ML20236U696
| ML20236U696 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/23/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236U506 | List: |
| References | |
| NUDOCS 8712030300 | |
| Download: ML20236U696 (5) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLF.AR REACTOR R[GULATION SUPPORTING AMENDMENT NO. 57 TO FACILITY OPERATING LICENSE NO. DPR-34 PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION t
DOCKET NO. 50-267 I
- 1. 0 INTRODUCTION i
By letter dated June 25, 1987, Public Service Company of Colorado (licensee)
]
requested an amendment to its Technical Specifications to ensure sufficient j
helium coolant flow to prevent overheating of fuel while the reactor is in the low power or shutdown modes.
In the Fort St. Vrain reactor, four circulators force coolant downward from top to bottom through the reactor.
Thus, the hotter helium is at the bottom of the reactor.
Since hotter helium has a lower density, the direc-tion of natural circulation flow through the reactor is the opposite of this; i.e. from bottom to top.
Thus, in Fort St. Vrain, the natural cir-culatica flow opposes the helium circulator flow.
At full power the helium circulator flow is many times greater than the natural circulation flow.
However, at low power levels, core flows are low and the natural circulation flows may lead to flow stagnation and reverse flow in some channels.
Should this occur, heat would not be removed from the fuel and the core temperature would increase.
The thermal pneumatics in the Fort St. Vrain reactor are further compli-cated by the use of controllable crifice valves at the inlets of each of the 37 fuel channels or regions.
Since more power is generated in some fuel regions than others, these valves are used to orifice the helium coolant flow to the lower power regions.
The objective at higher power levels is to make the temperature of the helium coming out of the low power regions the same as that coming out of the higher power regions.
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However, at very low power levels, it is difficult to measure the tempera-tures accurately.
Hence, the reactor is operated with all orifice valves fully open at very low power levels.
For low power and shutdown cooling, these orifice control valves need to have different settings.
There are two reasons for this.
One is that at low power and low helium flow the same orifice valve settings will not provide equal outlet helium temperatures as they do at high 8712030300 871123 PDR ADOCK 05000267 P
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. helium flow.
The second is that the distribution of decay heat power is not the same as the distribution of fission power.
Thus, for the low power and shutdown modes all of the 37 orifice control valves are moved to the open position.
Then, at some time during a rise in power, they have to be reset to equalize the 37 region outlet tempera-I tures.
Each of these valves is moved by a stepping motor at a very slow rate; so, resetting these valves, without overheating any of the fuel regions, is an arduous, time-consuming task.
The objective of these proposed Technical Specifications (LCO 4.1.9) is to ensure that, with these rather complicated thermal-hydraulics, adequate core cooling is maintained while the reactor is in the low power or shutdown modes or in transition between the low and high
. power modes.
2.0 EVALUATION Technical Specification LCO 4.1.9 contains the requirements for the helium circulator flow when th~e reactor is producing less than 25 -
percent of its full power and when the " CALCULATED BULK CORE TEMPERA-TURE" (CBCT) is greater lhan 760 degrees F.
Since this is the time when the orifice valves have to have different settings from those at higher power, LCO 4.1.9 specifies the required helium flow both for the orifice valves adjusted for equal region outlet temperatures and for equal region coolant flows.
The CBCT is defined in proposed Technical Specification 4.0.4.
The temperature ~760 F was chosen as the limiting value for helium circulator flow.
Thus when the CBCT is above 760 F helium circulator flow is required.
When the CBCT is below 760 F all circulators can be turned off.
4 The staff had the Oak Ridge National Laborato'ry (ORNL) review the Public Service Company of Colorado's (PSC) proposed Technical Specification changes.
Oak Ridge's Technical Evaluation Report (TER) on these proposed Technical Specifications is contained in Appendix I to this safety evaluation.
Except as noted below, the staff agrees with and adopts the conclusions in the TER.
i As described in Appendix I, Oak Ridge also modified and used its ORECA-FSV I
computer program to independently calculate the fuel temperatures attained I
during startups and shutdowns of the Fort St. Vrain reactor.
Oak Ridge made numerous comparisons between their calculations and calculations the licensee made with the G. A. Technologies (GAT) RECA computer program.
l RECA is a steady-state flow program, which does not account for transient i
flow dynamics.
Themodifiedprogramdoesaccountfortransientdynamics.
Thus, the basis for the licensee s calculations is quite different from that for the ORNL calculations.
