ML19246C527

Safety Evaluation Supporting Amend 21 to License DPR-34

ML19246C527
Person / Time
Site:	Fort Saint Vrain
Issue date:	06/06/1979
From:	Office of Nuclear Reactor Regulation
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ML19246C522	List: ML19246C521 ML19246C524 ML19246C527
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NUDOCS 7907260155
Download: ML19246C527 (33)

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SAFETY EVALUATION REPORT BY THE 0.:FICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT 21 TO FACILITY CPERATING LICENSE NO. DPR-34 0F PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION DOCKET NO. 50-267 790726cWK 1s 3 358 296

TABLE ^T CONTENTS Page 1.0 Introduction 1

2.0 Stage III Fire Protection 3

3.0 Alternate Cooling Method 12

".0 Instrumented Control Rods 16 5.C Testing of Reactor Building Louver System 22 6.C High Pressure Helium System Valve Closure 25 7.C Firewater Booster Pump 27 8.C Conclusions 31 358 277

1.0 INTRODUCTION

Fort St. Vrain, a 330-MWe high temperature gas cooled reactor (HTGR},

was designed by the General Atomic Company and is being operated by the Public Service Company of Colorado (PSCo) near Platteville, Colorado.

On October 28, 1977, the Nuclear Regulatory Comission (NRC) authorized operation of the reactor up to 70 percent of rated thermal power. All of the power ascension tests have been completed up to 70% of thermal power.

The first refueling at Fort St. Vrain has been completed and the insertion of eight test fuel elements and PGX graphite corrosion surveillance specimens has : 3en finished in accordance with requirements delineated in Amendment 20.

This amendment deals with various design modifications that Public Service Company of Colorado will perform prior to start up and with several changes to the Technical Specifications it has requested; these include:

(1) Modifications to the fire protection system for the three room complex, the Auxiliary Electric Room, the 480 Volt Switchgear Room and the congested cable areas; this constitutes 5tage III fire protection implementation; (2) Conversion of the Interim Alternate Cooling method to 1.'1e final Alternate Cooling Method; (3)

Installation of two modified control rod drive and orificing assemblies which contain instrumentation to further study and evaluate the power, flux, and temperature fluctuations that were experienced; Technical Specifications -Sanges are not required by this installation.

358 2n

(4) Testing of the reactor building louver system on a quarterly basis; (5) Elir.ination of manual isolation of the high pressure helium supply fro the helium circulator buffer supply header; and (6) Addition of two firewater booster pumps to the firewater system to movide augmented capacity to operate a circulator water turbine and to supply feedwater for safe shutdown cooling.

Technical Specification changes dealing with the above are identified in later chapters of this report along with a discussion of the corresponding safety significance of each change.

The Fort St. Vrain reactor is described in the Final Safety Analysis Report (FSAR) submitted for our review in November 1969. The FSAR, as amended, formed a basis for our January 20, 1972 safety evaluation report and a first supplement, which was issued on June 12, 1973.

The operating license, DPR-34 was issued on December 21, 1973. The operating license has been amended twenty one times, including the amendment supported by this safety evaluation. The FSAR and other early documentation continues to support our safety reviews, as augmented by the additional information and the oper-ational reports referenced herein.

The reactor achieve' criticality on January 31, 1974, and low power physics testing was initiu.ed.

These low power tests, identified as the "8,

Series" tests, along with the "B Series," or power ascension, tests were reported in accordance with Section 7 of the Technical Sprcifications. 356 2??

Also, in accordance with the Technical Specifications, Public Service of Colorado provides " Reportable Event" reports ud " Unusual Event" reports on safety items relating to abnonnal, unusur.1 or unanticipated events that occur during the course of plant operaticas.

In audition to the reports received from the licensee, our safety reviews have benefitted from infor-mation on plant statur 'ad operations provided by the Office of Inspection and Enforcement, and by visits to the plant site by technical sr.ec'alists to review plant records and the "as-built" condition of the plant. Oar safety review has also included consideration of comparable light wate reactor experience and policies, infonnation developed on gas cooled reactor safety under the sponsorship of the Office of Nuclear Regulatory Research, and infor-mation develnped during the review of the General Atomic Standard Safety Analysis Report, GASSAR.

2.0 SR GE III FIRE PROTECTION

Background

In April 1975, a routine NRC inspection of Fort St. Vrain revealed that some fire stops in the electrical cable system had not been installed, and that the routing of some cables deviated from the installation criteria set forth in the FSAR. An audit, :..it:sted by PSCo in the Spring of 1975, included a physical audit of the electrical cable system to establish routing of cables and determined that there were a significant number of deviations in the installed electrical system from FSAR requiremcnts.

The initial discovery of deficiencies in electrical cable routine compared to the criteria in the FSAR came shortly after the occurrence of an electrical cable fire at the Tennessee Valley Authority Browns Ferry Nuclear Plant near 358 300

Attens, Alabama. As information on this event became available, it became evident to the PSCo and the NRC staff that the scope of consideration should be broadened beyond the topic of electrical cable segregation and separation deficiencies, and the corrective action required to bring these deficient items into compliance with the FSAR criteria.

