ML20214W452

Review of Proposed Tech Spec Change:Core Inlet Orifice Valves/Min Helium Flow & Max Core Region Temp Rise, Technical Evaluation Rept

ML20214W452
Person / Time
Site:	Fort Saint Vrain
Issue date:	10/16/1986
From:	Ball S OAK RIDGE NATIONAL LABORATORY
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
Shared Package
ML20214W440	List: ML20214W436 ML20214W452
References
CON-FIN-A-9351 TAC-52634, NUDOCS 8612100231
Download: ML20214W452 (17)

Text

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o ENCLOSURE TECHNICAL EVALUATION REPORT FORT ST. VRAIN NUCLEAR GENERATING STATION

. DOCKET 50-267 .

LICENSEE: PUBLIC SERVICE CO. OF COLORADO REVIEW OF PROPOSED TECHNICAL SPECIFICATION CHANGE:

CORE INLET ORIFICE VALVES / MINIMUM HELIUM FLOW AND MAXIMUM CORE REGION TEMPERATURE RISE (L.C.O. 4.1.9) s PREPARED BY:

S. J. Ball Oak Ridge National Laboratory Oak Ridge, TN. 37831 l .. .

October 16.- 1986 NRC Lead Engineer: R. E. Ireland - RIV Project: ORNL Assistance in Evaluating Licensing Request-PSV LCO 4.1.9 (FIN A9351) l l

8612100231 DR 861205

( ADOCK 05000267 l PDR l

NOTICE This report was prepared as an account of work sponsored by an agency of the United Statec Government. Neither the United States Government nor any agency thereof, or any of their duployees, makes any warranty, expressed or implied, or assumed any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report or represents that its use by such third party would not infringe privately owned rights.

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3 REVIEW OF PROPOSED TECHNICAL SPECIFICATION CHANGE:

CORE INLET ORIFICE VALVES / MINIMUM HELIUM FLOW AND MAXIMUM CORE REGION TEMPERATURE RISE (LCO 4.1.9) 2 ENTRODUCTION The objective of this task is to provide NRC Region IV with technical and analytical support in their evaluation of a request by Public Service Co. of Colorado (PSC) to amend the Fort St.

Vrain (FSV) High-Temperature Gas-Cooled Reactor (RTGR) Technical Specification - Limiting Condition for Operation (LCO) 4.1.9.

3 The intent of LCO 4.1.9 is to ensure that during low power and low flow operating conditions (0-25%), core region temperatures will be limited to acceptable maximum values. The major basis for the concern is that at low core flows (and hence low core pressure drops), the effects of higher buoyancy forces of the pressurized helium coolant channels may lead to flow stagnation and reversals-in some channels. The uncertainties of the region heat removal processes under these circumstances make it desirable to ensure that region flow stagnation and reversals do not occur. The objective of the original LCO 4.1.9 is to specify a set of conservative operating limits for both startup and shutdown, hot and cold, and pressurized and not. NRC, PSC, and GA Technologies (GAT) have all identified problems with the consistency, accuracy, and conservatism of the original and interin technical specifications. It was concluded that an independent analysis should be done to provide a basis for the licensing action required to resolve questions about the l

operating limitations. Resolution of acceptable operating limits may result in changes to the FSV Technical Specifications in order to ensure conservative thermal margins.

, APPROACH The approach taken to help resolve the questions raised in determining acceptable operating limits made use of an existing ORNL code (ORECA-FSV), which calculates the dynamic thermal hydraulic behavior of the FSV core (Ref.1). The problems of flow stagnation' and core overheating were explored for a variety of representative and conservative startup and shutdown scenarios, in some cases requiring that special routines be added to the code. The objective was to determine if LCO 4.1.9 and/or accompanying tech specs provided adequate protection for all forseeable circumstances of plant operation.