However, ORNL concluded that the limiting conditions predicted by GAT are at least as conservative as conditions predicted by ORNL.
Thus, ORNL concluded that the helium circulator flow that is required by these Technical Specifications will in fact prevent core overheating.
' The recommendations, comments and concerns ORNL had on other aspects of these proposed Technical Specifications are listed on pages 8 through 10 of Appendix 1.
There are 8 of these with items 2, 5, 6 and 7 classified as having been resolved satisfactorily.
Items 4 and 8 were discussed by ORNL and the staff after the ORNL TER was issued. Item 4 is on the relationship between the limits on helium flow at high powers (LCO 4.1.7) to those at low power (LC0 4.1.9).
In this regard PSC has a sentence in the proposed BASIS for LC0 4.1.9 (page 4.1-23 q
of Appendix I) which states:
i
... fuel integrity is ensured for power levels from 0 to 100%
by limiting the INDIVIDUAL REFUELING REGION OUTLET TEMPERATURES to falues given in Specification 4.1.7."
The staff has concluded that this statement, in conjunction with the operating procedures which require conformance to LCO 4.1.7, are sufficient for safe operation.
The staff did not agree with ORNL that additional cross references to LCO 4.1.7 were desirable.
It is the staff's general policy not to include excessive cross references in plant Technical Speci-fications.
However, the licensee did include a cross reference to LC0 4.1.7 in the Bases section.
In further telephone discussions with ORNL on September 10, 1987, ORNL agreed this was a minor concern.
The staff finds that the licensee has provided an acceptable resolution of this issue.
Item 8 is on specific experimental verification for the analysis codes.
The staff believes that there are sufficient conservatism in LCOs 4.1.7 and 4.1.9 to assure that adequate core cooling would be maintaihed.
Addi-tionally, the code has been compared to plant data and has shown good agreement.
In subsequent discussions by telephone on September 10, 1987.
i ORNL agreed that the Technical Specification is based on adequate data.
Thus, the staff concludes that sufficient data exists for the basis of this Technical Specification to assure safe reactor operation.
ORNL classified the remaining two iteras,1 and 3, as needing follow-up work.
Item 1 is on the definitions of core flow and core power that.
are used in the LCOs to determine the operating limits.
PSC proposes to primarily use its Power-o-Flow Measurement System.
This is des-cribed in Section 7.3.11 of the FSAR.
While this gives accurate power to flow values at high powers, ORNL states that it does not at low powers, especially when some of the helium circulators are shut down.
In the proposed BASIS for LCO 4.1.9, it is stated for the measurements used for the circulator flow that:
"The uncertainties associated with control room indication of these parameters were accounted for in the analysis."
4 And it goes on to say:
"Other flow determination methods are acceptable provided the associated uncertainties are accounted for and the calculated circulator flow is adjusted accordingly."
The staff recommends that PSC examine ORNL's suggestion of using circulator speeds in conjunction with the circulator performance maps as one of the "other methods," and that the operating procedures should use it as a check on low values of he'lium flow.
However, the staff concludes that the conservative nature of the analysis provides sufficient margin to insure reactor safety at low power even with measurement uncertainty.
The final followup item, item 3, in Appendix I is on " Mechanized Calculations" of optimum orifice valve positions.
Although, this is not a safety issue per se, the staff recommends the use of calculated orifice positions, rather than trial and error settings, when going to another operating condition.
However, the staff finds the current use of trial and error setting continues to be acceptable.
All of the above issues pertain to times when the CBCT is above 760 F.
The 760 F limit, along with the method in proposed LCO 4.0.4 for calculating the CBCT and the time it will be below 760 F, were also reviewed by ORNL.
Its Technical Evaluation Report on these is in Appendix I as an ADDENDUM to the main report.~ As stated there ORNL found that:
"The 760 F upper limit for the bulk core temperature is a con-servative limit.
No damage of any type would be expected at this temperature."
On this basis the staff finds the 760 F CBCT limit acceptable.
In the method for calculating the CBCT, ORNL found that a few clarifica-tions are needed to ensure that the method will always give conserva-tive answers.
These clarifications are:
1.
It should be made clear to the operators that the " power histury" used in.the decay heat calculations is " thermal" not " electrical" power.
2.
The procedure should ensure that the " power history" used is long enough to provide all of the significant afterheat.
3.
The procedure should specify how the initial value of bulk core tempera 1ure is conservatively measured.
The staff finds that when these clarifications are made in the operating procedures this method is acceptable.
The staff has allowed a 30 day period to implement these minor changes.
