ML19331D988

Safety Evaluation Supporting Amend 22 to License DPR-34

ML19331D988
Person / Time
Site:	Fort Saint Vrain
Issue date:	08/19/1980
From:	Office of Nuclear Reactor Regulation
To:
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ML19331D986	List: ML19331D985 ML19331D987 ML19331D988
References
NUDOCS 8009040513
Download: ML19331D988 (35)

Text

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SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT 22 TO FACILITY OPERATING LICENSE NO. DPR-34 0F PUBLIC-SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION DOCKET NO. 50-267 8000040 g jg i

4 TABLE OF CONTENTS Page 1

1.

Introduction 3

2.

Auxiliary Electric System 4

3.

Administrative Controls 5

Diesel Generator 4.

5 5.

Fire Protection 8

6.

Reactor Protective System Surveillance 9

7.

Shock Suppressors la 3.

Design Features 14 9.

Accident Reanalyses 32 10.

Conclusions

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1.0 INTRODUCTION

Fort St. Vrain, a 330-MWe high tenperature gas cooled reactor (HTGR), was designed by the General Atomic Company and is being operated by the Public Service Companj of Colorado near Platteville, Colorado.

On October 28, 1977, the Nuclear Regulatory Commission 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, which was authorized by Amendment la dated October 28, 1977.

This amendment deals with various design and administrative modifica-tions that Public Service Company of Colorado will perform to ensure increased reliability of operation, maintenance and safety of their Fort

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St. Vrain nuclear generating station. These modifica tions have been requested by means of letters and requests for changes to the Technical Specifications and include:

1.

Changing the amount of diesel fuel in each diesel generator set day tank to 325 gallons.

2.

Company reorganizations based on NRC requirements.

3.

Changing 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.

Completion of the Fire Protection Technical Specifications to follow the requirements of Standard Fire Protection Technical Specifications.

5.

Changing the frequency and method of Reactor Protective System Surveillance to satisfy the requirement of IEEE-279-1971.

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6.

Changing the table listing all snubbers in the plant to reflect an updated status as a result of additions.

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7.

Changing the fissile particle thorium to uranium ratio to reflect "as manufactured" specifications.

j-8.

Changing the values for core region peaking factors and outlet temperature dispersions to reflect existing values in conjunction with accident reanalyses in support of full power operation.

This last item is one of three that limit the operation of Fort St. Vrain to 707, power. The other two are Helium Depressurization and Moisture Injection tests; these items will be discussed in the next Safety Evaluation Report of Amendment 23.

The Fort St. Vrain reactor is described in the Final Safety Analysis Report submitted for our review in November 1969. The final Safety Analysis Report, as amended, formed a basis for our January 20, 1972 safety evaluation

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report and a first supplement, which was issue <. on June 12, 1973. The operating license, DPR-34 was issued on December 21, 1973. The operating license has been amended twenty-two times, including the amendment supported by this safety evaluation.

The Final Safety Analysis Report and other early documentation continues to support our safety reviews, as augmented by the additional information and the operational reports referenced herein.

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The reactor achieved criticality on January 31, 1 974, and low power physics testing was initiated. These low power tests, identified as the "A Series" tests, along with the "B Series," or power ascension, tests

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were reported in accordance with Section 7 of the Technical Specifications.

Also, in accordance with the. Technical Specifications, Public Service of Colorado provides "ReportaP e 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, cur safety reviews have benefitted from infor-mation 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 plant. Our safety review has also included consideration of comparable light water reactor experience and policies, information developed on gas cooled reactor safety under the sponsorship of the Office of Nuclear Regulatory Research, and information developed during the review of the General Atomic Standard Safety Analysis Report, GASSAR.

2.0 AUXILIARY ELECTRIC SYSTEM Public Service Company of Colorado requested a change to the Fort St. Vrain Technical Specifications, Limiting Conditions for Operation of the Auxiliary Electric System in their letter dated March 7,1979 (D-7 905 5).

Specifically, the Technical Specifications, among other require-cents, state that the reactor shall not be operated at power unless both the diesel-generator sets are operable and 500 gallons of fuel exist in each cay tar.k.

Public Service Company o-f Colorado proposes to change the amount cf fuel in each day tank to 325 gall.cns.

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Discussions with the applicant and the NRC Resident Inspector revealed that the diesel generator set day tank level control system does not start refilling the tanks until the level reaches 350 gallons.

Each day tank is of a 500 gallon capacity and when the diesel generator is operating, the

' tank level obviously decreas'es below the 500 gallon mark without being refilled.

If operation of the diesel generator is stopped when the day tank level is between 500 and 350 gallons, the fuel oil pump is not started automatically and the day tank level will remain below the 500 gallon limit thereby requiring manual topping-off.