The scope of activities was therefore broadened to include consideration of fire prevention, detection, and suppression, and alternate methods for accomplishing orderly plant shutdown and cooldown in case of loss of nonnal ant preferred alternative shutdown and cooldown cable systems for any reason,

e. E., a fi re. These points had been raised in NRC Office of Inspection and Entorcement Bulletins 75-04 and 75-04A following the fire at Browns Ferry.

Discussions with PSCo also included consideration of alternative methods for cooling the reactor which would be capable of operation independent of the occurrence of disruptive faults or events in the congested cable areas; this eventually led to the design of such provisions, and a comitment by PSCo to install the equipment to perform this function.

The items associated with alternative methods of shutdown and cooldown using installed equipment, the installation of equipment for cooldown inde-pendent of events in the congested cable areas, and some acpects of the provisions for fire protection, which have been comittet to by PSCo, are plant upgrading actions rather than actions taken to correct deviations frcm the specific provisions of the FSAR or license. The upgrading actions were considered at the same time as the electrical cable corrective actions because of the interrelationships of equipment, equipment locations, and safety requiremen ts.

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358 301

The discussions held between PSCo and NP.C led to an overall approach for corrective and upgrading actions.

The PSCo proposal entails the following elentents:

(1) A number of specific electrical modificatiens to correct deficiencies; (2)

Improved fire protection, detection and suppression, incle"ing the application to certain cables of a fire retardant mate;1al (Flamemastic);

(3) An alternate conling method to ensure that, even in the event of a major electrical fault or fire in the congested cable a ra,

means would be available to depressurize the reactor and provide continued cooling to the reactor; and (4)

Procedures to cover degraded cooling conditions.

Because of procurement and installation schedules projected to extend beyond the time when the plant was expected to be ready to resume rise-to-power test operation, PSCo proposed that the actions outlined be implemented in three stages to allow the plant to resume rise-to-power test operation at an earlier time than otherwise possible. Three stages of reactor operation have also been pmposed corresponding to the implementation schedule. A sumary of the program proposed by PSCo follows:

Stage 1:

To resume critical operation and operate througn a 22.5-day operational test sequence including a two-day run at about, but not exceeding, 40 percent of rated reactor power:

a.

Complete all modifications to the electrical installation to correct the deficiencies identified and provide assuranc^

that redundant equipment will remain operable in tne event of 358

02

a major electrical fault or fire in the cable system. Action to be taken includes:

(1) Re-routing of a number of electrical cables (including all essential cables) in accordance with the FSAR criteria; (2) Rearranging of load assignments to essential electrical busses; (3) Evaluation and appropriate derating of cables having fire retardant applied or asbestos cloth installed; (4) Correction of any overcurrent protection deficiencies; (5) Modification of the Class IE power source for the plant protection system equipment supplied by 120v a-c Bus 3; (6) Correction, as needed, of cable tray loading to assure that tray fill does not exceed FSAR criteria requirements; and (7) Review electrical cables to ensure that conductor sizes and insulation ratings meet the objectives of the FSAR electrical design criteria.

b. Provide fire protection as follows:

(1) Around the clock roving watch will augment the plant staff.

Specific instructions will be issued covering his duties.

(2) Auxiliary Electrical Equipment and 480 Volt Switchgear Rooms and cable areas adjacent to the "G" anc "J" walls will be equipped with temporary self-contained fire detection units with local audible alarms; 358 303

(3) The number of portable air packs will be increased from six to 16; (4) Specific procedures will be issued for extinguishing an electrical fire using the hose water system; (5) Surveillance requirements for air breathing equipment will be issued; and (6) Personnel will receive training in these measures.

As a result of our review (see Amendment 14), we required that NOTE:

additional fire water pumping capability, namely that two diesel-driven pumps be used in connection with the IACM, will be available during Stage 1 operation. Also as a result of our review, we required that the procedures discussed under item a of Stage 2 below be available for Stage 1 operation.

Stage 2:,

To operate above 40 percent and up to 100 percent of rated reactor power, but no longer than the first refueling, the following corrective action would be implemented:

a. Implement the procedures developed as a result of a study of This includes core cooling under degraded plant conditions.

appropriate operator training.

b. Provide an interim version of the ACM (to be known as the IACM) for Stage 2.

The IACM will be operabic prior to proceeding A

beyond Stage 1, and until the permanent ACM is implemented.

major difference between the ACM and the IACM is that, in the IACM, firewater will be provided by one of two temporarily., n.

installed diesel engine-driven pumps. The second pump will be used to supply an alternate source of water for PCRV cooling or for the purpose of fire fighting. These pumps will take suction from the Circulating Watar Cooling Tower Basin and pump the water through short # ire hose connections into the permanently installed fire water system and then into the PCRV Liner Cooling Water System via existing piping cross connects. The return water is then directed back to the Circulating Water Cooling Tower for cooling and recirculating through pennanently installed plant piping.

Provisions for both the ACM and IACM include depressurization, release of reserve shutdown material, detailed procedures, training and surveillance.

c. Fire protection and detection improvements to be provided for Stage 2 are listed below:

(1) Continue to employ temporary fire detecters and firewatch described under Stage 1; (2) Continue in effect t:

mcedures and instructions for use of water systems on electrical fires; (3)

Install a manually actuated Halon fire suppression system in the Control Room, Auxiliary Electrical Equipment Room, and 480 Volt Switchgear Room. The first two rooms will be activated simultaneously. -

Install three-zone Halon sampling lines in each room.

