ML20041E001
| ML20041E001 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/02/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20041D997 | List: |
| References | |
| NUDOCS 8203090782 | |
| Download: ML20041E001 (15) | |
Text
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UNITED STATES
[ h.7 ( ~,j NUCLEAR REGULATORY COMMISSION WASHINGTO N. D. C. 20555
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- e SAFETY EVALUATION REPORT BY THE
-n 0FFICE OF NUCLEAR REACTOR REGU'ATION SUPPORTING AMEN 0 MENT 25 T0 FACILITY OPERATING LICENSE NO. OPR-34 0F PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION DOCKET NO. 50-267 1.0 Introduction Fort St. Vrain, 330 MWe high temperature gas-cooled reactor (HTGR), was designed by the General Atomic Company (GAC) and is operated by the Public PSCo
. Service Ccmpany of Colorado (PSCo) near Plattesville, Colorado.
was issued a construction permit on September 17, 1968 and submitted the Final Safety Analysis Report as Amendment 14 to its application for a construction permit and operating license for the Fort St. Vrain Nuclear Generating Station (FSV) on November 4, 1969. A Safety Evaluation Report dated January 20, 1972 and a first supplement which was issued on June 12, 1973 concluded that FSV can be operated, as proposed, at power levels up to 842 MWt, full 100 percent power, without endangering the health and safety of the public.
The operating license, DPR-34, was issued on December 21, 1973 and has been amended twenty-five times including the amendment supported by this safety evaluation. A listin9 and brief asscription of the' twenty-four prior amendments is presented in Appendix. A.
The reactor achieved criticality on January 31, 1974, and low power physics testing was initiated. These low power tests, identified as the "A Series" tests, along with the "B Series," or power ascensi6n, tests were reported in accordance with Section 7 of the Technical Specifications.
Also, in accordance with the Technical Specification, Public Service of Colorado provides " Reportable Event" reports and " Unusual Event" reports on safety items relating to abnormal, unusual or unanticipated events that occur during the course of plant operations.
In addition to the reports received from the licensee, our safety reviews have benefitted from information on plant status and operations provided by the Office of ~
Inspection and Enforcement, and by visits to the plant site by technical specialists to review plant records and the as-built" condition of the Our safety review has also included censideration of comparable plant.
light water reactor safety under the spons6rsnip of the Office of Nuclear Regulatory Researcn, and information developed during the review of the General Atomic Standard Safety Analysis Report, GASSAR.
I 8203090782 820302 PDR ADOCK 05000267 P
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This amendment modifies the license to permit possession of additional sources and revises the Technical. Specifications (TS) to:
(1) specifj the period of time and conditions,under which the Unit Auxiliary Transformer can be removed from service with the reactor at power, substitute the requirement for an Annual Operating Report with an (2)
Annual Occupational Exposure Report and a Monthly 6perat ng Rerort, i
(3) allow manual reset of the 30% bistables for operation at less tnan 30% but greater than 10% power, (4) require operability of snubbers when the reactor is at power, (5) establish an upper time limit for a loss of voltage to 480 'olt buses v
and the method to be used during surveillance testing, s
(6) substitut.e outdated requirements with those that comply to 10 CFR 20.103, (7) revise the method of performing thermocouple testing, (8) specify the testing of carbon sample canisters-as per NRC requirements, and (9) include two additional snubbers.
2.0 Radioactive Sources 2.1 Backcround 20, 1978, Public Service Company of Colorado By letter dated January submitted an application for an amendment to Facility Operating License No. DPR-34 for the Fort St. Vrair..sclear Generating Station.
The licensee has requested that Item 2.C.(4)a. be amended to include Radium.
223 and 226 not to exceed 5 millicuries per nuclide.
2.2 Evaluation and Conclusion The additicaal sources requested by this change are required for calibration The additional sources requested are consistent 'dith current licensing uses.