We have reviewed the proposed change and conclude that a 325 gallon supply of fuel oil in' each day tank is sufficient to bring emergency electric supplies on line at load without interruption owing to inadequate fuel supoly prior to transfer of fuel from the main storage tank. There-fore, the change is found to be acceptable.

3.0 ADMINISTRATIVE CONTROLS Public Service Company of Colorado requerted changes to the Fort St.

Vrain Technical Specifications, Administrative Controls sections dealing with organization and procedures in their letters dated January 23, 1979 (P-79015) and January 11,1980 (p-80003). These changes reflect a company reorganizations based on NRC requirements related to Fire Protection and ether issues during the past year.

We have reviewed the proposed changes in light of the requirements presented in the Standard Technical Specifications dealing with Fire protection and the requirements of Quality Control and find them acceptable, contingent on receipt of notification that the program is fully in effec:.

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4.0 DIESEL GENERATOR Public Service Company of Colorado requested a change to Fort St.

Vrain Technical Specifications, Limiting conditions for Operation of the ACM Diesel Generator in their letter dated August 29,1979 (P-7916 4).

'Specifically, the Technical ' Specifications state that the 10,000 gallons of fuel for the diesel generator provides ivr one week of operation of the generator.

Public Service Company of Colorado performed a fuel oil consumption test for the ACM diesel generator and determined that the initially established time that the 10,000 gallons of fuel oil would provide for full ACM load used a fuel consumption rate which was more conservative than that determined by the actual test. The actual test established the time that the 10,000 gallons of fuel oil will provide the diesel generator with operation at a full load of 900 KW or 41/2 days (108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br />) and not the originally established one week.

We have reviewed the proposed change and determined that the 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br /> is still sufficient time to obtain additional fuel oil from off-site sources.

We therefore, find the proposed change acceptable.

5.0 FIRE PROTECTION In the NRC Safety Evaluation Report supporting Amendment 21, June 6, 1979, the history of a three-stage fire protection improvement program initiated in 1975 was described together with approval of stage II impl ementa tion.

Stages I and II had been approved earlier on June 18, 1976 and October 28, 1977. One additional item outstanding was noted

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in the June 6 safety evaluttion. This was the revision of existing plant fire protection Technical Specifications to apply to other safety-related plant areas consisten't with the requirements of the Standard Technical Specifications dealing with Fire Protection.

By letter dated P-79170 dated August 13, 1979, the licensee proposed the following new or revised Technical Specifications:

1.

Specification LCO 4.2.6 - Fire Water System / Fire Suppression Water System, Limiting Condition for Operation.

2.

Specification SR 5.2.10 - Fire Water System / Fire Suppression Water ^3: tam, Surveillance Requirement.

I 3.

Specification SR 5.2.24 -

Circulating Water Makeup System, Surveillance Requirement.

4.

Specification SR 5.10.3 - Snake Detectors and Alarm, Surveillance Requirement 5.

Specification LCO 4.10.5 - Fixed Water Spray System, Limiting Condition for Operation.

6.

Specification SR 5.10.6 - Fixed water Spray System, Surveillance Requirement.

7.

Specification LCO 4.10.6 - Carbon Dioxide Fire Suppression System, Emergency Diesel Generator Rooms, limiting Condition for Operation.

8.

Specification SR 5.10.7 - Carbon Dioxide Fire Suppression Syttem, Surveillance Requirement. **

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Specification LCO 4.10.7 - Fire Hose Stations Limiting Condition for Operation.

10.

Specification SR 5.i0.8 - Fire Hose Stations, Surveillance Requirement.

11.

Specification LCO 4.10.8 - Yard Fire Hydrants and Hyd-ant Hose Houses Limiting Condition for Operation.

12.

Specification SR 5.10.9 - Yard Fire Hydrants and Hydrant Hose Hoses, Surveillance Requirement.

13.

Section Introduction 4.10 - Fire Suppression Systems - Limiting Conditions for Operation.

14. Section Introduction 5.10 - Fire Suppression Systems -

Surveillance Requirement i

These fire protection Technical Specifications were proposed in a format consistent with the fonnat of the existing Fort St. Vrain Technical Speci-fications rather than a format conforming with Standard Technical Specifi-cations. The licensees' Technical Specifications were reviewed with the aid of a step-by-step comparison with the Standard Technical Specifications, submitted by the licensee on September 28, 1979, and found to meet the intent of the Standard Technical Specifications. Our review concluded that the licensees' fire protection Technical Specifications are acceptable.

This review closes out the last remaining open. item in the fire protection program review of Fort St. Vrain.