1
3.0 CONCLUSION
The staff concludvs that when the proposed Technical Specifications LCO 4.1.9 is complied with, the Fort St. Vrain reactor will be adequately cooled.
These proposed Technical Specifications are therefore accept-able.
4.0 ENVIRONMENTAL CONSIDERATION
The amendment involves a change in the installation or use of a facility component located within the restricted area.
The staff has determined that the amendment involves no significant increase in the amounts of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radia-tion exposure.
The Commission has previously issued a proposed find-ing that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amend @ent meets the eligibility criteria for categorical exclu-sion set forth in 10 CFR S51.22(c)(9).
Pursuant to 10 CFR 551.22(b),
no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
5.0 CONCLUSION
The staff has concludea, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
November 23, 1987 Principal Contributor:
E. Lantz, SRXB/ DEST
Attachment:
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APPENDIX I 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 l.9) s.
PREPARED BY:
S.
J.
Ball Oak Ridge National Laboratory Oak Ridge. TN. 37831 October 16, 1986 NRC Lead Engineer:
R.
E.
Ireland - RIV Project:
ORNL Assistance in Evaluating Licensing Request-PSV LCO 4.1.9 (FIN A9351)
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NOTICE This report was prepared as an account of work sponsored by an agency of the United States Government.
Neither the United States Government nor any agency thereof, or any of their deployees, makes any warranty, expressed or implied, or assumed any legal. liability or responsibility for any third party's uset, 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 rightst I
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1 1
3 REVIEW OF PROPOSED TECHNICAL SPECIFICATION CHANGE:
CORE INLET ORIFICE VALVES / MINIMUM HELIUM FLOW AND MAXIMUM CORE l
REGION TEMPERATURE RISE (LCO 4.1.9)
INTRODUCTION 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 (HTGR) Technical Specification - Limiting Condition for Operation (LCO) 4.1.9.
The intent of LCO 4.1.9 is to ensure that during low power and i
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 interim 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 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 G AT RECA code (Refs.2-3).
In simulating typical startups, the ORECA calculation begins with a zero-power un i f orm-t emper a t ure core and follows user-input time-varying functions of total
I 4
circulator flow, th'ermal power, primary system pressure, and core inlet temperature.
Guidelines for typical startup scenarios werc 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 j
to determine how close the operating parameters approach the
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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 i
positions and the various core and refueling region bypass flow 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 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 thermocouple, which have fairly long response times -- especially at the low flows associated with startup and 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 operators to determine compliance.
Typical startup and shutdown runs were studied in some detail.
Magnetic tapes with plant data logger outputs for the November 3, 1983, startup and the January 17, 1984, shutdown were adapted for use on ORNL computers, along with PSC's " HISTORY" program for reading, deciphering, printing, and plotting the 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 (Re f. 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 j
bypass flows were well within expected ranges.
l The PSC HISTORY code was modified to do further investigations of possible problems with tech spec limitations.
l 4.
5 The LCO 4.1.9' and 4.1.7 watchdog rout ines 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 worstacase margin (maximum allowable region outlet temperature minus the worst-case measured outlet temper *ature) is calculated both for the'startup case (average core outlet (950 F) and for the conditions (>950 F) specified by Fig. 4.1.7-1.
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 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-p.ower 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 routines.
The'three major advantages of this technique over the GAT steady-state models are the 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 pover operating regions is the calculated stagnation point for intra-region flow, and since the ORNL and GAT approaches to the problem were so different, a set of benchmurk calculations were set up by Rick Kapernick of CAT.
Some apparent discrepancies had appeared, as well, when in many cases the operating limits dictated by GAT analyses were more restrictive than ORNL's.
It was decided that for all benchmark runs, there would be two worst-cose refueling
s 1,
6 regions.
'The hotter region had a region peaking factor (RPF) of 3.0 and's 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, 14 and 54,
'with a fixed helium inlet density corresponding to 250 F and 367 psia at 14 power, and 350 F and 419 psia at 5* Power.
Total reactor flows were varied between 1 and 10* for the-14 power.
case.,and 54 and 134 for the 54 power case.
The ORECA code, which noreally 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
' o ri fices 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 1* power cases and 134 for the 54 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, flow stagnation with flows as low as St.
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
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equalize outle'. 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 14 more flow was required j
'to prevent. stagnation without intra-region conduction'than for cases with normal conduction-included.
GAT's calculated minimum flow'to prevent. stagnation at 1*
f power was 6.9%, vs. 54 per ORECA.
