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Home My WebLink About851181.tiff 0 no, PEGV4 ro9L UNITED STATES co a o NUCLEAR REGULATORY COMMISSION w ; WASHINGTON.D.C.20555 oY�o 4 November 5, 1985 Docket No. 50-267 ?ElE Ctl NE-iFIn 3 "" N^V45t9 Mr. 0. R. Lee, Vice President p„ Electric Production Public Service Company of Colorado + EELeY• O°`o P. 0. Box 840 Denver, Colorado 80201 SUBJECT: FORT ST. VRAIN EQUIPMENT QUALIFICATION Dear Mr. Lee: On October 29, 1985, the NRC staff met with you to discuss the Fort St. Vrain Equipment Qualification Program and your request for an extension of the deadline for compliance with the equipment qualification rule. At the conclusion of that meeting, I stated that we still had many unanswered questions about both your extension request and overall program, and agreed to provide as soon as possible a list of the additional information we required to resolve these questions. The required information is listed in the enclosures. These questions, and any others that arise during the course of our review, must be resolved before we can conclude that FSV is in compliance with the equipment qualification rule. Draft copies of the enclosures were provided by telecopy to Mr. Holmes of PSC on November 1, 1985. During the meeting we also committed to provide an NRC staff position on the question of whether the rapid depressurization of the PCRV (DBA #2) must be considered in establishing environmental requirements for the equipment qualification program. Although we are still considering this issue, our preliminary conclusion is that this event must be considered. We will provide our final conclusion on this question as soon as our review is complete. The information requested in this letter affects fewer than 10 respondents; therefore, OMB clearance is not required under P.L. 96-511. Sincerely, Edward J. Butcher, Acting Chief Operating Reactors Branch No. 3 Division of Licensing Enclosures: As stated cc w/enclosures: See next page 851181 ,3aMYa IFld-85 Mr. 0. R. Lee Public Service Company of Colorado Fort St. Vrain cc: C. K. Millen Albert J. Hazle, Director Senior Vice President Radiation Control Division Public Service Company 4210 East 11th Avenue of Colorado Denver, Colorado 80220 P. 0. Box 840 Denver, Colorado 80201 J. W. Gahm Nuclear Production Manager Mr. David Alberstein, 14/159A Public Service Company of Colorado GA Technologies, Inc. P. 0. Box•368 P. 0. Box 840 Platteville, Colorado 80651 Denver, Colorado 80201 J. K. Fuller, Vice President Public Service Company of Colorado P. 0. Box 840 Denver, Colorado 80201 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. 0.Box 640 Platteville, Colorado 80651 Kelley, Stansfield & O'Donnell Public Service Company Building Room 900 550 15th Street Denver, Colorado 80202 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1800 Lincoln Street Denver, Colorado 80651 Enclosure 1 FORT ST. VRAIN REQUEST FOR ADDITIONAL INFORMATION EQUIPMENT QUALIFICATION PROGRAM EQUIPMENT QUALIFICATION BRANCH 1. Provide a description of the aging studies being performed and any interaction between equipment degradation and harsh environment accident conditions. (i.e., clarify the following statements: "The accelerated aging that occurs during the 4 minute isolation is very much different from that which occurs during a faster isolation time." "These lower temperature profiles will have a favorable impact in the areas of aging.") 2. Provide a description of the operability studies being performed, specifically how operability is being demonstrated when test duration is less than required equipment post-accident operating time. 3. Provide sample files (at least 3) which demonstrate how the above are being factored into the FSV EQ program. 4. Provide assurance that the equipment within the scope of 10 CFR 50.49 is being qualified to the most limiting environment resulting from a spectrum of break sizes. 5. Provide assurance that equipment accessibility is possible within the required time frame in the most limiting environment resulting from a spectrum of break sizes. 6. Provide assurance that all design basis events, as defined in 10 CFR 50.49, have been considered in the determination of harsh environments. Enclosure 2 FORT ST. VRAIN REQUEST FOR ADDITIONAL INFORMATION EQUIPMENT QUALIFICATION PROGRAM AUXILIARY SYSTEMS BRANCH 1. Provide a detailed description of the steam line rupture detection and isolation system (SLRDIS). This discussion should include the systems design basis including its capability to assure environments within acceptable limits following steam line breaks concurrent with a single failure. 2. Confirm that previous pipe break analyses have addressed equipment quali- fication concerns for failures in systems other than the steam lines, e.g. , main feedwater. Verify that these analyses have addressed protec- tion from flooding. 3. Provide the results of an analysis of a spectrum of postulated breaks in the main steam, and hot and cold reheat lines in the turbine and reactor buildings. Include the resulting temperature profiles. Confirm that small breaks, i .e. , those less than a full double ended break can be detected and isolated by the SLRDIS prior to exceeding the equipment, qualification envelop or unacceptably preventing access for required manual actions to achieve shutdown. In addition, provide response to the attached information request sheet in order to permit us to perform an independent calculation to verify your temperature profiles. 4. Provide information on the capability of the SLRDIS temperature sensors (thermistor cabling) to adequately detect elevated temperatures in the areas of concern. Verify that the sensitivity of these sensors is sufficient to provide proper indication/actuation in the event of localized temperature effects following steam line breaks. Include any available manufacturers test data and/or performance information on similar detectors in comparable applications. t - PRESSURE AND' TEMPERATURE PROFILES FOR PIPE BREAKS OUTSIDE CONTAINMENT The following information is- required for each pipe break analysis performed by the applicants. 1 . With respect to the pipe to be broken, we need to know the: a. Type of fluid (water or steam); b. Temperature; c. Pressure; d. Source of the fluid; e. Flow rate (or assumed flow rate) versus time; and f. Enthalpy versus time 2. With respect to the compartments being analyzed: a. Number of compartment analyzed; b. For each compartment: I. initial temperature ii. initial pressure • iii. initial humidity • iv. floor area including floor space taken by equipment (square feet) v. number of vents and vent areas (square feet) for each vent; and vi. compartment.wall height (feet) and , • c. Simple compartment and interconnection diagram. ' 3. All assumptions used, including but not limited to the: a. Orifice coefficient: b. Fluid expansion factor; and c. Heat transfer coefficient for heat through the walls 4. Utilities analysis results: a. Temperature versus time curve (peak temperature specified); b. Pressure versus time curve (peak pressure specified); and c. Humidity versus time curve (peak humidity specified) ', ``to Filet,/ 'W j°y° UNITED STATES NUCLEAR REGULATORY COMMISSION w ; WASHINGTON,D.C.20555 0 ~4o +°`tea October 25, 1985 Docket No. 50-267 Mr. 0. R. Lee, Vice President Electric Production N Z 5 1985 Public Service Company of Colorado iii P. 0. Box 840 J� COL-O. Denver, Colorado 80201 ::: EELEX. Dear Mr. Lee: Re: Recentralization of Licensing Functions