The FSV version of the ORNL ORECA code has been used extensively in code verification studies, and, in general, has shown good agreement with both FSV data and calculations by the GAT RECA code (Refs.2-3). In simulating typical startups, the i

ORECA calculation begins with n zero -power uni form t emperat ure core and follows user-input time-varying functions of total

T. .-

4 circulator' flow, thermal power, primary system pressure, and core inlet temperature. Guidelines for typical startup scenarios were initially obtained from the FSV DC-5-2 (Issue C) manual both for startup from refueling conditions and for startup with full helium inventory. Subsequently, plant data logger records were obtained from PSC to get representative information on startup and shutdown' operating procedures, to check data consistency, and to-determine how close the operating parameters approach the prescribed. limits (Ref.4). For shutdowns, the code requires inputs specifying the power and flow rundown conditions. In both

cases, orifice manipulation routines are executed to go from

, approximately equal-flow to equal region temperature rise

settings, or vice versa, at specified times. Other user inputs include the refueling region peaking factors and orifice positions and the various core and refueling region bypass flow s fractions.

A watchdog routine was added to ORECA to detect violations of the original LCO (both for LCO 4.1.9 Fig. I and 2 conditions),

, noting the beginning and ending times for the violations and, for 4

the Fig. 2 case, the value of the maximum region temperature rise. An additional watchdog routine was added to look for violations of LCO 4.1.7, which governs adjustments of the core inlet orifice valves, as this turned out to be more effective in limiting core temperatures in many cases than did LCO 4.1.9. .

The ORECA code includes a model of the dynamic response of

• the region outlet thermocouples, which have fairly long response times -- especially at the low flows associated with startup and 1

shutdown (Ref.5). Calculations for LCO 4.1. 9 of core thermal power and region temperature rises are made based on these simulated thermocouple measurements rather than " actual" region outlet temperatures, since the measurements are used by the j ,,

, operators to determine compliance.

Typical startup and shutdown runs were studied in some i detail. Magnetic tapes with plant data logger outputs for the

! November 3, 1983, startup and the January 17, 1984, shutdown were l adapted for use on ORNL computers, along with PSO's " HISTORY" i program for reading, deciphering, printing, and plotting the i

data. PSC also supplied calculated region peaking factors (RPFs) at crucial points in the runs so that the ORECA code could be set up to simulate the runs (Ref.4).

The ORECA code was set up and run for major " stopping

, points", and good agreement between the steady-state calculations and data was found. In each case the agreement was optimized by varying the assumed core bypass flow fraction, and the optimized bypass flows were well within expected ranges.

The PSC HISTORY code was modi fi ed to do further investigntions of possible problems with tech spec limitations.

5 The LCO 4.1.9 and 4.1.7 watchdog routines added to ORECA were adapted to HISTORY and run with the startup and shutdown data.

For LCO 4.1.9, flow " margins" (actual core flow / minimum allowable flow) are output when LCO 4.1.9 Fig. 1 (equal-flow orifice settings) is applicable, and region temperature rise " margins"

'(maximum allowable delta-T minus worst-case measured delta-T) are output when LCO 4.1.9 Fig. 2 (orifices anywhere) is applicable.

For LCO 4.1.7, the worst-case margin (maximum allowable region outlet temperature minus the worst-case measured outlet temperature) is calculated both for the startup case (average core Fig. 4.1.7-1.

outlet (950 F) and for the conditions (>950 F) specified by In the latter analysis, all region outlet temperature readings are taken at face value rather than using .

, comparison regions for some, as is done in more recent versions of the tech spec.

A major effort involved model and code development to include intra-region flow dynamics in the ORECA+ code. This i

addition was needed to determine the limits of flow stability and stagnation more precisely than was possible by representing each refueling region _by an average channel. The model that was

' developed simulates the two worst-case regions, using the computed overall core pressure drop, inlet plenum gas temperature, and power as inputs. Each of the two regions is represented by a single high-power density column (where the

" tilt" power factor may be as high as 1.6) in parallel with the average-power-density combination of the other six columns in the '

region. The common " upper plenum" for these two column models is the space just below the inlet flow orifice. As in the ORECA model, each column model is represented by six fuel, three reflector, and one core support axial nodes. Radial conduction is included but axial conduction is conservatively neglected.