The practice and do not involve a significant hazards consideration.
st'aff, therefore, finds the request acceptable.
3.0 Auxiliary Electric System 3.1 'Backcround 20, 1978, the Public Service Company of Colorado By letter dated January requested a change to Technical Specification LCO 4.6.l(a). The requested change specifies the period of time and conditions under which the Unit Auxiliary Transformer can be remov2d frcm service with the reactor at
- power, 1
3.2 Evaluation and Conclusions The present Technical Specification does not specify the time frame or conditions for the removal of the Unit Auxiliary Transformer from The addition to.the Technical Specification of a 24-hour service.
period when the Transformer may be inoperable, as well as the require-ment that for removal, both diesel generator sets must be started immediately prior to taking the Transformer out of service, their controls subsequently placed in the automatic mode, and all three
'480 volt AC essential buses operable is consistent with the present requirements for the Reserve Auxiliary Transformer.
Specifying the conditions under which the transformer may be removed from service provides for an alternate source of power to equipment necessary for safe shutdown of the plant.
The licensee's proposed change has been evaluated against the basis for LCO 4.5.1(a) and our review has resulted in our finding that the proposed change will not impose an adverse affect on either the health Thus, we find the and safety of the public or on the environment.
licensee's proposed Technical Specification change acceptable.
4.0 Recortino Recuirements 4.1 Backcround The Nuclear Regulatory Commission requested of occupational exposure data be deleted from the Tec Specification.which meets the desired NRC objectives.
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4.2 Evaluation and Conclusions AsrequestedbytheNRC$etterdatedSeptember 19,1977, the Public Service Company of Colorado submitted a change to their Technical Specifications deleting the requirement for an Annual Operating Report and incorporated the requirements for an Annual Occupational Exposure Report and for a, Monthly Operating Report.
The Technical Specifications define the reporting requirements for both the Annual Occupational Exposure Report and the Monthly Operating Report.
The staff has reviewed the submittal against the Draft Regulal tory Guide 1.16, Reporting of Operating Information - Appendix A Technical Specifications and the requirements specified in the NRC letter dated September 19, 1977.
Based on this review, we
' find the licensee's proposed Technical Specification change acceptable.
5.0 Reset of Rod Withdrawal Prohibit Bistables 5.
Backoround The Technical Specification section LC0 4.4.1 specifies the limiting conditions for operation o.f the plant protective system instrumentation.
Note (n) of the tables refers to the trip setting of the rod withdrawal prohibit (RWP) part of the plant protective system. Originally note (n) of the Technical Specifications stipulates the trip setting of the linear channels 3-8 for the high power RWP af ter reactor power initially exceeds 30 percent. The proposed Technical Specification change takes into account reactor power decrease from greater than 30 percent.
5.2 Evaluation and Conclusions The Technical Specifications were originally written to include' a trip setting for the Rod Withdrawal prohibit after the reactor power initially exceeded 30%. After initial reactor power increase, the plant protective system RWP exhibited trips in situations inv61ving. operation at less than 30% power but greater than 10% power after reactor power was decreased from greater than 30%.
The revised technical specifications allow manual reset of the 30% bistables for the above mentioned situation.
In this manner, the Interlock Sequence Switch can be placed in-the Law Power position and the rod withdrawal prohibit manually cleared, as necessary, without having to decrease reactor power to 10%.
The staff has reviewed the proposed Technical Specification change and found that the change will not impose an adverse effect on the health and safety of the public. We find the proposed change acceptable.
6.0 Snubber Ooerability 6.1 Backorcund The Technical Specification LC0 4.3.10 stipulates the limiting condi-tions for operation for the shock suppressors or snubbers. The original Technical Spacifications stated that during all modes of raattor operation except shutdown and refueling the. snubbers on Class piping systems shall be operational. Also, if Tnubbers are not operational, the plant shall be in a shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
By letter dated May 9, 1978.the Public Service Company of Colorado requested a change to the Technica1 Specifications such that the snubbers are required to be operational whenever the reactor.
is operated at power. Also, if the snubbers are not operational, the reactor shall be in a low power condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
6.2 Evaluation and Conclusions The proposed Technical Specification change utilizes the definitionL of reactor operation at power as being any operation with the Linear Power Range instrumentation indicating more than 2% of' rated thermal power. Also, the proposed change permits operation of the reactor at power levels up to 2% of rated thermal power if a snubber is found to be inoperable.