for the Fort St. Vrain Nuclear Generating Station As you are aware, the retransfer of Fort St. Vrain licensing activities to NRR was effective October 4, 1985. A Federal Register Notice to amend 10 CFR 50.4 and implement this change was published on October 9, 1985. A copy is enclosed for your information. The licensing project manager for Fort St. Vrain is Kenneth L. Heitner, Operating Reactors Branch No. 3 (ORB No. 3). He will report to Edward J. Butcher, Acting Chief, ORB No. 3. Sincerely, pit . Lainas, Assistant Director , for Operating Reactors Division of Licensing Enclosure: As stated cc w/enclosure: See next page Mr. 0. R. Lee Public Service Company of Colorado Fort St. Vrain cc: C. K. Millen Albert J. Hazle, Director Senior Vice President Radiation Control Division Public Service Company 4210 East 11th Avenue of Colorado Denver, Colorado 80220 P. O. Box 840 Denver, Colorado 80201 J. W. Gahm Nuclear Production Manager Mr. David Alberstein, 14/159A Public Service Company of Colorado GA Technologies, Inc. P. 0. Box 368 P. 0. Box 840 Platteville, Colorado 80651 Denver, Colorado 80201 J. K. Fuller, Vice President Public Service Company of Colorado P. 0. Box 840 Denver, Colorado 80201 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. 0.Box 640 Platteville, Colorado 80651 Kelley, Stansfield & O'Donnell Public Service Company Building Room 900 550 15th Street Denver, Colorado 80202 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1800 Lincoln Street Denver, Colorado 80651 (7590-01) NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 Regional Licensing Program Fort St. Vrain Nuclear Generating Station AGENCY: Nuclear Regulatory Commission. ACTION: Final Rule. SUMMARY: The Nuclear Regulatory Commission is amending Part 50 of its regulations concerning the domestic licensing of utilization facilities to provide information concerning the NRC's regional licensing program. This amendment states that authority and responsibility for implementing NRC's nuclear reactor licensing program pertaining to the Fort St. Vrain Nuclear Generating Station will be carried out by the Director of Nuclear Reactor Regulation and specifies where communications and applications relating to that facility should be sent. The amendment is necessary to inform the licensee and the public of current NRC practice and organization. EFFECTIVE DATE: October 4, 1985 FOR FURTHER INFORMATION CONTACT: Hugh L. Thompson, Jr, Director, Division of Licensing, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555. Telephone (301) 492-9595. SUPPLEMENTARY INFORMATION: The Commission has completed a pilot regionalization program to demonstrate that specific reactor licensing activities can be effectively accomplished in NRC regional offices. Since December 1, 1982, - 2 - certain licensing activities for the Fort St. Vrain Nuclear Generating Station (Utility Licensee: Public Service Company of Colorado, License No. DPR-34, Docket No. 50-267) have been carried out by NRC's Region IV (RIV) as part of the pilot regionalization program. However, regionalization of the reactor licensing function will not occur in the near future. Consequently, all licensing responsibilities for the Fort St. Vrain Nuclear Generating Station should be returned to the Office of Nuclear Reactor Regulation. The delegation of authority to the Regional Administrator of NRC's Region IV will be rescinded effective October 4, 1985. Copies of the memorandum effecting the recentralization of Fort St. Vrain licensing responsibilities have been placed in the Commission's Public Document Rooms at 1717 H Street, N.W. , Washington, D.C. , at the RIV Office, 611 Ryan Plaza Drive, Suite 1000, Arlington, Texas, and at the Greeley Public Library, City Complex Building, Greeley, Colorado 80631 (the local public document room for the Fort St. Vrain Nuclear Generating Station) where they are available for inspection and copying by the public. This amendment to 10 CFR 50.4, is necessary to inform licensees and the public of current NRC practices and organization. As amended, section 50.4 requires that inquiries concerning NRC regulation of all types of production and utilization facilities, including the Fort St. Vrain Nuclear Generating Station, be sent to the Director of Nuclear Reactor Regulation and specifies the proper address. The amendment deletes subparagraph (c) and references to subparagraph (c). The amendment does not change the requirements for direct communication between the licensee and RIV. Since this amendment is nonsubstantive and relates to matters of agency organization and procedure, the - 3 - notice and comment procedures of the Administrative Procedure Act (5 U.S.C. 553) do not apply and good cause exists for making the amendment -effective on October 4, 1985. ENVIRONMENTAL IMPACT: CATEGORICAL EXCLUSION The NRC has determined that this final rule is the type of action described in categorical exclusion 10 CFR 51.22(c)(3). Therefore, neither an environmental impact statement nor an environmental assessment has been prepared for this final rule. PAPERWORK REDUCTION ACT STATEMENT This final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.L. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget approval number 3150-0011. LIST OF SUBJECTS IN 10 CFR PART 50 Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. - 4 - For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is adopting the following amendment to 10 CFR Part 50. PART 50 - DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES 1. The authority citation for Part 50 continues to read as follows: Authority: Sec. 161, as amended (42 U.S.C. 2201); sec. 210, as amended 42 U.S.C. 5841). 2. Section 50.4 is revised to read as follows: § 50.4 Communications. (a) Except where otherwise specified, any communication or report concerning the regulations in this part and any application filed under these regulations may be submitted to the Commission as follows: (1) By mail addressed to - Director of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555. (2) By delivery in person to the Commission offices at - (1) 1717 H Street, N.W. Washington, D.C.; or (ii ) 7920 Norfolk Avenue, Bethesda, Maryland. - 5 - (b) Before making any submittal in microform, the applicant or licensee shall contact the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Telephone (301) 492-8585, to obtain specifications and copy requirements. Dated at Bethesda, Maryland, this 3 d day of Jr 1985. For The Nuclear Regulatory Commission. i lam . ircks, Executive Director for Operations. °`app RECL, %)0‘' UNITED STATES tu NUCLEAR REGULATORY COMMISSION • m ; WASHINGTON,D.C.20555 °� November 1, 1985 Docket No. 50-267 Nov 15 'ass J Mr. 0. R. Lee, Vice President Electric Production aREacY, Public Service Company of Colorado P. 0. Box 840 Denver, Colorado 80201 SUBJECT: FORT ST. VRAIN - APPENDIX R SAFE SHUTDOWN MODELS - REQUEST FOR ADDITIONAL INFORMATION Dear Mr. Lee: We are reviewing your submittals dated November 16, and December 17, 1984, January 17 and April 1, 1985 concerning your fire protection evaluation of Fort St. Vrain. One portion of our review concerns the systems which are being relied upon to provide safe shutdown (Trains A and B for noncongested cable area fires and the ACM components for congested cable area fires). Our initial review disclosed certain deficiencies in the submitted information which were discussed with your staff during a telephone conference on October 17, 1985. The questions raised during this conference