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It must be noted, however, that while Surveillance Requirements

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5.2.10 - Fire Water System / Fire Suppression Water System and 5.2.24 -

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Circulating Water Makeup Syst,em are acceptabl e for fire protection

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purposes, they are not yet acceptable from the standpoint of meeting the 4

new inservice inspection and' testing requirements which the licensee is developing.

In its letter of November 30, 1979, these two systems were iden+,1fied by the licensee as Category II systems and as detailed their surveillance requirement will be substantially more detailed than now expressed in the Technical Specifications.

6.0 REACTOR PROTECTIVE SYSTEM SURVEILLANCE Public Service Company of Colorado requested a change to the r

Fort St. Vrain Technical Specifications, Surveillance and Calibration Requirements for Instrumentation and Control System (SR 5.4.1) in their letter dated March 23,1976 (P-76075).

Speci fically, Public Service Comcany of Colorado requested to change table 5.4-4, startup channel calibration fraquency (Item ic) 2 and method, and Table 5.4-1, startup channel test method (Item 3c).

The staff concludes that item Ic under " frequency" column of Table 5.4-4 is not consistent with the other portions of the Technical Specifications and should be changed as requested to provide consistency.

The proposed change does not affect considerations related to the health and safety of the public.

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i The staff has discussed the requested changes to item 3c under

" Method" column of Table 5.4-1 and item lc under " Method" column of Table 5.4-4 of the Techn.ical Specifications with the licensee. The staff and the licensee have agreed that the following sentence shall be added: "The interna'l test signal shall be checked and calibrated to assure that its output is in accordance with the design requirements.

This shall be done after completing the external test signal procedure by checking the output indication when turning the internal test signal switch." The staff requires this addition and concludes that the requirements of IEEE-279-1971 are satisfied and that these changes are acceptable.

7.0 SHOCK SUPPRESSORS, SNUBBERS The Fort St. Vrain Technical Specifications dealing with the Limiting Conditions for Operation of shock Suppressors or Snubbers state (Section 4.3.10e) that shock suppressors may be added to Class I systems without prior License knendment to Table 4.3.10-1, provided a revision to Table 4.3.10-1 is included with a subsequent License Amendment request. The Public Service Company of Colorado elected to update Table 4.3.10-1 listing all Class I snubbers as they are added as a result of refined seismic analyses of their piping systems.

l As recommended by IE Bulletin 79-14, PSCo has reviewed their " safety-i related" piping systems to verify that the deisgn drawings correspond to the "as-built" configurations.

As a result of this review and additional requirements, PSCo has started a reanalysis of their piping systems and additional snubbers will be added or deleted as the analysis dictates.

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8.0 DESIGN FEATURES TFe Public Service Company of Colorado requested a change to the Fort St. Vrain Technical Specifications, Reactor Core' Design Features, in their letter dated January 11, 1980 (P-80003).

During fuel fabrication. in" 1971, the coatings on the fissile particles were manufactured slightly thicker than specified and a problem arose in squeezing enough particles into a fuel rod to yield a thorium to uranium ratio of 4.25 to 1.

8.1 Background

On September 14, 1971, prior to final AEC approval of the then-proposed Fort St. Vrain (FSV) Technical Specifications, the FSV fuel design specification was modified to allow a decrease in the nominal Th:U ratio of fissile fuel kernels from 4.25:1 (+0.5) to 3.6:1 (+1.2, -0.2).

However, no action was taken to modify the Th:U ratio describeo in Technical Specification D.F.6.1, which contains a description of the reactor core design features, including those of the fissile and fertile fuel particles.

In the meantime, fuel with the 2.6:1 nominal Th:U ratio was used in the latter stages of FSV initial core production as well as in reload segments 7 and 8, and it is intended for use in future segments containing (Th/U)C2 fissile. fuel. The major impact of the decrease in Th:U ratio is a slight increase in the fissile kernel peak burnup for six-year-old fuel in FSV.

The peak burnup for 4.25:1 fuel is expected to be 20% FIMA (fissions of initial metal atoms) whereas the peak burnup for 3.6:1 fuel is expected to be 22.4% FIMA.

The overall ratio of Th:U in the core is unaffected by the change in tne fissile particle specification.

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. 8.2 Summary of Regulatory Evaluation The major part of the technological basis for permitting the proposed technical specification change was provided in the attachment to the February 8,1980 letter from Don Waremburg (PSC) to William P. Gammill (NRC).

In that attachment, fuel' performance under normal and design basis accident conditions was discussed in terms of the potential effects of the Th:U ratio change on fuel failure phenomena.

In each case, test reloits were used to support the assertion that there were no discernable effects.

8.3 Fuel Performance 8.3.1 Normal Reactor Operation For normal operation, three phenomena were considered:

kernel migration, fabrication defects, and pressure vessel failure.