Provide external sampling staticn adjacent to each manual Halon control val o.

Provide portable three-channel Halon detector.

Provide hand Feld Halon cet!ctors for personnel entering fire. paces; (4)

Instr'l ver.tilation dampers in Control Room, Auxiliary Electrical Equipment Room and 480 Volt Switchgear Room.

Provide remote actuation damper controls adjacent to manual Halon control stations; (5)

Install continuous air breathing hose equipment system in Control Room, and 480 Volt Switchgear Room; (6) Expand in-plant communication capability to include an alternate communication system to assure communications during a fire emergency; (7) Apply a fire retardant coating to cables in the Auxiliary Equipment Room, the 480 Volt Switchgear Room, Peactor Building "J" and "G" walls, Turbine Building side of the "G" wall and underground cableways; (8)

Issue procedures for :ranual Halon activation including ventilation system operation; (9)

Institute personnel training for Stage 2 fire protection measures; and (10)

Issue surveillance requirements for Halon system.

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358 306

The Stage 1 review of the Fort St. Vrain fire protection program was evaluated in Amendment 14 dated June 18, 1976. Stage 2 review was evaluated in Amendment 18 dated October 28, 1977. This amendment deals with the Stage 3 fire protection provisions along with the conversion of the IACM to the ACM.

Stage 3:

Prior to reactor start-up following the first refueling outage the following corrective action would be implemented:

(1) Install a pennanent central fire detection and annunciation system, the fire detection syst ' n is to automatically activate Halon in the Auxiliary Electrical Equipment Room and 480 Volt Svitchgear Room; (2) Install fixed water spray systems in the Auxiliary Electrical Equipment Room, 480 Volt Switchgear Room, and in cable areas adjacent to "J" and "G" walls; (3) Install drip shields over electrical cabinets in the above areas; (4) Modify the fire protection water system as required to support additional spray water requirements; (5) Provide an automatic closing feature for the Auxiliary Electrical Equipment Rocm and 480 '. alt Switchgear Room dampers; (6) Modify tne common room feature of the Control Room and Auxiliary Electrical Equipment Room ventilation system; (7) Close the Control Room floor openings; 358 307

(8) Issue operating procedures for the fixed water spray extinguishing systems and automatic Halon operation feature; (9) Institute personnel training for use of fixed spray systems;

(.10) Issue surveillance requirements for fire detection system cnd fire protection water system; and (11) Establish a plant fire brigade and a brigade training program.

Conclusions These modifications have been reviewed as part of the Fort St. Vrain Stage 3 fire protection program and have been found to follow the applicable guidelines of Appendix A to BTP 9.5-1.

We have also reviewed the plant locations in which the ACM along with its associated cable routing is located using the acceptance criteria that no single fire should be able to simultaneously result in failure of the ACM and the primary systems for cooling down the plant.

From our review, we have concluded that the cable routing through an independent and separate duct bank between the new diesel generator power supply, and the equipment necessary to cool down the plant, satisfies the above requirement and is, therefore, a cceptabl e.

This review terminates the three stages of the fire protection program implamented at Fort St. Vrain as outlined above and ;uthorizes the continuation of the rise to power program except as modified by other issues, viz., fluctuations, accident analyses, etc., discussed in previous amendments.

It should be noted that cne additional fire protection item is still pendir.9, viz., revision of the existing plant fire protecticn Technical Specifications to apply to other safety-related plant areas consistent with the requirements of tne StandarG Technical Specifications dealing with fire protection.

This item will be addecssed in a later amendment and be imole-mented befom operation of the Fort St. Vrain reactor at 100 nercent power.

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3.0 ALTERNATE COOLING METHOD

Background

As a result of discussions with the NRC staff concerning the cable separation and segregatior, problem and considerations regarding additional fire detection prevention and suppression measures, Public Service Company of Colorado has agreed to provide an alternate means of providing continued reactor cooling in the event of the occurrence of disruptive faults or events, such as a major fire, in the cong2sted c.able areas. This alternative cooling method (ACM) will ensure that conditions nd public health and safety conse-quences, analyzed and presented in resian Basis Accident number 1 in the FSAk, are not exceeded in the case of such disrupti:s fwlts or events in congested cable areas.

The ACM is designed t. rcomplisn the following functions:

(1) To maintain the reactor subcritical by means of local, manual actuation of to Reserve Shutdown System; (2) To provide for continued cooling of the PCRV and therefore the core by manually initiating resumption of PCRV liner cooling; (3) To allow manual, local depressurization of the PCRV through the purification system; and (4) To establish manual local operation of the Reactor Building Exhaust system and radiation monitoring of the exhaust effluent to the atmosphere.