call and questions related to manual operations and instrumentation are listed in the enclosure to this letter. Additional questions may be developed as a result of our continued review or from your responses to these initial questions. We request that you provide a response to these questions within 30 days of your receipt of this letter. The information requested in this letter affects fewer than 10 respondents; therefore, OMB clearance is not required under P.L. 96-511. Sincerely, Edward J. Butcher, Acting Chief Division of Licensing Office of Nuclear Reactor Regulation Enclosure: Fort St. Vrain Fire Protection Review Questions cc w/enclosure: See next page Mr. 0. R. Lee Public Service Company of Colorado Fort St. Vrain cc: C. K. Millen Albert J. Hazle, Director Senior Vice President Radiation Control Division Public Service Company 4210 East 11th Avenue of Colorado Denver, Colorado 80220 P. 0. Box 840 Denver, Colorado 80201 J. W. Gahm Nuclear Production Manager Mr. David Alberstein, 14/159A Public Service Company of Colorado GA Technologies, Inc. P. 0. Box 368 P. 0. Box 840 Platteville, Colorado 80651 Denver, Colorado 80201 J. K. Fuller, Vice President Public Service Company of Colorado P. O. Box 840 Denver, Colorado 80201 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. 0.Box 640 Platteville, Colorado 80651 Kelley, Stansfield & O'Donnell Public Service Company Building Room 900 550 15th Street Denver, Colorado 80202 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1800 Lincoln Street Denver, Colorado 80651 • • Enclosure Fort St. Vrdin Fire Protection Review Questions . 1. There are situations in which a single failure can defeat a function common to both trains proposed for forced circulation safe shutdown. a. The service water strainer (F-4201) and the associated flow valves (HV-4257, HV-4225, HV-4221-1, and HV-4221-3) are required to be operated in both proposed trains. What provisions have been made to provide for continued system function if a fire disables or precludes positioning of these components? b. Have all other situations where single failures could disable the function been evaluated? 2. Review of the Tables contained in Section 2.1 of Report 1 (as revised) shows that not all of the required valves are listed. a. What criteria was used to determine if a component should be included in the Table? (e.g. , Table 2.1-3, sheet 3 of 5 lists V75595 to be closed but does not indicate that parallel valves V75600 and V75605 would need to also be closed; also, other valves off the same line would need to be properly positioned but are'.not listed.) b. Do the implementing procedures require checking the position of all valves in the flow path? (e.g. , A valve in the flow path that is designed to fail close and is required to be close should be verified to be closed; HV-2223 should be verified closed.) 3. The condensate storage tank is common to both safe shutdown trains. a. What is the minimum storage capacity necessary to complete the assumed function(s). b. This minimum capacity should be incorporated as a Technical Specification limit, as should all equipment for which credit is taken in the shutdown models. 4. The shutdown models for forced circulation cooldown, for a fire in a noncongested cable area, do not contain provisions for maintaining PCRV liner cooling. In light of the acceptance criteria B.2.b. contained in the PSC letter dated August 17, 1984, provide an evaluation of the necessity of maintaining liner cooling. 5. Provide an evaluation of the need of a main cooling tower fan in safe shutdown model Train B. 6. A list of required ACM fire protection shutdown components is provided in Table 3.2 of Report 1. -2- a. Provide a listing of the equipment identification numbers and the normal power supplies for these components. b. Provide a description of all changes made to the "ACM shutdown system" since its use was approved in License Amendments Nos. 14, 18, and 21. 7. Drawing PI 31-1 shows a removable spoolpiece between the fire water system and the line from the main feedwater pumps. a. Is this spoolpiece normally installed? If not, describe how its storage and installation are controlled. (In similar applications at other facilities, the spoolpiece is a Technical Specification controlled item.) b. Describe all other spoolpieces which are being used in the shutdown models and the controls used. 8. Since provisions are not included in the shutdown models for its operation, provide an evaluation of the necessity of the Buffer Hel'tum System for circulator operation. 9. The shutdown models for forced circulation cooldown consist of providing cooling water to the economizer-evaporator-superheater section of one of the steam generators via a low-pressure pump (condensate or fire water). Since the steam generators will be at relatively high temperatures (greater than 1000°F at shutdown) and are helical wound tubes with no storage capacity, provide an evaluation of the effectiveness of this mode of cooldown. This evaluation should include a study of the possibility of damage, caused by occurrences such as water hammer or over pressurization, during such a cooldown. 10. The capacity of the diesel firewater pump is indicated on drawing PI-45 to be 1500 gpm; Figure 2.1-9 of Report 1 indicates that this pump will provide 155 gpm for a Helium Circulator and 1050 gpm for a steam generator. Provide an evaluation which verifies that adequate capacity • is available to perform the safe shutdown functions in addition to providing sufficient fire suppression water flow. 11. Provide a description of the testing program which will be implemented to verify the operability of the proposed safe shutdown models. 12. There are numerous operations contained in the Tables of Section 2. 1 of Report 1 which require the operator to "De-energize and open or close" manually (HV-3133-1 and HV-3133-2 in Table 2.1-3) or "Remove power and open or close" locally (HV-4221-2 of Table 2.1-6). Operations such as removing fuses and, in most cases, opening power supply circuit breakers are considered repair operations and are not allowable for the train required for hot shutdown (III.6.1). • _3_ • Provide a description of what actions are necessary to accomplish the operations indicated in the Tables of Section 2.1 along with an evaluation of the time required to perform these actions. 13. Provide an evaluation that the Technical Specification required, onshift, crew size is sufficient to perform the actions proposed for the various shutdown models without reliance on Fire Brigade members. 14. It is the staff's position that monitoring of core flux provides the only direct indication of the reactor shutdown condition and therefore provisions for postfire source range flux monitoring are necessary to meet Section III.L.2 of Appendix R and, therefore, A.3.a. and B.2.b. of the PSC letter dated August 17, 1984. The position stated in Section 2.3 of Report 1, that, since a fire cannot credibly prevent control rod insertion, neutron flux monitoring is not required, has not been adequately justified. Include provisions in all shutdown models far source range (startup channels) monitoring or provide an evaluation of alternatives available to monitor core reactivity conditions. 15. Provide an explanation of how the reactor decay heat removal function will be monitored following a fire in a noncongested cable area (Criteria ' B.2.c. and d. ) without maintaining Primary Helium temperature instrumentation operable. 