The GAT analysis was done for l
e higher density gas (107.5% vs. 904 inventory in the benchmark) and neglected conduction (a 14 flow effect), so these differences could easily account for the discrepancy.
A comparison of the predicted benchmark core conditions at 10* flow also showed no 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 RPFs and tilts, i.e.
RPF = 1.80 and tilt = 1.36.
These also showed intra-region flow reversals when the' total flow was 3
reduced to 44.
However, if the outlet temperature equalization
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scheme is used to reposition the orifices, the reversals didn't 1
occur until the flow was reduced to about 24.
For the 5* power benchmark cases, excellent agreement
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1 between'ORECA and GAT calculat' ions'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 54 (74 neglecting conduction), although very high core and gas' temperatures were calculated.
The GAT analysis (at 107* vs.
90* inventory) gave 10.34 as the minimum flow to avoid j
stagnation.
Further investigation, however, showed that ORECA j
probably would have eventually calculated flow stagnation at flows higher than 5-7*.
The ORECA calculations'had been
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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.
For cases in which our "more realistic" worst-case RPFs ad tilts were used. ORECA runs indicated that no stagnation occurred at 5* flow and that orifice manipulation was able to give reasonable core temperatures.
Further variations on the tienchmark runs were studies to investigate potential problems with the use of equal outlet temperature orifice settings at the low power / flow conditions.
At 14 power, for the minimum peaking factor region (RPF 0.4 and column tilt =
1.507), which is limitang for the equal-outlet
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1 i
8 temperature mode, stagnation occurred at 44 flow.
However, by i
simply limiting the minimum orifice setting to 30% 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 i
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 I
surveillance requirements.
A list of meetings attended by ORNL 1
is in Attachment 1.
The following summary is a chronology of the more pertinent issues and questions raised, where the eventual 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 circulator performance maps.
We also recommend incorporating calculated afterheat in the power estimate when approaching or following shutdown.
2.
"C'ap" in coverage of operating conditions (R) -
In the figure limiting core outlet terapearture 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 660 F.
From the data we observed, all the region outlet temperatures are typically low enough so that there would not be a real problem with overheating the core.
o 9
3.
" Mechanized Calculations" of opt imum orifice positions (F)
We recommended that PSC should use a simple program to calculate optimum orifice positions for each desired operating i
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 relatiocchip between LCos 4.1.7 and 4.1.9 has not been universally well-understcod.
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 ere 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 auximum fuel temperatures.
Simply -
assuring no stagr.ation (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 een be related to fuel temperatu're.
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 temperatu.re rises are not shown as_being limited for equal
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region flow cases where the system pressure is')5O psia.
We
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Selieve it should be noted in the table, or at least in a s
footnote, that these cases are to be limited by restrictions in LCO 4.1.7.
5.
Surveillance vs. Corrective Action time requirements (R) -
The time limits are required corrective action and the surveillance time intervals are not consistent.
For example, t
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 IS-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.
S R 4.178-f o rmFr ly- ~
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.
6.
Questions on procedures for determining " bulk core temperature" limit of 760F (R) -
These questions were addressed in detail and reported in our A9351 monthly report for January 1986 (See attachment 2).
We i
10 recommended that reference be made in the Basis to the procedures used to calculate bulk core temperature.
7 Equal-Flow ori fice range specification (R)
The'LCO did not specify a limiting range for orifice positions corresponding to the equ'al flow setting mode.
We recommended 10-20% open.
The opening size is crucial to'the flow stagna, tion 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 variationa 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 wer.e recommended proposed relatively small flow perturbations'(within current tech spec limits).
These tests would provide data on V
whole-region, not intra-region, flow redistributions.
SUMMARY
AND CONCLUSIONS 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 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.
4)
Point out problems and suggest alternatives.
As in the case cf previous versions of LCO 4.1.9, the limits imposed by the final LCO (P-86451) range from equivalent to conservative as compared to those derived by ORNL analyses.
i Hence, the major concern of the Tech Spec, that of providing limits that will in fact prevent core overheating, has been
______________a
11 addressed and confirmed satisfactorily.
Remaining disagreements and concerns were primarily.with details of clarity and style.
REFERENCES 1.
S.
J. Ball. ORECA-1:
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. Cor,t rol and Testing Symposium, Gatlinburg, Tennessee (March 1980).
l 3.
S.
J.