In the case of kernel migration, coating failure is assumed to occur if a migrating kernel contacts the structural layers of a TRISO coating.

Since the coating dimensions are the same for 4.24:1 and 3.5:1 kernels, this phenomena could be affected only if the change in Th:U ratio were to cause a change in the rate of kernel migration.

However, test results indicate that the rate of migration is unaffected over a range of Th:U ratios from 4.25:1 to 1.60:1.

The change in Th:U ratio from 4.25:1 to 3.6:1 should not, therefore, have any impact on coating failure associated with kernel migration.

In the case of the second coating failure phenomenon, viz., missing or defective coatings, the small decrease in Th:U ratio does not result in a change in specified coating properties or coating pressures.

There fore,

the number of particles manufactured with missing or defective coatings would not enange, and coating failure and fission product release resulting from this phenomena would not,,be affected.

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, The third failure phenomenon, called " pressure vessel" failure, is a function of kernel burnup, coating design, and operating temperature.

From a, summary of irradiation test results provided wi,th the attachment to the February 8,1980 letter from PSC, it was observed that the failure probability for fuel irradiated to burnups 27% FIMA was the same (0.004) as fuel irradiated to 2.0% FIMA.

Hence, the irradiation test results indicate that the small increase in burnup (from a previous maximum of 20% to the current 22.4% FIMA) should have a negligible effect on pressure vessel failure.

8 S.2 Design Basis Accident Conditions The loss of forced circulation (LOFC) is the most severe accident, with regard to fuel performance, analyzed in the FSV FSAR.

Test results show that the major failure mechanisms include sic-actinide metal reactions, sic-fission product reactions, Sit decomposition.

In the FSV FSAR analysis of fuel performance during the LOFC accident, it was assumed that the fuel failure fraction was 0.05 for temperatures less than 1725'C and 1.0 for temperatures greater than or equal to 1725'C.

More recent (1977) data, however, obtained on irradiated particles subjected to a heat-up that was conducted out-of-pile, show that fuel failure should not become significant until temperatures exceed 2100*C.

Although that trst was conducted on particles having 18.2% FIMA burnup, the effect of a slightly higher burnup (22.4% FIMA) should be negligible relative to the FSAR assumptions.

As discussed in the attachment to the May 28, 1980 letter from PSC, heat-up tests of (Th/U)C and UC TRISO particles have snown that 2

2 the failure rate is relatively insensitive :0 burnup. We, therefore, conclude that decreasing the Th:U ratio frem 4.25:1 to 3.6:1 should not l

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8.4 Regulatory Position We have completed our review of the documentation provided in support of the request for a change to Fort St. Vrain Technical Specification D.F.6.1.

Based on our evaluation of the information supplied, we conclude that there is reasonable assurance that a decrease in the nominal Th:U ratio from 4.25:1 normal operation and design basis accident conditions and that the technical specification change is, therefore, acceptable.

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4 9.0 Accident Reanalysis 9.1.0 Background

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On November 1, 1977, the Public Service Company of Colorado submitted analyses in support of operation of thg Fort St. Vrain plant at 100% ci design power.

i The power level of the Fort St. Vrain plant was originally limited to 70% of design power because of limitations in the helium purification system which must be used for depressurization in the event of a loss of forced circulation accident.

These limitations were discovered during the review of the Alternate Cooling Method provided subsequent to the Brown's Ferry Fire and are addressed elsewhere.

In addition, a separate problem arose in that tests disclosed that firewater delivery to the circulator Pelton wheels and steam generators was insufficient to keep predicted temperatures at or below those originally reported in the FSAR.

PSCO justified, through analysis, that at a power level c7 70% of design power, temperature predictions would fall at or below the original FSAR values.

It was during these reanalyses that discrt-pancies between the values for core region peaking factors and outlet temperature dispersion used in the FSAR safety analyses and the values used in the plant technical specifications (which were higher) were identified.

Accident reanalyses using the more limiting initial operating conditions permitted by the technical specifications were then submitted in support of proposed full power operation for Ft. St. Vrain.

Additionally, the reanalyses for cores after initial refueling included the effects of pressure booster pumps wnich have since been installed in the firewater feedlines to the circulator pelton wheels.

This modification was required to provide suf-ficient circulator flow to maintain acceptable fuel temperatures for the firewater cooldown accident case with the reactor at full power.

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. The evaluatio.n which follows addresses'the accident reanalyses (Nov. 1977 submittal) in support of full power operation.

9.2.0 Licensee Analyses 9.2.'1 Scope The licensee has submitted analyses of three accidents which are considered to be the most limiting.

These are (1) Cooldown on one firewater-driven pelton wheel, (2) Rapid Depressurization/ Blowdown, and (3) Permanent loss of Forced Ci rculation.