16. Provide an explanation of how the reactor pressure control function will be monitored for a fire in a congested cable area (Criteria A.3.b. and d. ) without maintaining Primary Helium pressure instrumentation operable. MEETING SUMMARY DISTRIBUTION Licensee: Public Service Company of Colorado *Copies also sent to those people on service (cc) list for subject plant(s). Docket File NRC PDR L PDR ORB#3 Rdg KHeitner te' EButcher BGrimes N0V 151985 OELDJ EJordan, IE V '--� ACRS-10 CzREELEY. COLO. PMorriette NRC Meeting Participants: GLainas TKing RLaGrange PShemanski AMasciantonio JWermei1 NWagner Docket No. 50-267 October 31, 1985 MEMORANDUM FOR: Edward J. Butcher, Acting Chief Operating Reactors Branch No. 3, DL FROM: Kenneth L. Heitner, Project Manager Operating Reactors Branch No. 3, DL SUBJECT: SUMMARY OF MEETING WITH PUBLIC SERVICE COMPANY OF COLORADO (PSC) TO DISCUSS SCHEDULE EXTENSION TO EQUIPMENT QUALIFICATION RULE (10 CFR 50.49) FOR FORT ST. VRAIN The purpose of the meeting was for PSC to make a presentation to the staff on their extension request. It also afforded PSC an opportunity to discuss with the staff their equipment qualification program, and the new steam line rupture detection and isolation system (SLRDIS). Attendees at this meeting are listed in Enclosure 1. The material used by the licensee in their presentations is in Enclosure 2. During these discussions, certain issues were noted for further attention by the staff and the licensee. These included: - The aging methodology being used to qualify the equipment. - The adequacy of quality assurance in PSC's EQ program. - The spectrum of helium releases from the reactor or other pressurized helium lines should be considered by the PSC EQ program. - The environments that could be created by small steam leaks that would not immediately be isolated by the SLRDIS. These environments could affect EQ or operator access following an accident. The staff noted that additional information would be requested concerning these issues that affected both the extension request and the entire EQ program. These questions would also cover: - SLRDIS and, - Models used for determination of the accident environment. The staff stated it would send a request for additional information to PSC shortly on these issues. Kenneth L. Heitner, Project Manager Operating Reactors Branch No. 3, DL Enclosures: As stated • cc w/enclosures: See next page ORB#3:DL6J ±1 KHeitner;ef to /31/85 ATTENDEES AT OCTOBER 29, 1985 MEETING WITH PUBLIC, SERVICE COMPANY OF COLORADO ON EQUIPMENT QUALIFICATION AT FORT ST. VRAIN Name Organization . Ken Meitner USNRC r E. Butcher USNRC G. Lainas USNRC T. King USNRC R. LaGrange USNRC P. Shemanski USNRC A. Masciantonio USNRC J. Wermiel USNRC N. Wagner USNRC S. Ball ORNL M. Harrington ORNL 0. Lee PSC L. Brey PSC M. Holmes PSC D. Warembourg PSC Mr. 0. R. Lee Public Service Company of Colorado Fort St. Vrain cc: C. K. Millen Albert J. Hazie, Director Senior Vice President Radiation Control Division Public Service Company 4210 East 11th Avenue of Colorado Denver, Colorado 80220 P. 0. Box 840 Denver, Colorado 80201 J. W. Gahm Nuclear Production Manager Mr. David Alberstein, 14/159A Public Service Company of Colorado GA Technologies, Inc. P. 0. Box 368 P. 0. Box 840 Platteville, Colorado 80651 Denver, Colorado 80201 J. K. Fuller, Vice President Public Service Company of Colorado P. 0. Box 840 Denver, Colorado 80201 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. 0.Box 640 Platteville, Colorado 80651 Kelley, Stansfield & O'Donnell Public Service Company Building Room 900 550 15th Street Denver, Colorado 80202 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1800 Lincoln Street Denver, Colorado 80651 Enclosure 2 IQ MEETING AGENDA OCTOBER 29, 1985 METING PSC - ERR 1.0 INTRODUCTORY REMARKS - NPR PSC 2.0 SEPTEMBER 24, 1985 REQUEST FOR SCHEDULAR EXTENSION - OVERVIEW, EXCEPTIONAL CIRCUMSTANCES - RELATIONSHIP OF TECHNICAL ISSUES - PSC POSITION FOR OPERATION BEYOND NOVEMBER 30, 1985 - a - SCHEDULAR EXTENSION - CLARIFICATION NRC QUESTION 3.0 QUESTIONS/DISCUSSION 4.0 ENGINEERING EVALUATION - TEMPERATURE PROFILES - SLRDIS - OPERATION BEYOND NOVEMBER 30, 1985• (BRIEF ENGINEERING POSITION) 5.0 SUMMARY/CLOSING REMARKS s EXCEPTIONAL CIRCUMSTANCES • FORT ST. VRAIN IS A HIGH TEMPERATURE GAS-COOLED REACTOR, THE ONLY ONE OF ITS KIND IN THE USA. - THE MAJOR POINT IS THAT OUR UNIQUE CHARACTERISTICS DID NOT PERMIT PSC TO BENEFIT FROM INFORMATION SHARING AND GUIDANCE AVAILABLE IN THE LIGHT WATER REACTOR INDUSTRY THROUGH ESTABLISHED OWNER'S GROUPS AND NRC INTERFACES - PSC RECOGNIZED FSV'S UNIQUE HARSH ENVIRONMENTAL CONDITIONS AND PROCEEDED TO DEVELOP AN EQ PROGRAM AND ASSOCIATED EQUIPMENT TEST PROGRAM TO QUALIFY EQUIPMENT TO THESE UNIQUE CONDITIONS j= - IN - THIS RESPECT, FSV'S UNIQUENESS BY ITSELF DID NOT DIRECTLY CAUSE EQ PROGRAM DELAYS, BUT INDIRECTLY THE UNIQUENESS WHICH. TENDED TO ISOLATE FSV FROM INFORMATION AND GUIDANCE LED TO THE OVERALL EQ PROGRAM SHORTCOMINGS THAT WERE RECENTLY IDENTIFIED s EXCEPTIONAL CIRCUMSTANCES • THE EQ RULE (10CFR50.49) DOES NOT RECOGNIZE THE UNIQUE 11 CHARACTERISTICS OF AN HTGR. THERE WAS ESSENTIALLY NO HTGR, EQ 1 GUIDANCE AVAILABLE WITHIN THE NUCLEAR INDUSTRY - PSC WAS CHARGED BY AN NRC ORDER TO DEVELOP AN EQ PROGRAM UTILIZING DOR GUIDELINES TO THE EXTENT APPLICABLE TO AN HTGR - PSC ACCEPTED THIS CHARGE AND PROCEEDED TO DEVELOP AN EQ PROGRAM BASED UPON CERTAIN ASSUMPTIONS. THESE ASSUMPTIONS AS WELL AS VARIOUS PSC POSITIONS WERE SUBMITTED TO THE NRC. PSC RESPONDED IN A TIMELY FASHION TO THE VARIOUS BULLETINS, ORDERS AND REGULATIONS a NRC STAFF - WAS HEAVILY INVOLVED IN LWR EQ PROGRAM DEVELOPMENT. THE FSV EQ PROGRAM WAS DEVELOPED WITHOUT THE BENEFIT OF SIMILAR INVOLVEMENT OR FORMAL NRC REVIEWS z 0 UNIQUE CHARACTERISTICS - NO CONTAINMENT - NO RADIATION EFFECTS - EMERGENCY RESPONSE SYSTEMS ARE IN OPERATION FOR NORMAL PLANT OPERATION - REACTOR CONCEPT UTILIZES A MULTITUDE OF EQUIPMENT TO EFFECT SAFE SHUTDOWN WITH SEVERAL SHUTDOWN METHODS 9O m inu4c3- 4-o - REACTOR CONCEPT ALLOWS TIME FOR OPERATOR ACTION ' �' ►�3t}c IY Ycr4W{ liner - NO DESIGN BASIS EVENTS RESULTS IN OFFSITE RADIOLOGICAL cowl. CONSEQUENCES THAT APPROACH 10 CFR 100 GUIDELINES - REACTOR CONCEPT ALLOWS FOR PERMANENT LOSS OF FORCED CIRCULATION WITH OFFSITE RADIOLOGICAL CONSEQUENCES WELL BELOW - - 10 CFR 100 GUIDELINES t - HIGH TEMPERATURE/HIGH PRESSURE STEAM CONDITIONS RESULT IN EXTREMELY HARSH ENVIRONMENTS • • EXCEPTIONAL CIRCUMSTANCES • RECENT NRC STAFF GUIDANCE HAS BEEN HELPFUL. BUT THE FSV PROGRAM HAS NEVER BEEN GIVEN A COMPREHENSIVE REVIEW WITH RESULTING SER OR TER - PSC BELIEVES THAT MEANINGFUL AND CONSTRUCTIVE INTERACTIONS BETWEEN PSC AND THE NRC IN TERMS OF SER'S OR TER'S COULD HAVE RESULTED IN TIMELY DEVELOPMENT OF AN FSV EQ PROGRAM IN ACCORDANCE WITH 10CFR50.49 - IT IS EVIDENT THAT ONCE CONSTRUCTIVE INTERCHANGE WAS ESTABLISHED IN EARLY 1985, PSC HAS TAKEN AN AGGRESSIVE APPROACH TO RESOLVE ALL TECHNICAL ISSUES AND IDENTIFIED CONCERNS - _ a EXCEPTIONAL CIRCUMSTANCES • BASIC ASSUMPTION AND LICENSING BASIS OF FSV EQ PROGRAM HAS BEEN UTILIZATION OF A FOUR MINUTE OPERATOR RESPONSE TO ISOLATE AN HELB. CURRENT NRC POLICY SETS FORTH A MINIMUM OF TEN MINpTES FOR OPERATOR RESPONSE - FSV HIGH TEMPERATURE/PRESSURE STEAM CONDITIONS ARE SUCH THAT AN EQ PROGRAM CANNOT BE DEVELOPED FOR A TEN MINUTE OPERATOR RESPONSE - RESOLUTION OF THIS ISSUE (INSTALLATION OF A STEAM LINE RUPTURE DETECTION/ISOLATION SYSTEM) HAS A BROAD OVERALL IMPACT ON THE EQ PROGRAM - ALTHOUGH NOT SPECIFICALLY STATED IN THE SEPTEMBER 24, 1985 r SCHEDULAR EXTENSION REQUEST, THE FOUR MINUTE VERSUS TEN _ MINUTE OPERATOR RESPONSE CANNOT BE DIVORCED FROM THE OTHER TECHNICAL ISSUES RAISED BY THE NRC IN JANUARY 1985 (I.E. AGING, OPERABILITY TIME, EQ FILES, ETC.) - THE GENERAL RELATIONSHIPS AND INTERDEPENDENCY OF OPERATOR RESPONSE TIME, AGING AND OPERABILITY TIME ARE REPRESENTED ON THE FOLLOWING TWO TABLES s o e J z • y !El jai �.9od are s I WW ! r fu y W ti sties e: Li s O WNJ� O 6 W1O J :j i »:O: C w . W lute _1 1 ma -it g 595 so' WI $SW dIgsE l; w \s-Lb T — 0 o I II er on lv to. h he tW y g' f�4 > O Wo >II q =1! ii Il IW = iCH } rr y�i W6a = < Ws< wg1 {�W �� yW�: ~ `at~ t� 2WO��o7. W W Z ~� Zs aO !WI i ~j l <ef < 111 Ii Ill to 1°�s W t lI il_ iigi ig ile sl§ so _e nt_� ill< W . t 3 f W liii le ! lII ill! $ W ii G W 2 -! <Z » I g� ■ I C • Cs 1 V-1 ■wt`§ I § Iii mc to §to ■$ , % i |§ a§ fa■ so Jr se$ !k7 $ § ril 'z or sg �m / `§ tin $E, §2§ IE §2§ §&s 1k §§ ■ ■ . P \ . § § % - § § g C - ■ 0 2 § I - § .s§ $ § & . ■ � §§ § f §2 §§U w . 