The following is a listi g of all equipment associated with thc ACM:

n Diesel-driven generator (2500Kw)

Plant lighting auto-transfer switches Electrical equipment transfer switches 358 309

4160 V to 480 V transformer Stack effluent radiation monitor Firewater pump Service water pump PCRV liner cooling water pumps Reactor plant exhaust fan Helium purification cooling water pump Motor operated valves Firewater pop house vent fans and louvers Service water tower fan Service water return pump Circulating water makeup pump Diesel oil transfer pump Plant lighting Reserve shutdowri system Electric power for the above equipment, about 750 hp, is supplied by the separate dedicated 2500 KW diesel angine driven ~ generator unit in the event offsite and normal onsite emergency power supplies are lost.

The Fort St. Vrain primary coolant system contains four helium circulators, any one of which is capable of supplyir.g adequate core cooling following a reactor shutdown. For shutdown cooling these circulators can be driven by steam from a flash tank or from the auxiliary boiler. They can also be driven by water turbine drives which can be supplied with water from any of three water systems containing a total of eleven separate pumps.

Section 14.4 of the FSAR addressed the short-term (30 minutes or less) loss-of-forced-circulation (LOFC) event. During the subsequent transient following loss of forced circulation, the average primary coolant temperatures within the core would rise to about 1600*F.

Forced circulation cooling of the reactor core must be re-establisted within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> or restart is precluded due to potential damage to the steam generator inlet ducts which pass through the core support floor. An average helium temperature of 2100 F would develop under such conditions.

(Ref. FSAR, Appendix D, Section 0.2.5) ro

i0 JO

A walk-through of the manual operations necessary to put the ACM into service to restore liner cooling, if it had been lost, indicates that the system can be made operational in approximately 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; well within the twenty-five to thirty hour interruption analyzed as the margin available for restoring the liner cooling system to service.

The liner cooling system for the PCRV is designed to remove a total of approximately 24 million BTU per hour with coth independent cooling loops operating. Section 14.10 of the FSAR established that 1 eat removal during the LOFC accident is adequate utilizing only one loop of the system.

To maintain the PCRV concrete temperatures within acceptable limits, to assure the basic integrity of the primary containment, and to provide for reactor core cooling, the PCRV cooling system must be maintained functional.

This is accomplished by using normal power suppifes or if necessary the ACM power supply to run necessa ' equipment.

Since the reactor core must remain in a shutdown configuration throughout the LOFC event, the reserve shutdorn material housed in the control rod drives would be manually released into the core by operator action. 358 3ii

For ACM operation, in addition to maintaining the PCRV liner cooling system in service, other plant systems will also be returned to service, as follows:

(1) Fire Protection System - to provide wa er for fire fighting; (2) Service Water System - to provide cooling water, as required; (3) Helium Purification, Liquid Nitrogen, Helium Storage System, and Purification Cooling Wat.er System to pennit depressurization of the PCRV; (4) Circulating Water Makeup System - to provide makeup water to the Service Water and Firewater Systems; (5) Reserve Shutdown System - to provide backup to the control rods, to assure the reactor remains, ubtritical; (6)

Reactor Plant Ventilation System - to pennit disposal of excess PCRV helium inventory thmugh normal discharge of the Reactor Building exhaust system and to provide for post-accident monitoring of gaseous discharges from the plant; (7) Auxiliary Boiler Fuel System - to transfer diesel fuel from storage to the ACM diesel - generators; and (8) Plant Lighting.

Evaluation The initial safety evaluation of both the ACM and the Interim ACM (IACM) was prm. Anted in Amendment #14 dated June 13, 1976.

This was followed by a Safety Evaluation of the IACM in Amendment #18 dated October 28, 1977. 358 512

We have reviewed the design and imp'ementation of the IACM/ACM conversion cable routing. The requirement to be satisfied was that no single event would simultaneously result in failure of the ACM and the primary systems for cooling down the plant. We have concluded from our review that the cable routing through an independent and separate duct bank between tha new diesel generator power supply and the equipment necessary to coni down the plant satisfies the above requirement and is therefore acceptable.

In addition, we have reviewed the design, installation and operational requirements of the Alternate Cooling Method and conclude tnat the ACM can provide all necessary functions to assure safe plant shutdown and emergency cooling under the degraded conditions caused by a large electrical fault or fire in the electrical system. Therefore, the operation of the plant under the conditions of the proposed revisions to the Technical Specifica.tions as modified by the staff pursuant to agreement reacheo with PSCo, is acceptable.

4.0 INSTRUMENTED CONTROL RODS

Background

Cyclic temperature fluctuations were first noted at the Fort St. Vrain reactor on October 31,1977 at 58% power during the initial rise in power above the previously authorized 40% lesel. The fluctuations were observed in outlet helium temperatures, external thermal neutron flux, steam tempera-tures, and FCRV movement. Temperature fluctuations have remained within design and Technical Specifi ution limits, are non-divergent and rep. aducible.