All of these reanalyses were performed with the RECA3 code, which was not used for'the original FSAR analyses.

Differences in Technical Specification Peaking factors and Outlet Temperature Dispersion factors from those used for the original FSAR analyses are summarized in Table 2.1 below:

Table 2.1 Peakin< Factor Outlet Temo. Dispersion Original FSAR 1.78 54 F Technical Specification 1.83 250*F In support of the three bounoing accidents identified, the applicant submitted the results of a review performed for all accidents originally analyzed in the F5AR.

For those accidents affected by either Region Peaking factor or outlet temperaturs dispersion, a set of enveloping accidents was identified.

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16-affected accidents ar,e the Rod Withdra'wal accident, the orifice closure accident, and steam in-leakage events.

For the orifice closure accident, the conclusion that the original FSAR analyses were bounding was based on new data which showed that the fully closed o;rifice valve loss c', efficient was approximately i

1/2 of the value used for the FSAR.

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The staff has reviewed the enveloping logic and the results of the review and finds acceptable the conclusion drawn by the applicant that the three accidents identificd are bounding.

9.2.2 Analysis Methods All reanalyses were performed using the RECA3 code.

This code was not used to perform any previous analyses submitted to the NRC (i.e., for the FSAR).

While the staff has not reviewed the code for applicability on a generic basis, we have determined the code to be acceptable for the specific analyses performed for the Fort St. Vrain Plant (See Section 3.0).

Tne applicant has also used the TAP and RATSAM codes to predict the core heiium inle't temperature versus time and the system pressure versus time resoe'ctively, for input to the RECA3 analyses.

Comparisons of these code predictions to alternate calculational methods, as well as the sensitivity of analysis results to uncertainties in these parameters are also provided in Section 3.0.

For the three accidents analyzed, the plant was assumed to be operating at 105 percent of full power, ano 105 percent of full flow, with an initial power t: flew ratio of 1.0.

The applicant gas stated that actual power to flo.

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ratios may be in excess of 1.0 at indicated full power as indicated in Table 2.2.

In addition, plant technical specifications permit operation at power-to-flow ratios in excess of 1.05 at power levels below 100% (see Figure 2.1).

Maximum temperatures however, would occur for the 105 percent I

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power level case due to the increased decay heat generation.

This was i

confirmed by independent calc 01ations by ORNL at selected points along :he power-to-flow operating limit curve (Figure 2.1) as described in Section 3.0.

Table 2.2 Indicated Actual (worst case)

FSAR Assumotion 1

Power 100%

102%

105%

Flow 95%

93.5%

105%

P/f ratio 1.05 1.09 1.0 9.2. 3 Acceptance Criteria The thermal limits for acceptable response of fuel and structures to postulated accidents are those originally approved in the FSAR ano are provided iq. Table 2.3.

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-75 100 PERCENTAGE OF DESIGN CORE THERMAL POWER, P (l:)

FIc.H.D.E 2.1 POWER.TO. FLOW RATIO TECHNICAL SPECIFICATION CURVE FOR FORT ST VRAIN 1

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.. 4 Table' 2.3 Item Temperature Limit Fuel 2900 F t

Steam Generator Inlet Ducts and Liners 2000 F Upper Plenum Insulation and 1500 F Cover Plates These limits do not represent ooints at which physical damage of the fuel or structure will occur, but rather are temperatures above which degradation is expected to increase significa.itly.

9. 2. 4 Analysis cesults The results of the RECA 3 reanalyses are compared to the temperature limits of Table 2.3 in Table 2.4.

9. 3. 0 Staff Ev&1uation

9. 3.1 Methods Review Tre staff nas cetermined the acceptability of the applicant's analysis methods cy (1) evaluation of key input assurations to which the output is sensitive.

(2) comparison of the results of applicable plant transient temperature data c

emperature predictions for those transients using the RECA3 code,

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Table 2.4 Event Limit RECA 3 Prediction 0-delay Firewater Fuel,

2900 F

<2600 F Cooldown/ Initial Core Steam Generator Inlet 2000 F

~1600 F Oucts & liners Upper Plenum 1500'F

<<l500*F*

Insulation and Cover Plates Rapid Depressuri-Fuel 2900 F

~2600 F zation Blowdown Steam Generator Inlet Ducts & Liners 2000 F 1760 F pper Plenum 1500 F

<<1500 F Insulation and Cover Plates Permanent Loss of Fuel 2900 F

<2900 F Circulation (LOFC)

Steam Generator Inlet 2000 F

<2000 F Ducts & Liners Upper Plenum 1500 F

~1500 F**

Insulation and Cover Plates

^ Calculated temperatures were not reported by the applicant, since forced circulation is not lost and core inlet temperatures will remain close to the feedwater temperature.