Ball, et al.,
High-Temperature Gas-Cooled Reactor Safety Studies for the Division of Reactor Safety R e s c a t e_ h.
Quarterly Progress Report. January 1-March 31. 1978 NUREG/CR-0179, ORNL/NUREG/TM-221 (July 1978).
4.
Letter from D.
W.
Warembourg (PSC) to S.
J.
Ball, " History Tape Transmittal for LCO 4.1.9' Evaluations", March 2, 1984 (P-84073).
j 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).
ATTACHMENT 1 Meetings on LCO 4.1.9 Proposed Changes attended by ORNL 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.
l l
l I
v ATTACHMENT 2 0
t.tJLt tmtlHL W *Sl rion t hl v f<epor i f or Januar, 1*rt <.
P A Gt: 1 1
l REVIEW 0F LCO 4.1.9 BASIS RELA 11NG lu /69 F AVERAGE j
CURE TEMPERATURE LIMIT The basis in the draft LCO 4.1.9 ( P -8 5 4 4 '.' ) pertaining to the 76u F maximum core temperature lamat is as fo!!ows:
l
" The c al cul ated bull. core temperature is the calculated average temperature of the core. ancluding graphite and fuel but not the reflector. that occurs following a loss of ali forced circulation of pri mary cool ant flow.
The calculation assumes that all decay heat is i
retained an the core wi th no heat transfer to the reflector, PCRV anternals or primary coolant.
If the decay heat is sufficiently low, with all primary coolant flow terminated, the calculated bulk core
-temperature will not exceed 760 degrees F.
this specification is 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 an thi s 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 11mit was apparently derived from the design value of core anlet temperature, wh1ch is
'j 760 F; hence there is no safety concern af a!! of the components in the reactor vessel are nominally limited to tha's temperature.
Several' questions and concerns do arise, however, regarding the amplementattom of this limit to specific reactor operation scenarios
- 1)
How is the afterheat calculated?
2)
How hot do critical PCRV-i n t r>rna l (metal) components get with a core average temperature of 760 F7 31 How conservative is the assumption that there as no heat loss to the surroundings during a no-ffow heatup?
l ITEM 1:
The means of determining afterheat is not specified in the LCO, and at should be.
Since there are no sensors in the core which can ef f ecti vel y 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 perturbations sin, _
'C"- t h e p er i od before shutdown.
(We have developed such an algorithm that could be implemented on the plant computer or a programmable calculator.)
ITEM 2:
A variety of heatup conditions were simulated using the severe-accident version of the ORECA-FSV code, ranging f rom rel,at i vel y rapid heatups (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 an PCRV temperatures were l ar ger in the faster heatupst however, at the time when the average core temperature reached 760 F.
ma:: 1 mum fuel temperatures were well below normal operating temperatures and PCRV metallic component A 1 l
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{
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I ENCLOSURE - A9351 Monthly Report for January 1986 PAGE 2 l
temperatures were well below 760 F.
Hence we would conclude that the 760 F core average temperature limit is sufficiently conservative.
ITEM 3:
Also of interest was the ext.ent of the conservatism in the i
assumption that there is no heat loss from the core during the heatup.
Again, by use of the ORECA-FSV code, it was shown that this j
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 f or 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 af terheat is 50 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 damages however, the sneans for computing the adiabatic l
temperature rises should be specified in the LCO or its basis.
k 9
O i
0 t
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A 2-2
i i
4 l
p Figure 1 - Comparisons of Actual vs. Adiabatic Core.Heatup Rates l'
l l
ORNL.DWC 86C.6189 ETO l
l
,I i
?
f 100 e-a j Q Ne
=
u $
< o 80 N
ai 7
60
=
40 e
e 20 -
I I
I I
O 300 400-500 600 700 800 AVERAGE BULK CORE TEMPERATURE (*F) l NOTES:
CURVE 1
" FAST HEATUP RATE **
(100% POWER + 6 OAYS OR 35% POWER + 3 HOURS).
HEATUP FROM 300*F TO 760*F IN 6 HOURS.
CURVE 2 - *MEOlVM HEATUP RATE *-
(100% POWER + 70 DAYS OR 35% POWER + 9 O AY51.
HEATUP FROM 300*F TO 760*F IN 22 HOURS.
CURVE 3 -
- SLOW HE ATUP, RATE *-
(100% POWER + 400 DAYS OR 35% POWER + 104 OAYS).
HEATUP FROM 300*F TO 760*F IN 5.5 OAYS, 1
e A P2-3
-- ---- -- - o