The fluctuaticns are cere-w 4de and generally out-of-phase from one refueling region to another with a range in period from 5 to 20 minutes. The average core thermal power and averagt helium temperatures remain relati'/ely constant during the fluctuations.} }j}

A total of 30 core fluctuation everts, some occurring spontaneously and some induced for test purposes, have been identified at Fort St. Vrain. Following the initial fluctuation event, plant parameters were monitored throughout plant operations to enable imraediate detection of fluctuations. In all cases it was demonstrated that reducing reactor power was a reliable means of stopping the fluctuations. As part of a continuing comprehensive diagnostic program to investigate and characterize the fluctuations, in-core instrumentation will be installed by modifying two control rod drive assemblies, as described in PSCo letter dated November 17, 1978. The temporary modifications to the two control and orificing assemblies consist of removing one control rod from each assembly and adding an instrtsnent package which occupies the space vacated by the removed control rod. Al though the instrument package /.oes occupy the space vacated by the removed rod, the package is not connected to the corresponding control rod cable; instead, the instrument package is isolated from and independent of normal operation of the remaining control rod. The in-c:;re instrumentatior, location is shown in Figure 1 and consists of the following: (1) Three thermocouples to monitor helium temperature one on each of three axial locations; (2) A self-powered neutron detector (SPND) with a compensating cable to monitor the local / region flux level at the core midplane; (3) A fission couple to monitor local / region flux level near the bottom of the core; (4) A fission chamber to monitor region flux level in the bottom block of the upper reflector; and (5) Two microphones to detect changes in turbulence (flow). 358 31'

Fig. Relative axial tocations v. ...sv.e ,i instrumentation ( ; e-'I; h a ~ l ! }}- l l t I, . l. $ _, j ;, I i Thermoccuple 6 j No. 1 i 4 ij i 1

i I

l 1 l ) Fission ~ Chamber e < j j 1 I, I. l .7 I ~9 1 d '*--~ ,u 1 3 (b*I. i I Gulton Microphone l I l ~ (on C50 assembly g S/N 43 only) .i i l l l' t h. i 1 / SPND l l 71 I I. l Nw l 'Nl] l j N Thermoccuple No. 2 i l' I ) g j Kaman Flicrcphone (en C50 assembly t p i S/N 20 only) i.b7' l [ I t I The r-".o CCup le t I No. 3 I l l./ 4 li n. i 4 W v I J

l i ission-ccuple i

4 W'u ib / ;:' ~l ljll f q I.0M...!'"_.. lim._E. ijf St ~ , \\u' l i N, I. e , p.. \\' \\ / 10 i i 's1 1 i c. b 358 315 7_ .,. u _ 7s t __;c,. - 1-- l i L-l l ~ O.s

These instruments are securely attached at various axial locations to a support rud. The axis of i.he support rod is maintained at a position near the centerline of the control rod channel by means of several spiderlike assemblies at various axial locations. The basic materials of construction are magnesium oxide, stainless steel, inconel and chromel-alumel and all joints are welded. In addition to the above noted instrumentation, the two control and orificing assemblies are provided with pressure transmitters located in the mechanism compartment. They are plumbed in a manner to measure the pressure differential across their respective orifice valves. One of the control and orificing assemblies is provided with a linear variable differential transducer (LVDT). It is also located in the mechanism compartment and is mounted at the upper end of the orifice valve drive mechanism. This device will provide information relative to the vertical movement (and/or growth) of the region's central column over which the con-trol and orificing assembly is mounted. Other temporary modifications were required in order to pass the temporary in-core instrumentation leads through the p*imary and secondary closures. These included meking two temporary ports in the primary closure piece of i.be control and orificing assembly, and the fabrication of temporary secondary closure plates containing four ports. After exiting the temporary secondary riosure plate, the temporary in-core instrumentation leads will pass between hold down plates through a notch provided to acconmodate their passage. 358 316

Following completion of all measurements requiring the use of the instrumented contml and orif!cing assemblies, all temporary instrumentation is to be removed (at the hot service facility) and disposed of. The drives will then t'e returned to their original operational conditions; i.e., the removed control rod will be reinstal'ed and all nonfunctional ports required for the tempu ary instrumentation will be sealed. Evaluation We have reviewed the proposed modifications with regard to their effect on those aspects of core neutronic and thennal-hydraulic behavior important to reactor safety. Public Service of Colorado has proposed to replace one of the control rods in each of two control rod pairs by a string of detectors. The detector strings would not be attached to the control rod drive and the reinaining cc'itrol rods of the pairs would be free to move in accordance with the present withdrawal sequence. Modifications to the control and orificing assemblies will pennit the leads from the detectors to be routed out of the cora. Four different pairs of refueling regions have been investigated as candi-dates for insertion of the detector strinas. They will be located initially in regions 35 and 5. The effect of locating the detector strings in each refueling region on core parameters has been calculated for cycles 1 and 2. The effect on shutdown margin, power distribution, and rod withdrawal sequence have been deter-mined. The Technical Specification requirement of a one percent shutdown margin is met throughout the remainder of Cycle 1 and all of Cycle 2 (the minimum calcu-lated shutdown margin is 2.1 percent). Since the modified control rod pairs are not expected to meet the scram time requirements it was assumed that these rods did not scram as well as the most reactive of the remaining rods. The absence of the second rod of the control rod pair during that por' ion of the rod withdrawal 358 31/