The programs required for implementing this model were largely

. drawn from existing ORECA coutines. The three major advantages i ..

of this technique over the GAT steady-state models are the j ability to simulate the dynamic situation (which allows estimation of limits on allowable times out of compliance), the ability to use a dynamic overall core pressure drop driving function that is derived from a detailed 37-region calculation. ,

and the ability to determine if and how stagnation conditions

! could be corrected.

BENCHMARKING CALCULATIONS FOR INTRA-REGION FLOW LIMITING CASES

' Since the limiting factor in many low power operating regions is the calculated stagnation point for intra-region flow, i

and since the ORNL and GAT approaches to the problem were so different, a set of benchmark calculations were set up by Rick Kapernick of GAT. Some apparent discrepancies had appeared, as well, when in many cases the opernting limits dictated by GAT l analyses were more restrictive than ORNL's. It was decided that

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i for all benchmark runs, there would be two worst-ense refueling

6 regions. The hotter region had a region peaking factor (RPF) of 3.0 and a column tilt factor of 1.17, while the other had an RPF of 1.6 and a tilt of 1.507. The assumed active core bypass fraction was 0.18625, and all region orifices were set at 20%

open. Runs were made at two different power levels, 1% and 5%,

'with a fixed helium inlet density corresponding to 250 F and 367 psia at 1% power, and 350 F and 419 psia at 5% power. Total reactor flows were varied between 1 and 10% for the 1* power case, and 5% and 13% for the 5% power case. The ORECA code, which normally includes intra-region conduction, was run both with and without it. (Conduction is not included in the GAT analyses, and this accounted for some discrepancies).

Several other important features of the analysis besides the yes/no determination of stagnant or reverse flow conditions in a region were noted. These are:

1) How hot does the fuel get, with or without stagnation?

2) How long does it take for the fuel to heat up and for the regions to stagnate?

3) Could the operator tell from region outlet temperature readings if the worst-case regions are in trouble, and are the tech spec limits on outlet temperature mismatches violated?

4) What are the effects on flow stagnation of setting the orifices for equal region outlet temperatures (rather than equal positions)? ~

5) How readily can stagnation conditions be remedied by

• increasing flow or adjusting orifice positions?

Also important is the fact that there are relatively large errors in measuring power and flow at the very low values of each; therefore allowances need to be made for these when operating limits are set.

The ORECA runs for the benchmark were set up with a high flow initially (10% for the 14 power cases and 13% for the 5%

power cases), and the flow was subsequently reduced to the lower limits. For the 14 power reference benchmark case, there was no problem with either total region or intra-region _f1_ow stagnation with flows as low as 5%. After about 4 hr operation at 4% flow, however, the intra-region flow in the region with the higher tilt factor stagnated. The measured gas outlet temperature of the region with the higher RPF, although not stagnated, exceeded the mismatch temperature limit (average + 400 F) after about 1.5 hr.

With 3% of full mass flow, flow stagnation / reversals occur right away, and the maximum fuel temperatures reached the 1600 C

" damage limit" after about 4 hr. After as long as about 7 hr, by adjusting the orifices using a simple algorithm that attempts to equalize outlet temperatures, the stagnation can be cleared, and the core conditions can be recovered to acceptable temperatures and flows. In another case in which the flow was reduced to 2%

(and stagnation / reversals occurred), a subsequent increase in the

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flow to 5% cleared the stagnation. Runs in which the intra-region conduction was neglected showed some effect on the stagnation threshold. Typically about 1* nore flow was required to prevent stagnation without intra-region conduction than for cases with normal conduction included.

GAT's calculated minimum flow to prevent stagnation at 1%

power was 6.9%, vs. 5% per ORECA. The GAT analysis was done for 4

a higher density gas (107.5% vs. 90% inventory in the benchmark) and neglected conduction (a 14 flow effect), so these differences 4

could easily account for the discrepancy. A comparison of the predicted benchmark core conditions at 10% flow also showed no d

differences. .