The NRC staff realizes that the proposed Technical Specification change is less conservative than the original in that the original sbecification requires the reactor to be in a shutdown condition under the same circumstances. However,,several analyses were performed by the staff resulting in findings that approve the change.
l As part of the approval for operating the Fort St. Vrain reactor above 2% and 40% of rated power, the staff performed analyses as'part of the Safety Evaluation Reports supporting Amendments No. 9 and 14, respectively.
The analyses evaluated the loss of operability of vital equipment resulting from an electrical fire or fault. As a-i result of the analyses, an Alternate Cooling Method (ACM) system was installed at the Fort St. Vrain plant. 'The objective in supplying an ACM i s to assure that means will be available to depressurize the plant and to continue reactor cooling via the liner cooling system in the remote event that normally required equipment should become inoperable. The staff has calculated that in a depressurized state, full power decay heat levels can be dissipated to the liner cooling system by conductive and radiant heat transfer and that localized heating caused by natural convection of low pressure helium would be insignificant.
Based on our review of the proposed Technical Specification change, the analyses performed incorporating the ACM functions and the Safety Analysis Reports of Amendments No. 9 and la, the staff finds that the proposed change will not impose an adverse effect en either the-health l
i and safety of the public or the environment. Therefore, we find the licensee's proposed Technical Specification change acceptable.
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7.0 Plant Protective System Instrumer.tation 7.1 Backcround By letter dated June 19, 1979, t$ePublicServiceCompanyofColorado requested a change to Technical Specification LCO 4.4.1 and -SR 5.4.1.
This superseded the change as requested in Pu'blic Service Company of Colorado's letter dated January 20, 1978.
The submittal was a' direct result of the inspection c6hducted
. January 1978, during which the NRC inspector determined that the Technical Specification specified a trip setting of less than 60%
rated 480 volt bus voltage for longer than 30 seconds with no upper time limit.
The basis for the setting is to provide a reactor scram to prevent any power to flow mismatches from occurring and a delay of 35 seconds follow-ing a power loss to allow for the emergency diesel generators to start and pick up the necessary loads.
The requested change established an upper time limit for a loss of voltage to 480 volt buses lA and IC, and in the method to be utilized*
during surveillance testing.
7.2 Evaluation and Conclusions The previous LC0 specificaticis required a trip setting of less than 60%
rated 480 volt bus voltage fcr longer than 30 seconds and a monthly test by application of 40% of 480 volt bus rated voltage.
The requested change specifies that a reactor trip will occur in no longer than 35 seconds.
A monthly test by simulated loss of'480 voltage is to be performed.
The change had deleted the 60% of rated 480 valt bus under voltage relay set point and the simulated test at 40% of rated 480 volt bus voltage frcm the' Technical Specificatiens.
By letters dated October 20, 1980, and December 15, 1981, Public Service Company is committed to completion of their degraded voltage condition modifications and to provide Technical Specifications related to under^ voltage protection.
The monthly test The licensee has, c'ces provice for a simulated loss of 480 volt voltage.
following discussions with the staff, agreed to perform an interim surveil-lance test on the 480 volt and 4160 volt under voltage rslays prior to startup from the current plant outage.
The NRC Region IV Resident Inspector will review the results of the surveillance testing.
The licensee's proposed change has been evaluated against the basis for LC0 4.4.1 and our review has resulted in the finding that the 4
proposed change will not impose any adverse effect on either the health and safety of the public or on the environment.