"Tne top head liner temperature is calculated to not exceed the 1500 F limit provided the system is depressurized within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> af ter LOFC.

( O comparison of temperatures predicted by RECA3 to temperatures predicted by ORECA, and (4) comparisen of analysis code predictions to hand calculations.

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. The ORECA code, which predicts the tra'nsient behavior of gas-cooled reactort.

is similar in function to the applicant's RECA3 code.

ORECA was developed by ORNL for the NRC.

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Tne plant data used for code verification were from three reactor trips which occurred from power, and from'one event in which all forced circulation was lost for approximately ten minutes.

9.3.1.1 Input Assumotions As discussed in Section 2.2, the power-to-flow ratio used for all of the reanalyses was 1.05 and was confirmed by ORECA analyses to be the most limiting value.

The results of these analyses are proviced in Table 3.1.

In our review of initial conditions with respect to the allowable power-to-flow ratios in the technical specifications, it was noted that for limited periods of time, the technical specifications allow full power operation at power-to-flow ratios greater than 1.05 based on steady state time-at-temperature limits for fuel damage.

The licensee considers operation in this region to be a degraded plant condition and has stated that normal practice is act to coerate with power-to-flew ratios greater than 1.05.

Since operation i

nis degraded moda has nct been considered in the accident reanalyses, delioerate operation at power-to-flow ratios in excess of the curve shown as F# pure 2.1 (Figure 3.i-1 of the technical specifications) is not acceptable to tr.s staf' If the power-to-flow ratio limits of Figure 2.1 are exceeded, we require tnat the operator act promptly to bring the plant within I

~s _

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7 P

TABLE 3.1 Result s of OltHL Confirmatory Calculations Power-to-Flow katio Technical Specification

.i P wer Flow DBDA LOFC Case //

Initial 1 core inlet P/F

~l"F)

Mwt lbm/ min Max.

Max.

Max.

Fuel Gas fuel 1

1.05 768.3 100 842 95.2 54,760 2617 2313 2808

')

2 1.095 731.6 80 673.6 73.1 41,871 2335 2106 2555 3

1.14 706.9 60 505.2 52.6 30,084 2094 1928

' 2317 4

1.17 686.4 40 336.8 34.2 19,511 1923 1923 2072 4

n 5

1..a 734.2 20 168.4 18.2

~10,418 1826 1736 1840 6*

1.091 740.4 102 858.8 93.5 53,601 2627 2322 2826 7**

1.043 773.0 104.3 878.3 100 56,500 2676 2351 2858

• Worst-case operational conditions

• • Reference Case i

allowable limits.

We will require the Ticensee to propose technical specifi-cation revisions to conform to this position prior to approval of 100 percent pcwer operation.

ORECA analyses were also performed to investigate the sensitivity of cdiculated coolant temperatures to variations in some of the input parameters.

Results are listed in Table 3.2.

TABLE 3.2 Peak TGas Out, Hot Streak

• F Temo.,

F Reference Case 2269 1927 Helium Flow' (-20%)

2348 1995 Coolant Friction Factors '

(Laminar & Transition)(+20%)

2275 1926 Effective Coolant Heat Transfer Coefficent (-20%)

2252 1917 Afterheat (+20%)

2433 2034 From these studies, the most sensitive parameters were determined to be the helium flow through the core and the decay heat rate.

The decay heat rate used for the RECA 3 analyses is the same as the decay heat rate curve approved in the FSAR.

e*

m

,-w v

r 24-The bypass fraction assumed for the accident reanalyses was 7.5% of the circulator flow *.

There is some uncertainty in bypass flow because of the inability to directly measure flow, as well as the inability to measure flow path resistances and therefore, determine relative flow splits.

Calculation of the apparent bypass flow using the RECA 3 model indicated good agreement with the initial e.

. ate.

For the four scram tests to date, bypass fractions of

<r l

0.076 (40% power), 0.076 (50% power), 0.070 (60% power), and 0.063 (70% power) were calculated, which are in good agreement with the value of 7.5% assumed for the safety analyses. Moreover, analyses using ORECA indicate that even for bypass flow.ncertainties upward of 20%, stern generator inlet temperature limits will not be exceeded.

Based on the above, we find the use of a bypass fraction of 7.5% in the~RECA analyses acceptable.

9. 3.1. 2 Code Verification

9. 3.1. 2.1 RECA3

\\

The RECA3 comparisons to available scram data indicate that predictions of helium temperature in the maximum peaking factor refueling regions are in good agreement with the measured temperatures.

However, the code underpredicted helium temperatures in the north-west quadrant of the core by as much as 50 F tc 100 F in the 40-70 second time frame as shown in Figure 3.1.