A variation on the 1* power benchmark runs was made using ORECA for what we had judged (based on a GAT design support physics analysis) to be more realistic worst-case estimates for 1 RPFs and tilts, i.e. RPF = 1.80 and tilt = 1.36. These also

showed intra-region flow reversals when the total flow was 4

reduced to 44. However, if the outlet temperature equalization scheme is used to reposition the orifices, the reversals didn't occur until the flow was reduced to about 24.

For the 5% power benchmark cases, excellent agreement between ORECA and GAT calculations was obtained for 134 flow, although ORECA showed an outlet temperature mismatch exceeding

• LCO 4.1.7 even for those conditions. For subsequent reductions s in flow, ORECA initially showed no reversals for core flows as low as 5% (7% neglecting conduction), although very high core and gas temperatures were calculated. The GAT analysis (at 107% vs.

904 inventory) gave 10.3% as the minimum flow to avoid stagnation. Further investigation, however, showed that ORECA

, probably would have eventually calculated flow stagnation at flows higher than 5-7%. The ORECA calculations had been i - ' .

terminated when outlet gas temperatures exceeded 3000 F and fuel temperatures exceeded 9000 F; gas flows in the critical channels were still' decreasing. Hence the apparent discrepancy between GAT and ORNL calculations at 5* power were judged not be

,' significant. It should be noted, however, that the use of a flow stability limit to prevent"high fuel temperatures under these

circumstances is clearly not appropriate.

I For cases in which our "more realistic" worst-case RPFs and i

tilts were used. ORECA runs indicated that no stagnation occurred at 5% flow and that orifice manipulation was able to give reasonable core temperatures.

l Further variations on the benchmark runs were studies to investigate potential problems with the use of equal outlet temperature orifice settings at the low power / flow conditions.

At 1% power, for the minimum peaking factor region (RPF = 0.4 and

, column tilt = 1.507), which is limiting for the equal-outlet l

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-w ~ , - c.--e, err.ve,,ew.-e .,-r,,.,,---.--,.,m,--,r--_,-vw-.,w-4~-.--w-m-.r_,-% ,,,--.-,mvy,--,37-r--,,--, ,-,--n p.------,--

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8 temperature mode, stagnation occurred at 44 flow. However, by simply limiting the minimum orifice setting to 10% open, flow stagnation did not occur until the flow was reduced to 24.

The conclusions drawn from the benchmarking exercise was

'that the ORNL and GAT predictions were consistent, and that limiting conditions predicted by GAT would be at least as conservative as ORNL's.

SUMMARY

OF ORNL RECOMMENDATIONS. COMMENTS. AND CONCERNS This task (A9351) was initiated in Dec. 1983 at the request of the FSV project manager, P. C. Wagner. During the course of the study, a large number of recommendations were made and concerns noted. Most of these have been resolved or accounted for either by discussions with PSC and NRC, by further analyses, or by eventual modifications to LCO 4.1.9 and supporting surveillance requirements. A list of meetings attended by ORNL is in Attachment 1. The following summary is a chronology of the more pertinent issues and questions raised, where the eventual 4

dispositions are labeled by:

(R) -

resolved satisfactorily; (0) - ORNL was overruled (recommendation not followed); or (F) - some followup work is still recommended. *

1. Low-flow and low-power measurements (F) -

The definitions ~for core flow and core power (which are used in the LCOs to determine operating limits) are not adequate at very low power and flow. For example, the tech spec does not specify how core flow is to be derived, and the various means for calculating it from the instrumentation available have given widely-varying estimates at low flow, especially when one or more circulators are shut down. We recommend using circulator speeds in conjunction with air culator performance maps. We also recommend incorporataug calculated afterheat in the power estimate when approaching or following shutdown.