Thus, we find the licensee's proposed Technical Specification change acceptable.
,. 8.0 Procedure's - Administrative Controls
8.1 Background
' Technical Specification AC O.4.d. specifies the procedural rr aments and provides for conditions and limitations regaruing the use of respiratory protective equipment.
In a letter dated July 24, 1980, the Public Servic Company of Colorado submitted a change to their Technica.1 Specifi-cations deletino enumerated and outdated requirements, and specifying a requirement to comply with those in 10 CFR 20.103.
8.2 Evaluation and Conclusions The requested Technical Specification change deletes all of the present AC 7.4.d requirements to replace them with a requirement to develop personnel radiation protection pro-cedures consistent.with the requirements contained in 10 CFR 20.103.
10 CFR 20.103 specifies the requirements for exposure of individuals to concentrations of radioactive materials in air in restricted areas.
The change updates the' Technical Specification to reflect current regulatory requirements and eliminates the disparity between the old Technical Specifications and 10 CFR 20.103 requirements.
It also eliminates the need for future chances to this Technical Specification.
The staff has reviewed the submittal and we, therefore, fi'nd the licensee's proposed Technical Specification change acceptable.
9.0 ThermocouDie Testinc 9.1 Backcround Technical Specification SR 5.4.1 specifies the surveillance checks, calibrations and tests of reactor protective systen and other critical instrumentation. By letter dated July 24, 1980 (P-80229), the Public Service Company of Colorado requested a change to the Technical Speci-ficaticns to permit a different method of calibrating the tnermocouples in both the Scram and the Loop Shutdcwn Systems.
9.2 Evaluation anc Conclusions The requested technical specification change requires the comparison l
l of each thermocouple output to a NBS traceable standard instead of the i
l previously specified calibrated RDT laboratory standard. The necessity to calibrate the subject thermocouples in plant environments which do not permit use of a laboratory standard required a change in the method of calibration. The proposed method assures that an~ acceptable calibration will be performed without direct use of a l_aboratory l
standard in the plant environments.
The staff has reviewed the requested change and found that it will not impose any adverse effect on either the health and safety of the public or on tne environment.
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Therefore, we find the licensee's proposed Technical Specification change acceptable.
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10.0 Reactor Building Exhaust Filters, Surveillance 10.1
Background
Technical Specification SR 5.5.3 specifies the test procedures and the frequency of performing the tests on reactoi building exhaust filters.
Public Service Company of Colorado added external carbon sample canit,ters across the charcoal adsorbers in the reactor building ventilation system exhaust filters. By letter dated December 31, 1980,"the Public Service Company of Colorado requested a change to their Technical Specifications to incorporate the NRC staff recommended practice for filter surveillance.
10.2 Evaluation and Conclusion The external carbon samole canisters were added to allow periodic sampling of charcoal effectiveness without affecting the integrity of the adsorber beds.
The change to the Technical Specification incorporates the NRC staff recommended practice presented as Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2 dated March 1978.
The staff has reviewed the proposed Technical Specifications change and found it acceptable, 11.0 Shock Suppressors, Snubbers 11.1 Background By letter dated December 3),1980, the Public Service Company of Colorado requested a change to Technical Specification LC0 4.3.10, Table 4.3.10-1.
The change adds two snubbers to those included.in the Technical Specifi-cation for the Boiler Feed System.
The Fort St. Vrain Technical Specifications dealing with the Limiting Conditions for Operations of Shock Suppressors or Snubbers states (Section 4.3.10c) that shock suppressors may be added to Class I systems without prior license amendment to Table 4.3.10-1, provided a ravision license amendment request.
to Table 4.3.10-1 is included with a subsequent l.1.2 Evaluation and Conclusion The Public Service Ccmpany of Colorado elected to. update Table 4.3.10-1 by the addition of two Class I Snubbers (SFS-166 and SFS-154E) as a result of an Audit program for 2h" and larger piping. The audit identified SFS-166 and BFS-154E as located on, Class I piping, but was not included
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in Table 4.3.10-1 of the Technical Specificaticns.
