". 2-11 6 Change "50" to "150". 2203 285

D-2 Page Line Comments 2-14 24 Change to read: "... located about one and one-half miles downstream of the plant..." 2-17 23 Change to read: "... unregulated except for the influences..." 2-21 1 Change to read: "... bedrock (fresh to moderately weathered slate)..." 2-22 11 Change to read: "...mainly red beds of sandstone, congleuarate, and siltstone." 2-24 10 Change "Quanternary" to Quaternary". 2-25 21 Change to read: "...the applicant as being 12, 9, and 14 miles long..." 2-26 18 Change " exposed to " inferred". 2-30 11 Delete " multiple". 2-31 5 Change to read: "We have visited the site on four occasions. The third site visit was made during the spring, 1974.... fault. The last site visit was made in January,1975, with the ACRS Subcommittee." 2-36 9 Change to read: "...f ault and probably fc tmed at the same time." 2-36 12 Change to read: "...although, according to the applicant, splays of the shear zone indicate tensional movement as well. These tensional movements may be relarad to Triassic stress conditions." 2-36 16 Change " zone" to " zone (s)". 2-37 11 Change "382 feet msl" to "362 feet ms1" and delete "to moderately". 2-38 4 Change " concreted" to " concrete". 2-38 7 Change " Modified Proctor" to " ASTM D1557 method". 2-39 2 Change to read: "... factor of safety obtained from the analyses was 1.88. Dynamic analysis showed the slope was stable during the SSE." J-4 17 Delete "and cleanup". 2203 286 - s '.

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D-3 Page Line Comments 3-6 M Change to read: "The design tornado specified for the plant has a tangential wind velocity of 290 mph and a translational velocity of 70 mph. In addition, the rate of pressure drop to be used in the design of buildings requiring tornado protection is 2 psi per second. These specified tornado designs are consistent with Regulatory Guide 1.76." 4-14 15 Change " spheres" to " pellets". 5-7 20 Change to read: "The PCRV pressure relief system is designed and analyzed in accordance with ASME Code Section III, Division 2." 5-7 22 Delete: "The remainder of the piping up to and including the safety relief valve is designated and analyzed in accordance with Section III of the ASME Code, Class I (Quality Group Class A)." 5-9 13 Delete: "The plans to avoid sensitization are in general conformance with Regulatory Guide 1.44 and include controls on composition, heat treatment, welding processes and cooling rates." 5-9 18 Change to read: "The use of materials with satisfactory service experience provide reasonable assurance...". 5-10 15 Change comma after 1 to a period. 5-10 16 Delete: " including Appendix G of the Summer 1972 Addenda." 18 17 5-10 22 Change "1.3 x 10 nyt" to "6.8 x 10 nyt". 5-10 19 Delete " and weld..." 5-11 7 Change to read: "All penetrations and closures will be designed to the impact requirements that the lowest service temperature will not be lower than RTg + 60*F." 5-11 14 Change comma after 2 to a period. 5-11 14 Delete: "... including Appendix G, Summer 1972 Addenda, and Article CB-2500." 2203 287 y ss

D-4 Page Line Comments 5-13 6 Change to read: "...ll50'F to develop the optimum mechanical properties, including toughness in this material for this application." 5-13 21 Delete: "The water purit y will be maintained to the requirements of Regulatory Guide 1.56." 5-14 5 Change to read: "...a minimum of three percent and a maximum of twelve percent in delta ferrite...". 5-14 7 Change comma after III to a period. 5-14 7 Delete "... Paragraph NB-2433." 6-2 6 Change " hemispherical" to "torispherical". 6-15 6 Change to read: "The system is comprised of three 50% capacity trains so that the failure of any single active..." 6-15 16 Delete "demister". 6-16 5 Change "all" to "the objectives" 6-19 5 Change ts read: ... valve to open and the steam inflow is terminated after the valve resets." 6-27 8 Change "2120 gpm" to "2150 gpm". 6-35 24 Change to read: ... system equipped with two redundant 3000 c fm... " 6-36 1 Change "1500 cfm" to "3000 cfm". 6-36 2 Change to read: ... remote air intakes. Each of the two filter trains will process 1500 cfm of this air in addition to 1500 cfm of recirculating control room air. The staff..." 6-39 Table 6-1 Change "42 psig" to "45.5 psig" and change "0.15% per day" to "0.1% per day". 7-5 21 Delete " start-up detectors," and delete the comma after the words moisture detectors. 7-14 5 Change to read: ...and in addition may be used to remove..." 2203 288