sequence when these rods are deep in the reactor causes a significant increase (40-50 percent when the rods are fully inserted) in the peaking factor of the affected refueling region. However, the power in these regions is low (region peaking factors less than 0.5) and the increase in peaking factor is easily accommodated by adjusting the orifice valves. The differences are well within the Technical Specification limits (LC0 4.1.3). The effect of the removal of the controi rods on the worth of other rods or banks in the control rod withdrawal sequence was ir <estigated. It was determined that the conditions of the Technical Specifications (LCO 4.1.3) regarding maximum rod worth were met.. When operating at power the detector strings are still in the core. However, they have a negligible effect on core power distribution. The analysis of the effects of the control rod modifications have been performed using the calculations methods that were used in the final design of the core. The core startup following the modifications will be used to obtain rod bank worths, critical positions, etc., to verify the analyses. We have reviewed the core physics aspects of the proposed control rod modifications and find them to be acceptable. This conclusion is based on the following considerations. (1) The analysis methods re those previously used (for the FSAR analyses) and accepted. (2) The results of the calculations show comfortable margin to operating limits. (3) Confinnation of the rod worths and peaking factors will be performed during the first startup after the modifications. 358 318

Mechanical failure of an instrument package is enalogous to the failure of a control rod. The bottom of the instrument package has been designed such that in the unlikely event of a failure, the exit to the con-trol rod channel would not be blocked preventing the flow o f coolant. Nevertheless, thermal and fuel performance analyses have bet.n performed assuming the instrument assembly drops to the bottom of the channel and completely blocks the coolact flow while the plant is opera;ing at t ull These analyses shc* th:t, even under these severe conditions and power. if the failure of the instrument asr.embly went unr.oticed for 40 hours, the instrument package would not melt and the impact un fuel performance would be negligible. If the instrument package should fail, the fai'ure would result in anomalous readings which would give indication of such a failure. Also, the bottom of the instrument package has been designed to act as a catcher for small parts, should an unlikely occurrence of that nature happen. The presence of the instrument package in the coatrol rod channel will have an insignificant, effect on the normal helium flow through the channel. Calculations indicate that the flows through the channel containing the instrunent package, through the corresponding channel with the control rod inserted, are within 10%. These conclusions result from a comparison of the relative flow resistances for these cases. The values of flow resistance for the instrument package and for the control rod were detennined by means of full-scale flow tests. The two principal resistances to flow through the control rod channel are at the location where coolant passes through two 5/8-inch diameter holes in the guide tube inside the plenum elements, and at the exit of the channel 358 3i?

where the coolant leaves the channel through a 1-inch diameter hole. Friction and other losses in the channel are small compared to these for both the regular contrel rods and the instrument support structure of the ICRD. The largest flow resistance occurs when the control rod is in any withdrawn position. Control rod components are then opposite the two 5/8-inch diameter holes in the guide tube and require the entering flow to turn after entering. For the inserted control rod on instrument support structure, the resistance of the channel is less than 15% of the total resis-tance so the flow is little affected by the presence or absence of these assamblies. We conclude, based on the analyses and data presented, that the use of the instrumented assembly of the Instrumented Control Rod Drive (ICRD) assembly has a minor effect on the flow through the control roa channel. We conclude also that the temporary modifications to the closure assemblies to acconmodate instrument leads will not affect the performance or integrity of the closures. Having reviewed the core neutronic and thermal-hydraulic behavior of the instrumented assembly as it affects reactor safety and the health and safety of the public, we find use of the ICRD assembly acceptable. Existing Technical Specifications adequately cover operations with the ICP.0 assemblies in the reactor. 5.0 TESTING CF REACTOR BUILDING LOUVER SYSTEM Bac kground The Fort St. Vrain Technical Specifications, Surveillance Requirement 5.5.2, teactor Building Pressure Relief Service, indicates that the reactor buildir g louver system is to be exercised annually. Public Service Company of Colo ado has determined that this interval is too long to assure operability of these devices and PSCo's experience indicates that the louver system should be exercised quarterly to assure its operability. In a June 2,1978 letter 358 320

E 4 PSCo notificd the NRC that to increase the probability of successful operation of the reactor building louver system when called upon to operate, PSCo would intend to exercise the syster on a n,uarterly basis regaraless of reactor status. The NRC found this unacceptable since testing the reactor louvers regardless of reactor status could permit interference with the building ventilation system during fuel handling, for example, and exoose a pathway for activity to the environment without any treatment. We notified PSCo of this and recomend several These prerequisites prior to louver testing by letter dated December 7,1978. recommendations have been incorpated in a formal request for a change to the Technical Specifications dated April 22, 1979. Evaluation The following prerequisites must be adhered to prior and du ing quarterly louver testing. (1) Reactor shall be under normal steady state operating conditions; (2) Primary coolant pressure is within the nonnal envelope for existing conditions; (3) Reactor building ventilation system is operating per Technical Specifications; (4) No radioactive es waste releases are in progress, nor is fuel handling being performed; (R) No airborne activity above background as indicated by the building activity monitors; (6) Area radiation monitors and local alarms are operable per Technical Specifications; (7) No survefilance testing is being performed on the reactor ventilation system or the radiation monitoring systems; (8) Only one segment (group of louvers) of the louver system shall be tested at any given time; -q 3d.