This change corrects that error and fulfills the requirements of LCO 4.3.10(e).
As reccamended by IE Euiletin 79-14, Public Service Ccmpany has reviewed their " Safety-Related" piping systems to verify that the design drawings The recuested addition is a correspond to "as-built" configurations.
result of this review.
-9 12.0 Staff Conclusions Based on our review of the dncumenta'tlon referenced in this report, an evalua-tion of plant operations thus f ar, evaluations of the plant through site visits by NRC technical specialists, and favorable reports by the NRC Office of Inspection and Enforcement, we conclude that the Technical Specifications can be revised as requested. This safety report describes the basis for. this conclusion and notes the conditions which apply.
Since full power operation has previously been approved, this amendment presents the staff's continued app'roval of certain items for full power operation. The staff has determined that the issues addressed in this amendment will not result in any.significant safety or environmental impact not previously evaluated.
Since the Fort St. Vrain reactor is the first and only plant.of this size and type, and since a substantial base of experience comparable to that for light water reactors does not exist, the performance of the Fort St. Vrain-reactor continues to be closely monitored by NRC staff.
Environmental Consideration We have determined that the amendment does not authorize a et snge in ef fluent types or total amounts nor an increase in power lessl and Having made will not result in any significant environmental impact.
this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is ressonable assurance that the health and safety of the oublic will not be endangered by operation in tne proposed manner, and (3) such activities will be conducted in compliance with the Commissica's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: March 2, 1982 Principal Contributors:
Region IV Tom Westerman Bill Dickerson NRR/DL G. Kuzmycz l
APPENDIX A CliRON0 LOGY OF FORT ST. VRAIN LICENSING ACTIONS
-PERTAINING TO PLANT OPERATION, SAFETY EVALUATIONS AND LICENSE AMENDMENTS DATE TITLE September 17, 1968 Commission issued a construc' tion permit for the Fort St. Vrain Nuclear Generating Station.
November 4,1959 Public Service Company of Colorado sLbmitted the FSAR as amendment 14 -to its application for a construction permit and license.
Safety Evaluation by the Division of Reactor Licensing, January 20, 1972 O. 5. Atomic Energy Ccmmission in the matter of Public Service Company of Colorado - Fort St. Vrain Nuclear Generating Station, Occket No. 50-267. This document pertained to the review of the Final Safety Analysis Report prior to issuance of an operating license.
Supplement No.1, Safety Evaluation by the Directorate June 12, 1973 of Licensing, U. S. Atomic Energy Commission in the matter of Public Service Company of Colorado - Fort St. Vrain Nuclear Generating Station, Docket No. 50-267.
This document pertained to_ postulated high energy pipe ruptures outside containment.
December 2i, 1973' License No. DPR-3 issued for the operation of the Fo.-t St. Vrain Nuclear Generating Station.
Safety Evaluation by the. Directorate.of Licensing May 17, 1974 Supporting-Amendment No. I to License No. DPR-34.
(1) making Chtnges the Technical Specifications by:
exceptions to requirements for installation of secondary closures during certain initial low power physics testing, (2) revising specifications for monitoring during certain radioactive ef fluent releases, (3) revising
<j specification for tenden load cell and PCRV concrete crack surveillance, (4) revising certain specifications for checks, calibraticns, and testing'of loop shutdown system, and (5) redefining certain administrative responsibilities and authorities of the.offsite Nuclear Facility Safety Comittee.
Safety Evaluation by the Directorate of Licensing June 27,1974 Supporting Amendment No. 2 to License No. OPR-34 Changed the Technical Specifications to revise the organization of personnel for Fort St. Vrain nuclear Generating Station.