4 - B E 9 � $ �N1141 $E § °Ili :! ' < 8 § - ■ /ND, k§I § k &cU■ to 21 so 8 §2; • TASKS BtING PERFORMED TO SUPPORT THE FSV EQ PROGRAM • rn c s c 0 o+ c O r V 31 9 O Y i+ '- I- 4-3 O r CO A C.- O ]•0_n N'C .O N1O V L.- d 'O CAC L N I- 6 W. O C EQ RELATED TASKS a LL 4 a o F c. IF) ENGINEERING EVALUATIONS/ANALYSIS EE-EQ-001 I X I FSV EQ Functional Descriptions EE-EQ-002 I X I FSV Component Operating Cycles EE-EQ-003 XCIX Interaction and Effects of Non- Qualified Comp. in Circuits with Qualified SSD Comp. EE-EQ-004 XCIX X Interaction and Effects of Non- Qualified Comp. in Circuits with Qualified PPS Comp. EE-EQ-005 XCII Acceptability of Seismic Type D Valves in a Steam Break Environment EE-EQ-006 X C X Effect of, Fire Suppression System Actuation on Safe Shutdown Equipment EE-EQ-007 XCIX Interaction and Effects on Non- Qualified Comp. in Indicator Light Circuits with Qualified SSD Comp. EE-EQ-008 XCIXCC FSV Brake and Seal System Solenoid Valve Qualification EE-EQ-009 CIXCC Liner Cooling Method for SSD Cooling at FSV Following High Energy Line ' Break ' EE-EQ-010 X X Potential Failure of Components Subjected to HELB Environment EE-EQ-011 X C I C C Systems Review Forced Circulation • Cooling TASKS BEING PERFORMED TO SUPPORT THE FSV EQ PROGRAM C+ CC O 0 r V A + W ▪ 0 S N •- L �u a Cr- 0 � 7a+�n N t a �� L.- N CU v eta L W.- -. W N C n 0 z EQ RELATED TASKS h LL a o Z -. ENGINEERING EVALUATIONS/STUDIES EE-EQ-013 X I Accidents Resulting in Harsh Environment Areas at Fort St. Vrain EE-21-011 X C Submergence Concerns on BWST Level - Valves - Provide HELB Profiles for Manual I I I X Termination of Feedwater and Condensate Breaks Evaluate Other Trip Methods for I I X I Feedwater Condensate and Extraction Steam HELBS Provide HELB Profiles Utilizing I IIXI Variable Heat Transfer Coefficients Prepare Report on Slowdown Rates and X Building Air Temperature Response Evaluation - Reactor Building Louvers C X and Exhaust Filters Revise CONTEMPT G to Provide Humidity I X I Calculation Identify Control Functions Taken X X X X Credit for in the HELB Scenarios Prepare HELB Systems Review Document I I X I Including Single Failures Investigate Recalculation of MCA Si I I I X DBA 2 Profiles with CONTEMPT G Taking Credit for Heat Sinks Revise CONTEMPT G to Provide Harsh I I I X Environment Profiles for East 4A Wall, Turbine Deck and Refueling Floor • TASKS BEING PERFORMED TO SUPPORT THE FSV EQ PROGRAM rn c rn C 0r v G r d>, o 70 . g U - Z v RS.", 0 N W y p a L. in OJ C FL' L 61r - w ^i1 Yoil ' BEL EQ RELATED TASKS `� LL WIti r. c o J - d N ENGINEERING EVALUATIONS/STUDIES Investigate Min./Max. Human Tolerance I I X C for Personnel Entering Harsh Environments Evaluate Automation of all Manual C X Valves Associated with Liner Cooling Evaluate Automation- of all Valves X C Associated with Forced Cooling FIELD WALKDOWNS/REPAIRS Perform Walkdown of Safe Shutdown IXII Items Identify or Replace ASCO Solenoids IXII without Nameplates Review Field Walkdown Sheets IXII Prepare Report on Model I Discrepancies Review Work Completed/or in Progress I X I I Since February 1983 Install New Splices IXII Perform Walkdown of Cables in EQ I X I I Program as Modified by Cable P.O. Review Repair Items Identified by Walkdown IXII TASKS BEING PERFORMED TO SUPPORT THE FSV EQ PROGRAM rn c C ;o.- V a, .- CU 'CO Y 40 '- L 44 O I-- 'C r 0 _ 7 L U = Mr Li gip N C L r N u cI. S ie ow"— O.P c a A O -4- 01 C � . E O C EQ RELATED TASKS La_ IA- C C r F d`, v) DESIGN CHANGES CN Package - Hard Pipe Spool Pieces X C X and Provide Firewater Piping to HTFA, CN-2142 CN Package - Protective Equipment X IXCC Changes to Insure Proper Coordination Between SR and NSR Loads, CN-2119 r Replacement of Brake and Seal Valve X IXCC Seats and Seals, CN-2178 Prepare System Description for Steam X Line Rupture Detection/Isolation System (SLRDIS) ( Tagging of Components (Indicating X X Lights) not Previously Tagged, CN-2184 SLRDIS Modifications for E-H Valves. I IICX (PV-2229, 2230: PV-2243, 2244), CN-2180 PPS Modifications for SLRDIS, CN-2176 X • SLRDIS Modifications for Air Operated I C I I C X and Hydraulic Valves (18 Valves), a CN-2181 Cable Re-routing for Components X C X Affected by EQ Program Changes, CN-2182 CN Package - Move TR 8201, TR 8202, I C X I C C and TR 8203 to Mild Environment, CN-2151 CN Package - Replace GE Relays with ICXICC Qualified Devices, CN-2095 • CN Package - Replace United Electric I C X C C Temperature Switches with Qualified Devices, CN-2160 CN Package - Replace Foxboro TE's with I C X C C Qualified Devices, CN-2144 TASKS BEING PERFORMED TO SUPPORT THE FSV EQ PROGRAM • C C C C s o O. 7gg •g- '0 O Y g1+ 0- L i+ O J 'O `r O 7 10 au any C L � d U 9 _ -- C L d 0.4- p tit Cl r d C a E O 6' H IL !i Q O ..f 0O ell- N EQ RELATED TASKS DESIGN CHANGES (CONTINUED) CN Package - Replace ASCO's for I C X C C FV-21297 and FV-21298 with Qualified Devices, CN-2152 — . DATA BASE UPDATING - EQ Data Base Update-for Forced XCCX Circulation Cooling, Ten Change Notices Overall EQ Data Base Update, CN-2172 X I X. Preparation of EQ Books ICXICC Prepare Detailed List of Equipment X X X Required for Isolation of High Energy Line Break and Equipment Required for Depressurization and Liner Cooling Identify EQ Cable Numbers I X I I Provide Necessary EQ Documentation CCXCC for Liner Cooling Comp. MISCELLANEOUS Transmit Voltage Ranges for EQ X Electrical Equipment to S&L Environmentally Test Fuse Blocks X IXCC for CN-2119 LEGEND --X—:—Primary initiating areas/functions I - Areas/functions impacted C - Areas/functions needed for completion/input JUSTIFICATION FOR CONTINUED OPERATION O INSTALL SLRDIS O UTILIZE QUALIFIED REDUNDANT LINER COOLING PATHS ALONG WITH ASSOCIATED SYSTEMS TO PROVIDE REACTOR COOLING O PSC CONTENDS THAT INSTALLATION OF SLRDIS ALONG WITH QUAcIFICATION OF THE LINER COOLING FLOW PATHS AND ASSOCIATED SYSTEMS .WILL ESTABLISH NECESSARY CONDITIONS TO PROTECT THE HEALTH AND SAFETY OF THE PUBLIC AND MEET THE INTENT OF 10 CFR 50.49 FOR INTERIM OPERATION (NOVEMBER 30, 1985 THRU MARCH 31, 1986) IN THAT: - THE INTEGRITY OF THE REACTOR COOLANT PRESSURE BOUNDARY CAN BE MAINTAINED - THE REACTOR CAN BE SHUTDOWN AND MAINTAINED IN A SHUTDOWN CONDITION a c - THE CONSEQUENCES OF REACTOR COOLING UTILIZING THE PCRV LINER COOLING PATHS HAVE BEEN ANALYZED, AND IT HAS BEEN SHOWN THAT OFFSITE EXPOSURES REMAIN WELL BELOW 10 CFR 100 GUIDELINES O THE TECHNICAL ASPECTS AND SUPPORT FOR THE ABOVE POSITION WILL BE ADDRESSED BY ENGINEERING EVALUATIONS AND ASSOCIATED SAFETY EVALUATIONS FOR