This dis-crepancy may be due to excess bypass flow through fuel region gaps in this xThe design bypass flow, or that flow which does not enter the core barrel, is 2.9 percent.

The RECA analyses assume part of the unheated core ficw as bypass.

For these analyses, the bypass was input as 7.5 percent.

e l

n

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vv< a m l

1200 1*

x ucA3

.,-s a

%k DATA 1000 g

N N,

' o e i e<

t a

900 N

N s N 300 7g l%}N

=

a

=

,i w 700

,a w_ _+

il 5

}_l s

r 4., 600 8,

a E

g

=

[500 1

400 6

l 300 0

10 20 30 40 50 60 70 TIE (ninuits)

FORT ST. TRAIN SCUM FROM 39 PERr.ENT POWER

• an Ic/zs/n us1ON 35 FIGURE 3.1 l

a 1'

26-

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quadrant.

Such observations are consistent with region outlet temperature fluctuation pheno.aena observed during plant operation.

The fluctuations were most prominent in this region, and are believed to be due to the opening and closing of axial gaps between fuel blocks.

The dis:repancy between the predicted and measured region outlet temperatures is of concern to the staff. We will therefore require that the applicant perform at least one verification transient subsequent to corrective action taken to eliminate the core fluctuations.

This transient can be a reactor trip from power, and the verification should con? st of comparisons of i

measured to predicted region outlet temperatures.

Acceptable

• predictions of the measured data, including resolution of the previously observed northwest quadrant discrepancies, will be required before full power operation is allowed.

Alternatively, the licensee should identify an acceptable operating pcwe* level, based on accident analyses in which this uncertainty has been properly accounted for.

Ccmparisons of ORECA predictions to the plant trip data showed good agreement between the calculated and measured region outlet helium temperatures.

In acdition to the comparisons made to plant data, peak helium temperatures were a'sc predicted for two of the three bounding accidents; the Design Basis Deoressurization Accident (D8DA) and the Firewater Cooldown Accident (FWCD)

.#th a zero time-delay assumed for the initiation of firewater cooling.

For a precictive uncertainty should not be abnormally excessive for any efueling region when ccapured to the average.

~ ~ ~

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the firewater cooldown accident, predictions were made for two core loadings, equilibrium, and initial.

The results of these calculations compared to the applicant's predictions are given in Table 3.3.

s TABLE 3.3 t

RECA3 ORECA Event Prediction Prediction 0 - Delay Firewater Peak average gas 1525 F 1509 F Cooldown/ equilibrium outlet temperature Core from core Peak gas outlet temperature for maximum region 1900*F 1873 F 0 - Delay Firewater Peak average gas Cooldown/ initial outlet temperature core from core 1500 F 1479 F Peak gas outlet temperature for maximum region 1900*F 1901 F Gesign Basis Depressuri-Peak average gas i:ation Accident / equi-outlet temp. from librium core core 1700 F 1724 F Peak gas outlet temp.

2350 F 2269 F for maximum region Peak fuel temoerature 2600*F 2557 F

9. 3.1. 2. 2 RATSAM Code Tne RATSAM code is used to predict system pressure versus time as input to the RECA3 calculations.

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O ~

In ceder to assess the effect of uncertainties of the calculated pressure on the RECA3 calculated temperatures, the applicant performed both hand calcula-tions of the transient pressure as well as RECA 3 reanalyses using a constant helium pressure of 700 psia.

The hand calculations of the transient pressure showed agreement with the general trends of the RATSAM-calculated pressure during the first hour af ter accident initiation.

However, the RATSAM-calculated pressure was shown to sligi tly increase after one hour whereas the hand-calculated preessure con-tinued to decrease beyond one hour.

To show that the effect of calculated pressure uncertainties did not have a large effect on the results, the applicant performed reanal 1es with RECA3 3

assuming a constant 700 psia system pressure.

These reanalyses were for the twi accident analyses which require RATSAM input; the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the LOFC (prior to initiation of depressurization) and the first 1-1/2 hour delay of firewater to the pelton wheels. '

Tre main result of the LOFC : analysis assuming ccastant system pressure was that the time for the top head thermal barrier aveiege cover plate temperature tc reach 1500*F was reduced from 25 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The analysis also showed t st the top head liner remained intact for both cases beyond 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Ine results of the reanalysis of the FWC0 with a 1-1/2 hour celay assuming c:nstant system pressure showed that some temperatures in the core were re: ced oy 10 to 30 F frcm the case nere system was calculated by the RATSAM 9

i^

i

. code.

The peak fuel temperature was a' Iso reduced by 53 F for this case.