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2. " Gap" in coverage of operating conditions (R) -

In the figure limiting core outlet termpearture mismatches (for average T-out > 950 F), Fig. 3.2.2-1, there were gaps in the coverage for typical startup operations. These occurred when T-out > 950 F and the average temperature rise from the circulator inlet to core outlet was less than GGO F. From the data we observed, all the region outlet temperatures are typically low i enough so that there would not be a real problem with overheating the core.

- - - _ . - _ _ _ , _ _ _ - _ . _ . _ _ _ _ _ . ~ - _ , _ _ _ _ _ . _ , _ _ _ _ _ _ _ _ - - . _ _ _ _ _ _ _ _ _ _

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3. " Mechanized Calculations" of optimum orifice positions (F)

We recommended that PSC should use a simple program to calculate optimum orifice positions for each desired operating condition rather than let the operators find them with a slowly converging iterative process.

4. References to " Sister" LCO 4.1.7 (0) -

Apparently the relationship between LCOs 4.1.7 and 4.1.9 has not been universally well-understood. The limits imposed by LCO 4.1.9 are calculated entirely on the basis of avoiding operating conditions in which refueling region (or sub region) flows would stagnate or reverse. The limits of LCO 4.1.7 are somehwhat more direct, as they relate the inferred maximum fuel temperature to measurements of region outlet coolant temperature. Both LCOs are needed to effectively limit maximum fuel temperatures. Simply assuring no stagnation (LCO 4.1.9) doesn't assure that fuel won't overheat. For LCO 4.1.7, if the flow is stagnated, the region outlet thermocouple may not be measuring a temperature that can be related to fuel temperature. Hence, we recommended a closer tie-in of LCO 4.1.7 (it is now referenced in the Basis section).

However, as an operator guide, we feel that cross-references should be spelled out clearly. For example, in Table 4.1.9-1, core temperature rises are not shown as being limited for equal region flow cases where the system pressure is >SO psia. We '

believe it should be noted in the table, or at least in a +

footnote, that these cases are to be limited by restrictions in LCO 4.1.7.

5. Surveillance vs. Corrective Action time requirements (H) -

The time limits are required corrective action and the surveillance time intervals are not consistent. For example, with surveillance required only every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the alternative corrective actions are required in either 15 min or one hour.

(The 15-min limit implies an urgency not carried by the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance time interval.) PSC noted that operating procedures called for essentially continuous monitoring of the power, flow, * *

• and core outlet temperatures during orifice maneuvers. While this is a satisfactory response, we would prefer that reference to the more frequent monitoring be made in the LCO to assure the operator's understanding of the requirements. SR 4.1.~8 formerly~~

indicated only that flow was to be monitored continuously during power level changes. It has since been revised to require continuous monitoring during orifice maneuvers as well.

G. Questions on procedures for determining " bulk core temperature" limit of 760F (R)

These questions were addressed in detail and reported in our A9351 monthly repori for January 1986 (See attachment 2). We

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10 recommended that reference be made in the Basis to the procedures used to calculate bulk core temperature.

7. Equal-Flow orifice range specification (R) -

The LCO did not specify a limiting range for orifice positions corresponding to the equal flow setting mode. We recommended 10-20% open. The opening size is crucial to the flow stagnation calculations.

8. Experimental Verification (0) -

While benchmarking exercises for GAT and ORNL flow-stagnation analyses showed good general agreement, more comfort could be derived from some good experimental confirmation data.

In simulations of numerous startup tests, it was observed that for wide variations around normal operating paths, very little flow redistribution occurred. Redistributions,'which are precursors to stagnation, were shown to be readily observable by measuring changes in outlet temperature dispersions due to changes in total primary loop flow. The tests which were recommended proposed relatively small flow perturbations (within current tech spec limits). These tests would provide data on -

whole-region, not intra-region, flow redistributions.

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SUMMARY

AND CONCLUSIONS s.

The approach used in the review of the final and previous versions of the revised Tech Spec LCO 4.1.9 " Core inlet orifice valves / Minimum helium flow and maximum core region temperature rise" included the following:

1) Revise the ORNL ORECA (3-D FSV core thermal-hydraulics) code

'* as required to include intra-region flow and to simulate startups and shutdowns with both representative and conservative l assumptions. Benchmark calculations using ORECA and the GAT codes used to derive the limits employed in the new LCO showed good agreement.