D-5 Page Line Comments 8-8 3 Change to read: " Vital 120/208 volt a-c control..." 9-2 18 Change to read: "The fuel storage facility will consist of fuel storage wells embedded in a concrete monclith in the reactor service building which will store both spent fuel and reflected elements." 9-16 3 Change " Table 3.5.5-1" to " Table 3.5.3-1". 9-16 5 Change "500 psig" to "535 psig". 9-18 24 Change to read: "The purge lines from the PCRV penetra-tion out to and including the check valves will be designed to seismic Category I requirements." 9-24 7 Change "outside the turbine building," to "outside of Unit 1 circulating water basin,". 9-24 8 Change "outside the control building," to " portable extinguisher in the control room." 9-30 Table 9-1 Change Auxiliary Circulator Motor Water Coolers peak accident 6 Btu /hr". heat load "2 x 106 Btu /hr" to "0.5 x 10 9-30 Table 9-1 Change PCRV Liner Cooling Coils peak accident heat load 6 Btu /hr". "14 x 106 Btu /hr" to "15.7 x 10 9-30 Table 9-1 Add item f. " Containment Penetration and Fuel shipping Container Storage. Normal reactor operation heat load is 1.18 x 106 Btu /hr and the peak accident heat load is "1.18 x 106 Btu /hr". 11-1 18 Change to read: "Certain ventilation subsystems will treat gaseous streams by means of filtration and adsorption." 11-11 5 Change to read: "During normal operation, the reactor containment may require venting once a year to reduce containment pressure. In addition, the containment will be fully purged prior to refueling on a once-through basis at a rate of 48,000 cfm. The purge effluent will be discharged to the atmosphere through the ventilation exhaust duct on the reactor service building. The reactor service building will be ventilated by a once-through system. Exhaust air will be discharged to the atmosphere through the ventilation exhaust duct. Ventila-tion air from each turbine building will be exhausted to the atmosphere through a single roof top vent on each turbine building." 2203 289

D-6 Page Line Comments 11-16 2 Change to read: "... exhausted to the reactor service building ventilation system. Displaced air from drumming operations will be exhausted to the process vent system." 11-16 13 Change "45,000 Ci" to "6000 Ci". 11-20 Table 11-1 Change "35 scf" to "35 scfm". 15-15 18 Change to read: "... work properly, the radioactive gas leakage..." 15-15 20 Delete: "If cooling of the casks is not maintained, then due to the high heat capacity of the graphite elements there will be sufficient time to transfer the fuel to storage or back into the core before there would be a significant release of radioactivity." 17-2 11 Change to read: "...the QA Program at least once per twelve-month period." 17-3 8 Change to read: "The qualifications include technical competence, administrative skills, and ability to develop and implement Quality Assurance Program Policy." 17-4 18 Change to read: "...QA Program to implement specific drafts of ANSI Standards which are listed in Chapter 17 of the PSAR." 17-6 22 Change to read: " Engineering assurance policy for design and engineering is established by the Engineering Manager through the Chief Engineer, Engineering Assurance. Quality Assurance and Quality Control (QC) policies, in all areas except engineering and design, are established by the Vice President, QA. The Manager, QA, coordinates and implements the QA and AC policy and is responsible for the QA and QC manuals." 17-7 15 Change to read: "...is implementing corporate QA policies and programs. A detailed...". 17-12 13 Change to read: "... design work are designated by the responsible Department Manager...". 2203 290 m. 3

D-7 Page Line Comments 17-8 18 Change to read: ...to implementing specille drafts of ANSI Standards which are listed in Chapter 17 of the PSAR." C-6 15 Insert item 66(a): "U. S. Nuclear Regulatory Commission Regulatory Guide 1.76, ' Design Bases Tornado for Nuclear Power Plants,' USNRC, Office of Standards Development, Washington, D. C., April, 1974." 2203 291 .}}