SLRDIS AND LINER COOLING TO BE SUBMITTED FOR NRC REVIEW IN EARLY NOVEMBER, 1985 • EQ SCHEDULAR EXTENSION - PSC IS PROCEEDING WITH AN AGGRESSIVE EFFORT TO QUALIFY ALL ELECTRICAL EQUIPMENT NECESSARY FOR SAFE SHUTDOWN UTILIZING BOTH THE LINER COOLING METHOD AS WELL AS FORCED CIRCULATION - PSC IS CONCENTRATING INITIAL EFFORTS ON EQUIPMENT-NEEDED FOR SLRDIS AND LINER COOLING UTILIZING FIRE WATER, CURRENTLY THIS SET OF EQUIPMENT INCLUDES 51 EQUIPMENT ITEMS COMPRISING 438 COMPONENTS - PARALLEL EFFORTS ARE IN PROGRESS FOR QUALIFICATION OF ELECTRICAL EQUIPMENT NEEDED FOR FORCED CIRCULATION COOLING. THIS EQUIPMENT SET PRESENTLY INCLUDES 156 ITEMS REPRESENTING 1,246 COMPONENTS - PSC IS CONFIDENT THAT THE MAJORITY OF EQUIPMENT IN THE OVERALL EQ PROGRAM WILL BE QUALIFIED IN THE NOVEMBER 30, 1985 OR PRIOR - - TO REACTOR RESTART TIME FRAME INCLUDING REPLACEMENT OF AGE D SENSITIVE MATERIALS/COMPONENTS, REPAIR OF FIELD INSTALLATION DEFICIENCIES AND REPLACEMENT OF TAPED SPLICES - IT SHOULD BE EVIDENT, HOWEVER, THAT PSC IS INVOLVED IN A MULTI-FACETED EFFORT WITH MANY INTERDEPENDENT ACTIVITIES PROCEEDING IN PARALLEL. SIGNIFICANT PROBLEMS THAT COULD DEVELOP IN ANY ONE AREA COULD AFFECT MANY OTHER AREAS AND COULD HAVE NEGATIVE IMPACT ON THE OVERALL EQ PROGRAM SCHEDULES. PSC HAS EVALUATED THE CURRENT STATUS OF EVALUATIONS/STUDIES AND OVERALL EQ PROGRAM WORK ACTIVITIES. PSC BELIEVES THAT NECESSARY ACTIONS REQUIRED TO ESTABLISH A DEFENSE IN-DEPTH EQ PROGRAM CAN BE ACCOMPLISHED WITHIN THE MARCH 31, 1986 SCHEDULAR EXTENSIONS REQUEST. a PERSONNEL CURRENTLY WORKING ON EQ PSC AND ON-SITE CONSULTANTS FOR ALL PSC DIVISIONS (PRODUCTION, ENGINEERING, LICENSING, QA) , EQ IS TOP PRIORITY AND USES MAJORITY OF WORK EFFORT. PSC HAS CONTRACTED NUMEROUS CONSULTANTS TO WORK ON SITE TO SUPPLEMENT PSC STAFF WORKING ON EQ OFF-SITE CONSULTANTS E DI BENEDETTO ASSOCIATES GA TECHNOLOGIES INC GENERAL ELECTRIC IMPELL PROTO-POWER CORPORATION SARGENT & LUNDY STONE & WEBSTER TENERA WESTINGHOUSE WYLE TOTAL THE ABOVE WORK FORCE REPRESENTS THE EQUIVALENT OF OVER 400 PEOPLE WORKING FULL TIME ON EQ NRC QUESTION #1 0 WHAT IS THE CAPABILITY OF THE REACTOR BUILDING CONFINEMENT AND LOUVERS TO WITHSTAND A STEAM LINE BREAK AND STILL FUNCTION EFFECTIVELY TO REDUCE THE RADIOLOGICAL CONSEQUENCES OF THE DESIGN BASIS EVENT? • IN TERMS OF OFFSITE RADIOLOGICAL CONSEQUENCES, THE VARIOUS DESIGN BASIS EVENTS AS CONTAINED IN THE FSV FSAR DW-NOT TAKE CREDIT FOR THE REACTOR BUILDING LOUVER SYSTEM • THE REACTOR BUILDING LOUVER SYSTEM AS IT CURRENTLY EXISTS IS n NOT ENVIRONMENTALLY QUALIFIED • EVALUATIONS ARE UNDERWAY FOR THE FOLLOWING AVENUES OF APPROACH: - DETERMINE THE END EFFECT OF SLRDIS AND THE NEED TO QUALIFY -F4r(r THE LOUVER SYSTEM a, • QUALIFY THE REACTOR BUILDING LOUVER SYSTEM WITHIN THE AUSPICES OF FSV'S EQ PROGRAM - PERFORM EVALUATIONS AND ANALYSES TO SUPPORT NOT UTILIZING • OR DEPENDING ON LOUVER SYSTEM ACTUATION NRC QUESTION #2 0 WHAT IS THE EFFECT OF MOISTURE RELEASED BY THE BREAK ON THE EFFECTIVENESS OF THE REACTOR BUILDING EXHAUST FILTERS? - PRELIMINARY INVESTIGATIONS INDICATE THAT THE PERFORMANCE OF THE EXHAUST FILTERS WILL NOT BE SIGNIFICANTLY REDUCED BY THE STEAM LEAK CONDITIONS IN TERMS OF EFFICIENCY FOR REMOVAL OF ELEMENTAL IODINE - FILTER EFFICIENCY COULD BE REDUCED FOR ORGANIC HALIDES - PSC IS CURRENTLY EVALUATING THESE AREAS IN TERMS OF RADIOLOGICAL CONSEQUENCES. PRELIMINARY BUILDING TEMPERATURE PROFILES AND RELATIVE HUMIDITY PROFILES HAVE BEEN RECENTLY COMPLETED AND ARE UNDERGOING FINAL REVIEW. THIS INFORMATION WILL BE FACTORED INTO THE EXHAUST FILTER EVALUATIONS TO DETERMINE IMPACT - EVALUATIONS AND ADDITIONAL INFORMATION SHOULD BE AVAILABLE EARLY IN NOVEMBER looks 0r NRC QUESTION !3 • WHAT ARE THE SAFETY BENEFITS FROM OPERATING THE PLANT AT PARTIAL POWER DURING PERIOD OF NOVEMBER 30, 1985 TO MARCH 31, 1986- - FROM THE VIEWPOINT OF RESULTING EQ CONDITIONS FROM AN H!LB, THERE IS ESSENTIALLY NO BENEFIT OPERATING AT PARTIAL POWER LEVELS - PSC HAS NOT PERFORMED ANY DETAILED ANALYSIS TO QUANTIFY REDUCED RADIOLOGICAL CONSEQUENCES DUE TO A LOWER FISSION PRODUCT INVENTORY OF LOWER FUEL TEMPERATURES. SOME RADIOLOGICAL BENEFIT COULD OBVIOUSLY BE EXPERIENCED - THE OAK RIDGE FSV SCENARIO OF OCTOBER 14, 1985 FOR VARIATIONS OF DBA-1 INDICATES FUEL FAILURE WILL OCCUR FROM POWER LEVELS A 35% AND ABOVE. BUT THAT THE PERCENTAGE OF FUEL FAILURE AT 5 - DAYS IS SMALL AND THEREFORE FISSION PRODUCT RELEASE IS SMALLER c Y OAK RIDGE NATIONAL LABORATORY POET OR10E ICE cM Soot n9.SUrs V OWED n ANA MSNEM SOW stria PC October.14, 1985 G at ezefi k 01 ts0s • y[imuvna Mr. T. L. King KIST,T. J. MITOffice of Nuclear Reactor Regulation wI :/- Mail Stop P846 IwvIA IC. COAT. OOL• N U. S. Nuclear Regulatory Commission UIMOTa:+- [[MS.K. Washington, D.C. 20555 9IwL9Gn Mg: w. C. N. - Dear Tom: CIS.K. OtTPLENG.N J. S. ORECA Code Calculations of Fort St. Vrain DBA-1 Scenario Variations MEJOSSM..'A. TEL O.I. ICBM L.N. Can Per your request, I have enclosed results of analyses of Fort St. Vrain WICo9 . I. 10"9010. [. (PSV) DBA-1 scenarios for less-than-full power operation (see Table 1). 90C MOP KOT' The severe accident version of the ORECA-FSV code was used for these BUTS ±:O. S101i'fI.L. calculations. The main object vas to determine the maximum equilibrium mi. u99. M.QGA 1S: W000[1.K.N. power level FSV could run at such that if a permanent loss of forced wuuao. s. circulation (LOFC) event. (DBA-1) did occur, the maximum fuel temperature N1111N&J. would be such that essentially no fuel damage would occur, i.e. T-core • (max) <2900NF or 1600NC. The study showed that this power level is %35% [or 295 MW(t)] . The reference case assumptions for the parameter study were: - 1) Primary system scram, depressurization, and LOFC occur at time - 0. 2) Only one (of the 2) liner cooling systems (LCSs) is opera- tional. 3) The initial core average temperatures assumed are (conser- vatively) derived from ISV normal operating data (see Fig. 1). 4) A time-at-temperature fuel failure model is used to estimate fuel failure fractions*. Sensitivity studies were also done. It was seen from them that the maximum core temperatures were only very slightly less both for the case where both LCS trains are operational and for the case where the system is depressurized according to the emergency operating procedures (EOP- G), i.e., begin a 7—hr depressurization 2 hours after the LOFC. • D. T. Goodin, "Accident Condition Performance of Fuels for RTGRs", J. Amer. Ceramic Soc., 65, 5, May 1982, pp. 238-242. Mr. T. L. King 2 October 14, 1985 Several cases were also run in which it was assumed that no depressurization occurred (and the pressure was assumed to remain constant). Because of the enhanced convection heat transfer to the LCS, the maximum core temperatures are considerably less for corresponding powers for_the depressurized cases. While such a mode of operation is contrary to the current tech specs and not recommended, the analyses indicate that for operation up to MO% power, a DBA-1 scenario in which depressurization vas not achieved would not lead to a situation in which a subsequent rapid depressurization would result in the spread of significant fission products from failed fuel. The DBA-1 scenario accounts for overheating and eventual failure of the PCRC's carbon steel cover plates, especially in the upper plenum region. Noting that the ORECA models used are "approximate", the results indicated that the maximum cover plate temperatures peaked at 1200'F and 1500'F, respectively, for the 202 and 302 power depressurized cases. Hence beyond 302 power, significant cover plate failure would be predicted. Teo other comments on the results presented: 1) The Goodin fuel failure model used vas reviewed by ORNL and judged to be excellent for predicting relatively large failure fractions, but it was not intended as a means for predicting onset-of-failure situations such as we have here. Hence, while we believe that these results are good for-ballpark estimations, they are not accurate. 