2) Use FSV-supplied startup/ shutdown plant process computer

! data and PSC's " HISTORY" code to study both typical and conservative transients and to note operational problems.

3) Confirm that the revised Tech Spec meets its goals.

j 4) Point out problems and suggest alternatives.

As in the case of previous versions of LCO 4.1.9 the limits imposed by the final LCO (P-86451) range from equivalent to conservative as compared to t h a.s e derived by ORNL analyses.

Hence, the major concern of the Tech Spec, that of providing j limits that will in fact prevent core overheating, has been

11 addressed and confirmed satisfactorily. Remaining disagreements and concerns were primarily with details of clarity and style.

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o 12 LCO 4.1.9 (TER)

REFERENCES

1. S. J. Ball, ORECA-l? A Digital Computer Code for Simulating the Dynamics of HTGR Cores for Emergency Cooling Analyses, ORNL/TM-5159 (April 1976).

2. S. J. Ball, " Dynamic Model Verification Studies for the Thermal Response of the Fort St. Vrain HTGR Core",

Proceedings of the Fourth Power Plant Dynamics. Control and ~

Testing Symposium, Gatlinburg. Tennessee (March 1980). -

3. S. J. Ball, et al., High-Temperature Gas-Cooled Reactor Safety Studies for the Division of Reactor Safety Research Quarterly Progress Report. January 1-March 31, 1978, NUREG/CR-0179, ORNL/NUREG/TM-221 (July 1978).

4. Letter from D. W. Warenbourg (PSC) to S. J. Ball, " History

.. . Tape Transmittal for LCO 4.1.9 Evaluations", March 2, 1984 (P-84073).

5. S. J. Ball, et al., High-Temperature Gas-Cooled Reactor Safety Studies for the Division of Reactor Safety Research Quarterly Progress Report. October 1-December 31, 1978, NUREG/CR-0716 ORNL/NUREG/TM-314 (April 1979).

LCO 4.1.9 (TER)

ATTACHMENT 1 Meetings on LCO 4.1.9 Proposed Changes attended by ORhL s

1. August 23, 1984, at NRC-Region 4 Arlington, TX, with NRC, PSC, and GAT, to discuss the status of the review.

2. March 13, 1986, at NRC-Region 4 Arlington, TX, with NRC, PSC, and GAT, to address and resolve the oustanding issues on the most recent PSC drafts of LCO 4.1.9.

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ATTACliMENT 2 LfJi,1 tJSilHL ( W .',',1 Monthly f<epor t + or Januar , 1*/ tit. PAGt: I REVIEW OF LCO 4.1.9 BASIS RELAllNG lu /69 F AVERAGE CURE IEMPERATURE LIM 11 The basis an the draft LCO 4.1.9 (P-85442) pertaintny to the 760 F mansmum core temperature lamat as as follows:

"The calculated bu l l- core temperature is the calculated average temperature of the core. Including graphite and fuel but not the reflector, that occurs following a loss of all forced circulation of primary coolant flow. The calculation assumes that all decay heat as retained in the core with no heat transfer to the reflector, PCRV anternals or primary coolant. If the decay heat as sufficiently low, with all primary coolant flow terminated, the calculated bulk core temperature will not exceed 760 degrees F, this specification as not applicable. Below this temperature, there is no damage to fuel or PCRV internal components."

The 760 F bulk core temperature limitation as proposed for use in this and other revised LCOs as a means of assuring that no fuel or PCRV-internals damage will occur for periods when no primary coolant forced circulation is available. The 760 F limit was apparently derived from the design value of core inlet temperature, which is 760 Ft hence there is no safety concern af a!! of the components in the reactor vessel are nominally Inmited to this temperature. Several' questions and concerns do arise, however, regarding the implementatiot of this Inmit to specific reactor operation scenarios:

1) How is the afterheat calculated?