(9) Communication shall exist between personnel performing the tests and the control room operators; (10) Capability shall exist to manually shut the louver panels; (11) Testing of the louver system shall not exceed a total duration of six (6) hours in any one quart?r; and (12) Non-complianca with any of the above conditions will require testing to be discontinued and the louver system will be returned to normal. The louve's around the reactor building consist of 94 sections which are divided i.to 20 groups. 6 groups contain 4 louver sections and 14 groJos contain 5 louver sections. Therefore, one group is tested at a time on approximately 5% of the total. The functional testing of the louver system takes approximately one hour, so that if problems are encountered, an additional four hours are avail-able to identify and resolve the problems and another hc1r to retest that portion of the system involved. If the system were to be tested when the primary coolant system is pressurized and the reactor is in operation, there would be a short period of time during which one section of the louver system would be open, that there would be a possibility for a release of activity to the atmosphere from the reactor building without passing through the reactor building ventilation system HEPA filters and iodine absorbers. In the analysis of the " maximum hypothetical accident" (MHA) primary coolant helium and tt.e associated radioactivity is released into the reactor building over a periad of two hours. In order to account for uncertainties 35b 3Al

in the operation of the reactor building louvers, we have assumed that 10% of the primary coolant system inventory will be released unfiltered from the reactor building. Using our conservative assumptions, the exclusion area boundary (590 meter) doses, viz., Thyroid 4.6 rem, Whole Body 8.6 rem, Bone 36 mrem, are well below 10 CFR 100 guidelines. If an assumption is made that only are group of the louvers, out of a total of 20 groups, is open then the unfiltved release from the reactor building is approximately 5% and the analysis for the MHA is the bounding Based on the above revi ew, we find quarterly testing of the louver case. system, as outlined by PSCo, acceptable. 6.0 HIGH PRESSURE HELIUM SYSTEM VALVE CLOSURE

Background

while the reactor was shut down and depressurized, On February 25, 1976, The buffer two circulators were being operated in the self-turbining mode. helium was being supplied by high pressure bottles rather than by the helium Power purification system which is normally used durina reactor operation. was inadvertently disconnected from a pressure controller, allowing valves in the supply and recirculation lines to fail-open and connecting the 88 psi high pressure bottle header to a 12 psi buffer helium header, producing a This upset the helium buffer system by preventing buffer 76 psia difference. helium recirculation and interfering with drain of bearing water to the high .ressure separator. To reduce the chance 01 recurrence. PSCo hcs implemented a change in the control valve in the line from the hen :'A storage system so that loss of controller power will result in valve closure. 358 323 IPDI ! 1-Ol l2E ![Z1 .[ H S } gg7g r_ _ _ _ _ _,_ _ _ I-0213%I I-052366L2 rk IPDC l ' X5l / N l I-Ol!236_7[ ;2367 '[ g i 10 - sow A 1, q i IS Ps!G.,: _a. 6_g, {W .s 15 79 6cy_y_JA,_m_9 - -> FIS2368 (V23225 ) 2 36G-l n XEP i \\ PD1 (HV N

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\\- H E LIUM CIRC U L ATORS ?URTF. FO FO SYSTEM FC (V23223) 3 .J Pb l C-2101,C-2 i 0 2,C-210 3 (V 2,3 2 77D/_ 67-l I C-2 iO 4 j h - - -- FC ,rW~ g/ \\ (V2 322 0) 2 3 6 6-P-M_ _ Zi] v ~ 2 Q66;2 1 02.:2366-2 u (V2 322d)N u, co X u N -I-Figure 6.1 w LCO 4.3.9,liigh Pressure '~X llelium Supply System A @ 23221) HELIUM Q STORAGE SYSTEN

The Technical Specifications were also changed to require that valves V-M., and V-23221 be placed in the closed position to isolate the helium storage system from the helium circulator buffer helium system when the reactor is in operation. As a follow-up to the above mentioned changes, the system design was re-evaluated and modified so that press.ure regulator valve PDC-2367-1 fails closed upon loss of electrical power, thereby precluding reoccurrence of the original evant. Also in the design, the block valve HV-2366-2 i-nomally shut. This valve will only be opened when there is a loss or unavailability of purified helium flow from System 23, as was the operatino mode when the unusual event occurred. Evaluation Technical Specification LC0 e. 3.9, as written, insures against the enwanted accident flow only in the case of a double casualty, nama' loss of purified helium flow followed by failure of the PDC-2367-1 valve to re3 alate pressure by ncidently being energized to the wide open position. However, with the t C0 as written, a sequence of events could happen when the high pressure storage bottle system is used, which could cause a circulator trip. This is the type of occurrence which the LC0 set out to specifically avoid. We have reviewed the system design and pertinent infomation in the failure mode of pressure regulatory valve. We conclude that the present design of the instrumentation and controls for the system which provides buffer helium makeup from the helium storage bottle system requires two separate failures which could result in the loss of buffer heliur, to the 358 325

circulators. Therefore, since the design now satisfies the single failure requirements, we find it acceptable. Accordingly the proposed deletion of LC0 4.3.9 is acceptable. 7.0 FIREWATER B00 STER PUMP