Safety Evaluation by' the Directerate of Licensing
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July 12,1974 Sapporting Amend. Tent No. 3 to License No. OPR-34 Chunged tne Technical Specifications to allow low power reactor operaticn with a helium environment in the reactor during Phase I of tne ;:ower ascension program.
Date T,i tl e Nove:6er 11,1974 Safety EvaluatiSn by the Directorate of Licensing Supporting Amendment No. 4 to License No. DPR-34.
Changed the Technical Specifications to permit revision of (1) radial power peaking factors under certain operating conditions and (2) the number of core regions allowed the maximum deviation in outi&t temperature-from the average core outlet temperature.
Safety Evaluation by the Directorate of Licensing December 19,1974 Supporting Amendrent No. 5 to License No. OPR-34.
Changed the Technical Specifications to permit revised staffing requirecents for plant operati-ng shifts.
Safety Evaluation by the Division of Reactor Licensing, January 23, 1975 Supporting Amendment No. 6 to License No. OPR-34.
Chanced the Technical Specifications to permit a change in calibration frequency for one adjustment of the w'ide range power instrumentatior. and added a calibraticn
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requirement for the linear range power instrumentation.
Safety Evaluation by the Office of Nuclear Reactor April 17,1975 Regulation Supporting Arendment No. 7 to License No.
DPR-34 Changed the Techn'ical Specifications to permit
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bypass of the two-loop trouble scram when the reactor mode ' switch is in the " fuel loading" positics.
Safety Evaluation by the Office of Nuclear Reactor December 1,1975 Regulation Supporting Amendment No. S to License No.
Permitted the possession d use of adlitional-OPR-34 radioactive ' sources for the purpose of Calibration and instrument checks.
Safety Evaluation by the Office of Nuclear Reactor December 29, 1975 Regulation Supporting Amendment No. 9 to License No.
Changed the Technical Specification to permit OPR-34 a reduction in the helium circulator high-speed trip when operating on water-driven Pelton turbines.
Safety Evaluation by the Office of Nuclear Reactor January 27, 1976 Regulation Supporting Amendment No.10 to License No.
Changed the Technical Specifications to permit OPR-34.
a change in the procedures to be followed in'the event of trouble with the hydraulic power system.
Date Ti tl e_
Safety Evaluation by the Office of Nuclear Reactor April 15,1976 Regulation Supporting Amendm'ent No.11 to Licenso No.
4 Changed the wording in the Technical Specifi-DPR-34.
cations to eliminate an inconsistency in the plant protection system labeling and the FTnal Safety Analysis Repo r t.
Safety Evaluation by the Office of Nuclear Reactor April 26,1976 Regulation Supporting Amendnent No.12 to License No.
Changed the Technical Specifications to add DPR-34 surveillance requirements for h.elium circulators and helium circulator Pelton wheels.
Safety Evaluation by the Office of Nuclear Reactor June' 18,1976 Regulation Supporting Amendment No.13 to License No.
Changed the Technical Specifications to:
DPR-34 (1) add requirements for operation of analytical system moisture monitors between reactor shutdown and 5 percent power; also calibration frequency for these mon.itors is stated; (2) revise allowable primary system impurity levels and method of specifying moisture impurity from parts per million to dew point temperature; (3) add a definition of'ooerable dew point reisture monitors; (4) add functional checks and tests for dew point moisture conitors; (5) revise the core reactivity status surveillance and limiting conditions for operation; (6) isolate the helium storage system from the helium circulator buffer helium system when the reactor is in operation; (7.) allow bypass of plant protective system moisture monitors for' testing during the startup testing program; and (8) add reporting requirecents.
Jur,e 18,1976 Safety Evaluation by the Office of Nuclear Reactor Regulation supporting amendment no. 1,4 to licensee no. OpR-34.
Revised the Technical Specifications to add requirements for: (1) backup pumping capability to the fire water system; (2) surveillance for the added pumps; and (3) an additional class lE pcwer source for the plant protective system.
June 24, 1976 Sa fety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.15 to License No.