2) How hot do critical PCRV-internal (metal) components get with a core average temperature of 760 F7

3) How conservative is the assumption that there as no heat loss to the surroundings during a no-flow heatup?

ITEM 1: The means of determining afterheat is not specified in the LCO, and it should be. Since there are no sensors in the core which can effectively measure the mean temperature, it is important to have an accurate estimate of afterheat. This can be a complex calculation for cases where the power level has undergone maior perturbattons in _ _.

id"2the period before shutdown. (We have developed such an algorithm that '

could be implemented on the plant computer or a programmable calculator.)

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ITEM 2: A variety of heatup conditions were simulated using the severe-accident verston of the ORECA-FSV code, ranging from relatively rapid heatupu (6 days after 100% power operation or 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after 35%

power operation) to very slow heatups (1 year after 100% or 100 days after 35%). As expected. the nonuntformattes in PCRV temperatures were l arger in the faster heatuput however, at the time when the averaqo core temperature reached 760 F. mantmum fuel temperatures were well below normal operattnq temperatures and PChV metallic component A 1

ENCLOSURE - A9351 Nonthly Report for January 1986 PAGE 2 temperatures were well below 760 F. Hence we would conclude that the 760 F core average temperature limit is sufficiently conservative.

ITEN 3: Also of interest was the extent of the conservatism in the assumption that there is no heat loss from the core during thu heatup.

Again, by use of the ORECA-FSV code, it was shown that this conservatism is strongly dependent upon the heatup rate (the slower the rate the more conservative the assumption). The " actual" computed rates as percentages of the adiabatic heatup rates are shown for three representative cases in Fig. 1. The cases in which there is a relatively short time to restore circulation are also the cases with the least conservatism.

We would also recommend rewording the next-to-last sentence in the " basis". The way it currently reads, it implies that LCO 4.1.9 flow requirements are waived only in those cases where the afterheat is SO low that an adiabatic core would never reach 760 F.

In conclusion, it appears that the 760 F limitation is a sufficiently conservative means of protecting the' core and PCRV internals from damaget however, the means for Con,Juting the adiabatic temperature rises should be specified in the LCO or its basis.

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Figure 1 - Comparisons of Actual vs. Adiabatic Core Heatup Rates oRNL DWG 86C 6189 ETO

. l l l l u 100 - *- -

N'

- =E O u <

< 80 yt h

.K -

si y 60 -

= *. -

9 f 40 -

, h -

E .

20 -

t 0 ' I 300 400 500 600 700 800 '

AVERAGE 8ULK CORE TEMPERATURE (*F)

NOTES:

CURVE 1 -

• FAST HEATUP RATE *:

(100% POWER + 6 DAYS OR 35% POWER + 3 HOURS).

HEATUP FROM 300*F TO 760*F IN 6 HOURS.

s.

.- CURVE 2 - *ME01UM HEATUP RATE *:

(100% POWER + 70 DAYS OR 35% POWER + 9 OAYS).

HEATUP FROM 300*F TO 760*F IN 22 HOURS.

CURVE 3 -

• SLOW HEATUP. RATE *:

(100% POWER + 460 DAYS OR 35% POWER + 104 OAYS).

HEATUP FROM 300*F TO 760*F IN 5.5 DAYS.

A 2-3

T utto in1 b

OAK RIDGE NATIONAL LABORATORY eost orricE som v OPERATED ey taantsN edAl'NETTA ENERGv Sv$tEMS PeC October 16, 1986 Mr. Richard E. Ireland U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011

Subject:

Submittal of TER on FSV LCO 4.1.9 (FIN A 9351)

The final version of our Technical Evaluation Report (TER) on the Fort St. Vrain LCO 4.1.9 is enclosed for your review and approval. Please call me if you have any questions. s Yours truly, S. J. Ball, Manager HTGR Safety Studies for

. . . NRC SJB:djs Enclosure cc: .K. L. Heitner - NRC A. P. Malinauskas D. L. Moses

.