Background

While simulating safe shutdown cooling of the Fort St. Vrain reactor with firewater on 6/9/76, it was ascertained that flow through the steam generators and helium circulator speeds did not meet the published FSAR values. In this test firewater at 125 psig was supplied to the emergency condensate header using the condensate system while the steam generator discharge pressure was being maintained at 75 psig. With these conditions, the flow through the steam generator was about 540 gpm (270,000 lbs/hr) whereas the FSAR indicates one firewater pump should supply 1,406 gpm to the steam generator and drive one helium circulator to supply about 1.5% to 2% helium flow or approximately 700 rpm for a cold core. 'ubsequent testing identified and corrected the water flow problem, resulting in circulator helium flow consistent with FSAR comitments for shut-down cooling on firewater. However, subsequent reanalysis of the firewater cooldown event indicated that temperatures could exceed those predicted in the FSAR for full power operations. It has been established, however, that the present system without the booster pumps is adequate to keep temperatures within FSAR predicted values for operations up to 70% power. Provision of means for cooldown using firewater as a source of feedwater and circulator drive provides backup means for orderly cooldown in the event normally provided sources are lost. Cooldown in this mode is not essential to safety, since cooling can be provided via the liner cooling systam, using normal power 358 326

sources or the ACM installation previously described, in t:.e remote event t:.t forced convection cooling cannot be pm vided. Firewater for safe shutdown cooling is routed to the steam generators and helium circulator pelton wheel drives through the emergency condensate or emergency feedwater headers. Two boost pumps, including associated valves and piping, have been installed to boost the firewater pressure differential at the pelton wheel nc:zle to a minimum of 175 psid, even though operation of only one pump is required. These pumps are not needed for operations up to 70% power, but are being provided to assurs adequate cooling for operations at full power. Firewater supply to bath pumps is available from two separate headers. The manual supply and routing is from the emergency condensate header and back again through normally open yhlves. The only action required to obtain the boost pressure is to start either pump once firewater has been admitted to the emergency condensate header. The second source of supply to both pumps is directly from the firewate-header through two normally closed valves. The discharge from both pumps can also be routed to the emergency feedwater header. This use of the emergency feedwater header required the installation of a flexible metal hose spool an? the operation of two norn&lly closed val ves. The purpose of this removable spool is to prevent leakage from the high pressure feedwater header to the lower pressure rated pumps and piping components. Normal plant operation with the emergency condensate or feedwater headers remains unchanged. Both pumps are provided with suction and discharge pressure gauges for testing purposes. Pump operability can be verified by throttling emergency coridensate to firewater supply pressure end measuring at the pump discharge and at the nozzle of the operating pelton wheel. The electrical power for thr. firewater booster pumps is obtairad from essential Bus 1 and Bus 3. Evaluation We have reviewed the design and implementation of the electrical, ir strumentation and control systems for the two fire water booster pumps whicn are to be added to the plant design. The purpose of this design change is to provide adequate circulator speed and water supply to the steam gener-atore for safe shutdown cooling utilizing firewater and one helium circulator peltor' wheel drive. The requirement 40 be sat'sfied is that no single event will result in failura of both firewater booster pumps. We have reviewed the firewater booster pumps information contained in PSC's letters to NRC (P-76205), Fort St. Vrain FCN-4089 dated 3/27/78, and FCN-3651 dated 2/20/78, in the areas of physical and electrical separation, equipment qualification, IEEE279-1971, power supply assignments and operability assurance. From our review, we have determined the following: (1) The cable routing for each pump motor and its associated controls will be physically separated in accordance with the cable routing criteria for redundant safety related systems (reviewed and found acceptable by NRC in 1975). The cables in the three room control complex and any congested areas will be coated with Flamemastic. In all other areas adequate physical separation is maintained. (2) The motor and control power for each pump motor is assigned to a separate and independent Class lE power bus. As each bus is supplied power from a diesel generator, the bus loading values 358 328

were reviewed and it was found that the continuous rating of the diesel generator would not be exceeded while shutting down the plant using the firewater mode. (3) The environmental test reports for the additional equipment were audited and found to be in accordance with previous requireTolts found acceptable by NRC. The seismic reports were also audited and found acceptable on the same basis except for hand switches HS-21535 and 21536 and the local pressure indicators PI-21535-1 and -2 and 21536-1 and -2. The information on the hand switches is in accordance with previous requirements found acceptable by NRC. However, this information is not included in PSCo's system of records. The local pressure indicators were not seismically qualified and therefore cannot be depended on to provide assurance that the pumps are operating properly when they are started. Further review identified that the helium circulator speed indi-cators, which are Class lE and are located in the control room, will indicate a minimum speed increase of at least 200 RPM when the pump is started and operating properly. These indicators are required to determine pump operability. We consi .r this adequate. We conclude that the booster pump installation is acceptable and that the proposed Technical Specification changes associated with its surveillance and use are adequate. 358 329

8.0 CONCLUSION

S Based on our review of the documentation referenced in 11s safety evalu-ation report, an evaluation of plant operations thus far, evaluations of the plant through site visits by NRC technical specialists, and favorable reports by the NRC Office of Inspection and Enforcement on conpleted work, we conclude that: (1) because the taanges do not involve a significant increase in the probability or consequence of accidents previously considered and do not involve a significant decrease in a safety margin, the changes do not involve a sig-nificant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Comission's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public. 358 330}}