2) The 1600'C fuel limit is a convenient yardstick, but the Goodin model is better, and shows both that some failure occurs at lower temperatures and that not necessarily much more occurs at the higher temperatures reached, for *staple, in the 402 power depressurized case (Table 1). Note also that the fuel failure fraction is less than the fraction of fission product release to the primary system. The model does not calculate diffusion through the graphite element, etc. Please let us know if you have questions or comments. Yours truly, S. J. Ball, Manager MGR Safety Studies for NRC Enclosure cc: J. C. Cleveland T. S. Kress R. B. Foulds - RES A. P. Malinauskas R. M. Harrington D. L. Moses K. L. Meitner - MRR J. H. Wilson M. H. Holmes - PSC R. P. Wichner R. E. Ireland - 14 TABLE 1. FSV DBA-1 SCENARIO VARIATIONS INITIAL DEPRESSURIZED POWER TCORE (MAX) PEAR TIME FUEL FAILURE* CASES LEVEL •F [•C] DAYS @ t-5 DAYS 20% 2180 1193 3.5 0 30% 2681 1472 3.5 0.042 35% 2902 1594 3.4 0.282 40% 3110 1710 3.1 0.752 PRESSURIZED CASES 40% 2520 1382 1.25 0 (P-CONSTANT) SO% 2795 1535 1.25 0.06% *GOODIN (GAT) MODEL USED. INCLUDES TIME @ TEMP. MODEL, NOT ACCURATE @ LOW FF'S. 'NOTE FAILURE 2 AT 402 POWER VERY SMALL. PP RELEASE IS SMALLER. J .5 F G. I_1-s_V [ac.A .3'1..l itirtc. teems enaaJ j` —o$EcA r v�� 14UC i 1 Illy i E I I I I I I I I I ! ` I:Vo I I i . • • I I : I. , #-:- :I_._.rr: .I -- I - I 11'40 1 .. ! I I t -i I ..- (' L • - -RIrA/ t l . I .. I . F. t :r7r; e:;c--- - - •. -t:. I I. :.. •- I .1. - 1 ,.....1,,F,,,,,___,__,..:,' -= --4-77I—•-- . - PE _:.7...r , ' I I .I , . I I I Ty=, PAl• DA IPPM- -1Y+ FS✓ r oPGX.:LaGi- I I I ..--- . us Past 401.—-y ri f i ! t • I E. 1 L. I 1 + t : t , R t o. -- a • "ft I - -.- j 'r 1. 1 : 1.... . - 1 ! _ I. 1:71..I! . I orizc I . . 1 ` I:. � Its i , I _ . .. I q-n t , . I t • - - j —1 t . I. I .�... - -_.. I .._� . . _.. . ... _ ' j i I i i I : ! I .• ! . . . I . 't �o ;� o 10 Zo 3O 40 to 60 7c . 8O 07 .e_ NEC QUESTION #5 0 WHAT ACTIONS WOULD BE REQUIRED TO RETURN FSV TO POWER OPERATION FROM A PERMANENT LOSS OF FORCED CIRCULATION COOLING FROM PARTIAL POWER CONDITIONS? - PSC HAS NOT PERFORMED ANY DETAILED ANALYSES CONCERNING FUEL TEMPERATURES AND ULTIMATE PCRV INTERNAL TEMPERATURES FOR A PERMANENT LOFC FROM PARTIAL POWER LEVELS - BASED ON ENGINEERING JUDGMENT, WITH NO OTHER CLAIMS WHATSOEVER, OPERATION AT ANY SIGNIFICANT POWER LEVEL THAT WOULD REPRESENT ECONOMIC BENEFIT WOULD IN ALL LIKELIHOOD RESULT IN PERMANENT PCRV INTERNAL COMPONENT DAMAGE. THE ORNL STUDY OF OCTOBER 14, 1985 INDICATES SUPPORT OF THIS CONCLUSION t r a I _ STEAM LINE RUPTURE DETECTION AND ISOLATION SYSTEM (SLRDIS) PURPOSE * Provide Continuous Monitoring of Area Temperatures in both Reactor & Turbine Buildings * Minimize Building Environmental Conditions following Steam Line Rupture to: ▪ Protect Functional Integrity of Safe Shutdown Equipment ▪ Enhance Re-entry into Plant Areas for Recovery SYSTEM SCOPE * Temperature Sensors (Thermistor Cabling) * Microprocessor Logic Cabinet * Interface with PPS for: ▪ Circulator Trip (1) Four Circulator Trip (2) Two Loop Trouble (3) Reactor Scram ▪ Valve Actuation a DESIGN FEATURES/BASES * Capability to Detect: - High Temperature Conditions (1) Pre-alarm @ 150-200 degrees F (2) Trip @ 250 degrees F (3) Rate of Rise Alarm @ 100 degrees F/min. - Trouble - ` * Ability to Withstand Single Active Failure E - - 2/4 Logic in Sensing Circuits - Redundant Microprocessors - Redundant PPS Logic - Redundancy @ Valve Actuations * Capability to Function Without Offsite Power * Necessary to Operate Without Operator Action for First Ten (10) Minutes * Operate in parallel with, but independent from existing Steam Pipe Rupture Detection System - Leave Existing SPRDS Intact - Keep Existing Tech. Specs. r • SYSTEM FUNCTIONS * Monitor RB & TB Area Temperatures ▪ RB - - TB t- * . Monitor Each Zone Redundantly - Four (4) Sensors per Zone - Redundant Microprocessors - 2/4 Logic Trip Signal to PPS * Isolate. Secondary Coolant & Power Conversion System (via PPS) e - Isolate FW (Valves 2, 3, & 4) to Eliminate Steam Production - Isolate MS (Valves 7, 8, & 9) to Eliminate Steam Flow - Isolate RH (Valve 15, 16, & 17) to Eliminate Steam Flow - Trip & Isolate Circulators (Valves 13 & 14, also Water Turbine Valves not shown) to Eliminate Helium Flow thru SG - Isolate Aux. Boiler (Valves 20, 21, 22, 31) to Eliminate Steam Flow * Initiate Reactor Scram (via PPS) SYSTEM RESET & RECOVERY * Consistent with Existing Reset Procedures, to Maximum Extent - Possible * Microprocessor Reset * PPS Reset - Circulator Trip - Valve Actuation (1) Reset 1 (2) Reset 2 FIGURE C : SPAS , SLRDIS , AND PPS REL 'IONSHIP (A LOu C SHOWN, B LOGIC SIMILAR) INPUTS FROM INPUTS FROM SPRDS SLRDIS INSTRUMENTATION - .INSTRUMENTATION 000 OOOO N tW 417- SPRDS . E STEAM LINE -LOGIC CIRCULATOR RUPTURE _ TRIP (ONE PER CIRCULATOR) DETECTION LOGIC RACK PPS (I-93543) t VALVE i ACTUATION .. LOGIC o r EXISTING ADDITIONAL OUTPUTS OUTPUTS IA N REACTOR =193S \' TE-93940, 45 Lza,. " 49zi (5141OW,V) 12 490478-93936 re-93939 A AUTSA9WN- 4BcI it —`�,r I 4644 Io TE- 93943 4949 9 Q Tae/Ate BCD 4. • 7E-9394/-\48Ern 4829 MARM/STOA GABLES �� got F u• 4eos 4811 , a?VEEN S.4402?EL.482B1 4191 S 4191 Sit PLAN 4151 4 I41K - 4111 3 re-93945 ` , r TE- 93944 a '= • iii 1 . ELEVAT/ON 1 (SNo/✓N) ( LOOK/Na EAST?) re- 93942 (MT si/a/N)' Nl.4112 • 2 }- MIr 11reets4 t . TE19.1 _ I- II I rE4craess K -- -1 re 9.J9, 8 II I I � J TAT-939,.9 V I L I a 14 • -- I ---�� i I - I I 7E 19i9481 I I 5 F ---- /7,e , q44 I a -- _ 17t_9,d9421 _TvR,e/NE BL D'G. D -- _ I I I _ C 711939 STEAM L/NE B -- I Ran-thee DETECT/c fYSTCMOVRV/CW AIL_ • PLAN V/EW rwURt I S i RAM LINE RUPTURE DETECTION ISOLATION INSTRUMENTATION J t >M LINE RUPTURE DETECTION INSTRUMENT BUS IA RACK (I-93543)(LOCATION IN CONTROL ROOM) SORB , l- ••• TEMPERATURE SENSING I 4 CONTROL ROOM 4 - I CABLES ....,� Al ANNUNCIATOR TEMPERATURE (TYPICAL) _ I ....a.. B —Th MONITOR �...= 0 A E EA 1 LOGIC • 2 OUT OF 4 TRIP N ANY ,---go(TO PPS) ZONE NPUTS SHOWN (DETAIL A) FOR ZONE I YP OTHER ZONES) I TEMPERATURE Im—S'A F MONITOR Y0 OK b ' II R D I I 4 - t TEMPERATURE I ANNUNC. 4---IN MONITOR .....—NI C I _ �� S LOGIC 2 OUT OF 4 TRIP N ANY —.(TO PPS) - I ZONE ()ETAL A) I TEMPERATURE I MP ANNUNC. OTHER MONITOR I ZONE 0 I . 1 t _ a ......_ a INSTRUMENT BUS IS DETAIL A 2 OUT OF 4 LOGIC 'A' BUS IA (TYPICAL FOR LOGIC 3' BUS IB) ZONE I ZONE 2 ;2 OUT OF AI OUT OF t ` 4 4" OUTPUT TRIP RELAYS • • Yl P21 a x a a. ! Y Yi 11 i I + �{ Yw blip- rt:� Y e �• g Yili w lei !!!'