DPR-34 Changed Technical Specifications to add requirements for operability and surveillance of shock suppressors.
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Day Title _
Safety Evaluation by the Office of Nuclear Reactor November. 17, 1976 Regulation Supporting Amendment No.16 to License Revised the section of the Technical No. DPR-34.
Specifications relating to administrative controls.
Safety Evaluation by the Office of Nuclear Reactor December 8,1976 Regulation Supporting Amendment No.17 to License No.
DPR-34 Tempor'arily revised the provisions in the Technical Specifications relating to operation of the bearing water makeup pumps in the primary coolant system.
Safety Evaluation by the Office of Nuclear Reactor 0-t:ber 28,1977 Regulation Supporting Amendment No.18 to License No.
Permitted Stage 2 operaticn up to 70 percent of rated thermal power.
safety Evaluation by the Office of Nuclear Reactor Februa rf 23, 1979' Regulation Supporting Amendment No.19 to License No. DPR-34.
Incorporates the Fort St. Vrain Amended Security Plan as part of the license.
April 20,1979 Safety Evaluation by the Office of Nuclear Reactor Regulation supporting amendment no. 20 to license no. DPR-34.
Revised the Technical Specifications to:
(1) install eight test fuel elements into the reactor core at the first refueling, and (2) install PGX graphi.te surveillance specimens into five bottem transition reflector, elements of the reactor core.
bJns 5,197 9 Safety Evaluation by the Office of "uclear Reactor Regulation supporting amendment no. 21 to license no. OPR-34 Revised the Technical 3pecifications to:
(1) modify the fire protection system for the three room complex, the Auxiliary Electric Room, the 480 Volt Switchgear Room and the concested cable areas; this constitutes Stage III fire protection implementation; (2) convert the Interim Alternate Cooling method to the final Alternate Ccoling Method; (3) test the reactor building louver system on a quarterly basis; (4) eliminate the manual isolation of the high pressure helium supply from the helium circulator buffer supply header; and (5) add two firewater booster pumps to the firewater system to provide adequate capacity to operate a circulator water turbine and supply emergency cooling water for safe shutdown cooling.
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Date Title August 19, 1980 Safety Evaluation by.the Office of N'uclear Reactor Regulation supporting Amendment No. 22 to License No.
Revised the Technical Specifications to (1) change the amount of diesel fuel in each diesel gener-ator set day tank to 325 gallons; (2) update company reorganization based on NRC requirements; (3) change the number of hours that the ACM diesel generator can operate with 10,000 gallons of fuel to 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br />; (4) alter the Fire Protection Technical Specifications to follow the requirements of STS on Fire Protection; (5)
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change the frequency and. method of Reactor Protective System Surveillance to satisfy the requirement of IEEE-279-1971; (6) update the listing of all snubbers; (7) change the fissile carticle thorium to uranium ratio to reflect "as manufar uured" specifications and (8) change the values for core region peaking factors and outlet tempera'ure dispersions to reflect existing values in conjunc' ton with accident reanalyses in support of full power operation.
March 16, 1981 Safety Evaluation by the Office of Nuclear Reactor Regulation supporting Amendment No. 23 to License No. DPR-34.
Revised the Technical Specifications to (1) extend the minimum sample flow limits to cover the reactor power range of 70 to ".00 percent, (2) define the times to depressurization, (3) extend the core residence time of the fuel test elements' and (4) specify operator action time limits for power-to-flow ratios as per S.L.
3.1.
November 9, 1981 Safety Evaluation by the Office of Nuclear Reactor Regulation supporting Amendment No. 24 to License No.
OPR-34. Revised the Technical Specifications to (1) 9 incorporate the shift Technical Advisor position and to reflect recent organizationa.1 changes and (2) permit the interspace between primary and secondary closures of steam generator module B-2-3 to be maintained at a pres-sure slightly above cold reheat steam pressure, on a t
temporary basis.
A