However, the average upper plenum temperature was shown to increase 138 F (to 1350 F) and the average PCRV top head thermal barrier cover plate temperature increased by 113 F.,(to 1152*F) at 1-1/2 hours.

In neither case were the temperature limits of Table 2.3 exceeded.

The applicant has demonstrated that the general trend of initial pressure raduction predicted by the code is supported by hand calculations, and that with the assumption of a constant 700 psia system pressure (approximately 100 psia greater than the RATSAM predicted pressure), temperatures in the core, fuel and structures did not change the results significantly.

Based on the above, the staff finds the use of RATSAM code acceptable for the purpose of predicting system pressures for the two FSV accidents analyzed.

9.3.1. 2. 3 TAP Code The TAP code is used to calculate the temperature of the helium exiting the steam generators and entering the upper plenum and core as input to the RECA3 calculations.

Although the TAP code has not been verified against data, hand calculations were performed by the licensee to confirm the calculational accuracy of the TAP code.

These comparisons are provided in Figure 3.2 and show the TAF calculations to be in good agreement with the hand-calculated values of helium inlet temperature.

Moreover,.ne applicant stated that because of the excessive heat transfer ca:amility of the steam generators at cecay heat levels, coupled with the low c:re flow assumed subsequent to accident initiation, the helium temperature at i

COMI'AltlSON OF I Ap AND llAND CALCULATED FOllT ST. VilAIN COllE INI.lil lil l illM 1 EMPEll Al ullE UNDEH FiflEWATEll COOLDOWN CONDITIONS

/00().

Inilial Comlitnms at Time of Accident 600 X

llamt Calculated Points Based on Accident Conditions at Indicated llours.

500 L

N:

g

. 5 400 g

5 t-E 300 h

TAP Input to RECA (Assumes 80*F Firewater)

,~&

I Fiiewater Inlet Temperature to SG = 120*F 200 g Fire r inlet Temperature to SG = 8(TF fland Calculated Curves p

/

-x

_2 X

4 _.

x i0o Constant 120* F Beyond 2 fluurs l

I I

I I

I I

I I

I o

O 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 18 2.0 Time (llours)

FIGURE 3.2

= the exit to the steam generators will 'be approximately the same ss the feedwater temperature, and therefore should not be a highly sensits) para-meter.

Based on the above considerations and the confirmatory hand calculations, the staff finds the use of the TAP code acceptable for the purpose of calculating the helium temperature exiting the steam generator.

9. 4. 0 Summary The staff has reviewed the accident reanalysis submitted by the licensee in support of operation of the Fort St. Vrain plant at 100 percent of design power.

Based on our review, we have concluded that the reanalyses provided are acceptable to justify full power operation.

However, prior to operating at any power level above the present 70 percent restriction, the licensee must perform the following:

N 1.

Provide for staff review ar.o approval a minimum of one additional RECA3 code verification analysis of plant transient response.

The transient response used for verification must be performed subse-quent to corrective actions taken to eliminate the core fluctuations.

Alternatively, an acceptable power level should be proposed which is based on accident analyses which account for this prediction uncertainty.

2.

The licensee must propose, for staff approval, revisions to the plant technical specifications which will specifically preclude cperation at power-te-flow ratios in excess of those for which the plant transient response has been shown to be acceptable.

I l

i l

10. ' Conclusions Based on our review of the documentation referenced in this report, an s

evaluation of plant operations husfar, 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. Tnis safety report describes ine basis for this conclusion, and notes the conditions which apply.

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 the NRC staff.

As related in Amendment No.18, three items form a formal hold to operation of the reactor above 70% power: (1) Depressurization, (2) Moisture Monitors, and 3

(3) Accident Reanalysis.

This ti;ird item has been reviewed and our comments f

and conclusions presented in Section 9 of this Safety Evaluation Report.

Items I and 2 are still under NRC review and should be resolved in the near future.

Investigations into the power / temperature fluctuations continue at both the NRC j

and PSCo.

Public Service Company of Colorado plans to perform acceptance tests of the installed Region Restraint Devices (Luci Locks) sometime in the fall cf 1990 to demonstrate their ability to eliminate the fluctuations.

The licensee l

plans to modify the Buffer Heluim System to provide two separate heluim lines, t

[

one for each loop.

Is is' anticipated that this split will eliminate most of i

the buffer-mid-buffer upsets experimenced at the Fort St. Vrain reactor.

The physical split of the loops is planned for the next refueling outage, tentatively scheduled for May 1981.

l

-e--

e-e--

-i,--v---

o

, These planned activities do not detract from the conclusion that tne plant may be operated at 70% of rated power.

Finally, the NRC staff requires and expects the licensee to proceed expeditiously to resolve those matters as described in this safety evaluation report, and to also expeditiously complete all work required for 100% power operations.

P 89 9

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