.g Q° ' �g I i € f "plait @ €€lyilyt lyY gLs 2 r ;i r• ti lap • - $ei la Ii Y Yl!Y Y Y Y ow-- a id; fb el i; .steild ai..i �-- $a 4 t , 7i1.1411 r�Y. >t; -a Qe ia ..3 r "' w' Ygog , YY Y•yi r E 'il Y . is ! • r Y • !'5ps 6w Uli' Pre ithentietzlitt.tl>i8:a$.ilh MO . laie a t9 writ=et:elsirt iSgg 000000 ©00000 O O GOO GOO 000000 on O l 1 F N i ! ; = mi E evw = dill r N ee. i g91II ! ! y B Man* t l lean " 12 — it E —r— --S$RS _ _ J _ .I g^ W P r) : 1 I I Z ' ii fr ISIZ ° ' r s' ,1 VI i 11 ®� as M �/ , III 3 at 6 ®00 s a s d � i 01 0 " � - - Z © € p € - a]o Olse os1 g l i - - a. __ —. _ ar I 0 0 a " I - . r 3 m ',-... i ` L. 6! .e -4CJ g o _ — ® a~ w U th _ W r': -::: _ a---- iblq f 2 It€ s at e O— a Fi ii-- J --- gi Og° it t TABLE 1 TAG SLRDIS ACTUATION RESET NO. LOOP DESCRIPTION LOGIC METHOD (1) ACTION SV-2105 1 CIRC 1A SPEED CONT A&B CT - (2) SV-2106 2 CIRC 1C SPEED CONT A&B CT . SV-2109 1 CIRC 1A WATER TURB CONT A&B CT SV-2110 2 CIRC 1C WATER TURB CONT A&B CT " SV-2111 1 CIRC 18 SPEED CONT A&B CT " SV-2112 2 CIRC 10 SPEED CONT A&B CT " SV-2115 1 CIRC 1B WATER TURB CONT A&B CT " SV-2116 2 CIRC 10 WATER TURB CONT A&B CT " HV-2109-1 1 CIRC 1A WATER TURB SUP A&B CT (2) ` HV-2110-1 2 CIRC 1C WATER TURB SUP A&B CT " - HV-2115-1 1 CIRC 18 WATER TURB SUP A&B CT H - HV-2116-1 2 _ CIRC ID WATER TURB SUP A&B CT " e JP- HV-2109-2 1 CIRC 1A WATER TURB DISCH AU CT (2) HV-2110-2 2 CIRC 1C WATER TURB DISCH A&B CT " HV-2115-2 1 CIRC 1B WATER TURB DISCH A&B CT " HV-2116-2 2 CIRC ID WATER TURB DISCH A&B CT " HV-2201 1 FW INLET B XCR (2) HV-2202 2 FW INLET A XCR - " FV-2205 I FW CONTROL A XCR (2) FV-2206 2 FW CONTROL B XCR HV-2203 1 EMER FW INLET B XCR (2) MV-2204 2 EMER FW INLET A XCR " HV-2223 1 SHT STM STOP CHECK A&B CT (2) HV-2224 2 SHT STM STOP CHECK A&B CT " PV-2229 1 SHT 5TM BYPASS A&B XCR (3) - PV-2230 2 SHT STM BYPASS A&B XCR " HV-2292 1 SHT STM STARTUP BYPASS A&B XCR (3) HV-2293 2 SHT STM STARTUP BYPASS A&B XCR " HV-2241 1 RHT STM BYPASS A&B XCR (3) HV-2242 2 RHT STM BYPASS A&B XCR " PV-2243 1 RHT STM BYP PRESS RATIO CONT A XCR (3) PV-2244 2 RHT STM BYP PRESS RATIO CONT B XCR " HV-2249 1 CIRC 1A TURB TRIP A&B CT (2) HV-2250 2 CIRC 1C TURB TRIP A&B CT " HV-2251 1 CIRC 16 TURB TRIP A&B CT " HV-2252 2 CIRC 10 TURB TRIP A&B CT " NV-2253 1 RHT STOP-CHECK A&B XCR (3) HV-2254 2 RHT STOP-CHECK A&B XCR TABLE 1 continued TAG SLRDIS ACTUATION RESET N0. LOOP DESCRIPTION LOGIC METHOD (1) ACTION PCV-5201 - AUX STM TO 150 PSIG HDR A&B XCR - (3) PCV-5213 - AUX STM TO CRH A&B XCR PCV-5214-1 - CRH TO 150 PSIG HRD A&B XCR F (3) PCV-5214-2 - CRH to 150 PSIG HDR A.&B XCR PCV-5214-3 - CRH TO 150 PSIG HDR A&B XCR PCV-5305 - 150 PSIG HDR TO DA A&B XCR (3) (1) XCR indicates valve is actuated thru an XCR in PPS. CT indicates valve is actuated thru Circulator Trip Logic portion of PPS. 17- (2) Requires: a. Return of ambient to below setpoint. b. Reset of microprocessor @ Control Panel c. Reset via exiting methods to recover from Circulator Trip (3) Requires: a. Return of ambient to below setpoint b. Reset of microprocessor @ Control Panel c. Reset of XCR via HS-93375 (HS-93376) on MCB c 4 TEMPERATURE PROFILES Using Bulk Building Temperature Curves o Consistent with nuclear industry ° Containment of water reactors F 135°F Maximum Temperature for Access/Manual Actions o Survey o EPRI Report NP2868 Developed Scenarios ° Offset rupture ° Single-active failure ° Existing ° plant protection and control systems SLRDIS o Safety-Related Equipment o Non Safety-Related Equipment SLRDIS - ° Detectors o Actuation times o Profiles Blowdown Rates - FLASH Building Temperatures - CONTEMPT-G Results U H f • Q 0== f W o N Q O N C .-1 m x Q W C7 Cce W Wger a -� - 1� W - ._N Q is •-• • _ •-4 0 1-1 m F - I E W H W N ►2r t� - •-• az =cc N F O _ _ co .-1 • J • • • • • Isn aL m v M 1 1 1 1 1 1 1 1 1 1 1 1 I I I I • CD m m Cu .r 1-WSa • cu. CI-SONaSWCW I Qwc U - g: wn F F S a t F o Q M O .r m LO L E` m N II) 1 = - Cr W = 0 Q WD W ▪ m r C a - N 012- _ I-I ►I - Z .. a - m • ., E - tD C=q IA1 g D.. '-• W in - � w C F I" - F m Q WW' N CZ O r = h.-,.... CI • H'1Y3$ 30 3ONVH0 ,_ . m . ..... z LA ,.. az .. m 0 in1O 17� r • - C. 1 1 1 1 1 1 1 1 I I IlIMMV 1 , , 1 • CD 1- en N � FWEO_ • OL1. CI-tOmd=WCCW I =WC,. W U S - �- I- 3 E i 0 G F LL - 0 Q N _ f7 N 0 - (.3 m F _ e - e e Ch0 1 � 2 et - W V U r Cr Y La O W T - a as r ILO Q a F O W - w. La - wd F w a - z CD t alW - 1 = N W a Cei c r W m o r - N I G J - 0 0 m wl 37YDS JO 33NVH3 z � m Ln IL m o _ O N N F C an 1 1 1 1 1 1 1 1 I ; 1 I I I I I I I I m CD CD 0 0 1 0 CD is. La en N (10.1 .Nr gwl in I—Wt0- • OW QE—rONd2WCEW_ I pWo W U rya ✓ F = C a C O e: f LL 0. n e 0 e en CD CD v m F m can gal 1 _ • 2 IX W V CO ► 'Cr a _ ix lie j W a a ,_ N O WF a r it r CL" 0, - •- E en 0 W I ir C to Wir ► h IIII W = ► N c ► IX a .. O 0 m .r m • .-1 - LA LL S e • a N I I 1 1 r T ^- m 1 1 1 1 1 • CD m m m m enCuN ..y FW>EO, 0 OIL QFtOtf10,SWKW I qWc La.. ENGINEERING EVALUATION OF LINER COOLING METHOD WITH FIREWATER FOR SHUTDOWN COOLING OF FORT ST. VRAIN - FOLLOWING A HIGH ENERGY LINE BREAK • - PURPOSE OF THE ENGINEERING EVALUATION o Document acceptability of PCRV liner cooling using firewater-as interim method of decay heat removal following a high energy line break (HELB). o Identify specific systems and components to be utilizedt including electrical components in the harsh environment which must function, and define manual actions required. a r BASIS FOR LINER COOLING METHOD o Liner cooling with firewater to be used as the design basis for decay heat removal (DHR) following a HELB during interim period until forced circulation systems/components are qualified. _ o Liner cooling with firewater is used to minimize reliance on electrical components in harsh environments. o Liner cooling with loss of forced circulation (LOFC) is addressed in FSAR as Design Basis Accident (DBA) No. 1. o Liner cooling for DHR based on worst case scenario: ' e --Double ended, offset rupture of steam line --No credit taken for non-qualified electrical components located in harsh environment. --Coincident loss of offsite electrical power --Due to nonqualified components, systems for forced circulation are assumed to be lost. --Liner cooling is last option for core cooling. It can be expected that many forced circulation components, although not qualified, will survive the HELB. c OVERALL PROCEDURE ° Depressurization of PCRV _ -- To limit heat transfer from core to PCRV internals. -- Via Helium Purification System (HTFA, HPC, LTA) to Reactor Building exhaust system. -- Must be initiated within two hours after LOFC to maintain acceptable heat load on Purification System components. o Actuation of Reserve Shutdown System -- To assure adequate shutdown margin throughout the event. -- Must be initiated within 5 hours. o PCRV Liner Cooling -- Decay Heat Removal -- DBA No. 1 assumes Reactor Plant Cooling Water System is available. Use of firewater is addressed as option. -- Must be initiated within thirty (30) hours after LOFC. o Procedures for PCRV liner cooling with firewater currently exist (Operations Order 85-17). These procedures will be revised to reflect design changes being made to support interim operation: -- SLRDIS ` -- Spoolpieces to HTFA li • 2. kw pi C 44 !aa! ! b. . .vi C 44 _ . . , a 2 of a ri .14 41 . I . | !cc la t-I =k | . w a , p!_ _ ! , al 0}} ( 1.1 a set ti kkE- - - I �P.U La / V rtaeg �VI�k j } ! ! ! ! ! I k Z / a lill e.; 8§ ■kE fl V r§C0$ + E 4.1 L m go r | 4.` • co ' -0 in . E �� 6 .a \ ! � r . /� i § to ! .� { / !f ! | . = ii- � 7 a ` - lie h- �� % ! , §� k/ I S. • 3 I S • � r a 'UEr. k : �� • at 4,_re t # � I $■ 4.-- Qt.\j 4 (2 ■ 4k k as §# a: ) n O. m% �f • .. . . . .EA . _i t ■ ® r V ! ! II 1 S. S. § ,,,,..rig ! a SI # - ■ [ ! § a �- 2 f q : 2 3. ! - - 2� in VI � !! a ! \)! t k k SELECTION OF FLOWPATHS FOR PCRV DEPRESSURIZATION AND LINER COOLING o Existing piping o No automatic valves in flowpath (except with manual overrides) o Minimize manual actions o Isolate non-essential users 6 o Provide redundancy a N N 0 In . 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