ML20141B745
| ML20141B745 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/28/1986 |
| From: | Heitner K Office of Nuclear Reactor Regulation |
| To: | Lynch O Office of Nuclear Reactor Regulation |
| References | |
| TAC-52634, NUDOCS 8604070096 | |
| Download: ML20141B745 (17) | |
Text
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~g UNITED STATES
! g NUCLEAR REGULATORY COMMISSION
, 5 E WASHINGTON, D. C. 20555
.....* March 28, 1986 Docket No. 50-267 MEMORANDUM FOR: Oliver D. T. Lynch, Jr., Section Leader Standardization and Special Projects Directorate Division of PWR Licensing-B FROM: Kenneth L. Heitner, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B
SUBJECT:
SUMMARY
OF MEETING WITH PUBLIC SERVICE COMPANY OF COLORADO (PSC) TO DISCUSS FORT ST. VRAIN (FSV) TECHNICAL SPECIFICATIONS (TS)-LC04.1.9OhMARCH 13, 1986
References:
- 1. Letter to E. J. Butcher, NRC, from H. L. Brey, PSC, dated November 22, 1985 on Revised Draft of LC0 4.1.9, Core Region Temperature Rise
- 2. Letter to E. H. Johnson, NRC, from S. J. Ball, ORNL, dated January 10, 1986, on Monthly Report "0RNL Assistance in Evaluating Licensing Request - FSV LC0 4.1.9"
, 3. Memo to H. Berkow, NRC, dated December 20, 1986, on FSV TS Upgrade Changes This memorandum sumarizes a meeting held with PSC at the Region IV NRC Office in Arlington, Texas on March 13, 1986. Discussions were held with PSC
- on their proposed draft for LC0 4.1.9 (Reference 1). The staff's contractor, Oak Ridge National Laboratory (ORNL), had reviewed this submittal and provided coments in Reference 2. Additional staff coments in Reference 3 were also discussed with PSC.
l The attendees at this meeting are given in Enclosure 1.
A sumary of the resolution of the coments in References 2 and 3 follows.
Reference 1 Coment 1 - PSC stated that plant operators monitor compliance with LC0 4.1.9 more frequently as they manipulate the reactor controls during power changes. The staff observed that the approach for this LC0 should be -
consistent with other LCOs that affect power', flow, and region temperature differential. Surveillance requirements at a higher frequency during power changes should be consistent with the time requirements for action in the LCO.
B604070096 860328 PDR ADOCK 05000267 P PDR
Coment 2 - PSC stated that the final curves would include the 26 percent power level. These curves are shown in Enclosure 2, a redraft of the proposed Technical Specification.
Comment 3 - PSC agreed that the proposed action statements for the reactor shutdown were unclear. The proposed LC0 would be revised with separate sections for the reactor in operation and the reactor in shutdown.
Comment 4 - PSC will explain in the basis why the reactor is operated in the equal flow mode below about 4 percent power. Potentially, Figure 4.1.9-3 could be labeled to show the range of equal flow operation.
The need to cross reference other LCOs will also be considered.
Coment 5 - No further discussion needed.
Coment 6 - PSC stated that the accuracy of the measured differential pressure used to calculate flow had been improved. Thus, accuracy of flow data at low flow rates was not a problem. The staff was concerned that if flow remained zero because of circulator problems, then no action would be required since thermal power would be zero. PSC agreed that a loss of flow with sufficient after heat would require consideration of LC0 4.2.18. The action statement would be modified to reflect this situation.
Comment 7 - PSC proposed that the equal orifice setting cover a range of 8 to 20 percent open. This would be clarified in the basis for the LCO.
All comments on the basis for the LC0 were accepted by PSC.
Reference 2 Coment 1 - PSC agreed with the suggested changes.
Comment 2 - PSC stated that they had examined the instrumentation which supported LC0 4.1.9 and most was covered by current surveillance requirements. They still needed to compare the instrumentation against the l
proposed surveillance requirements of the TS Upgrade Program.
! PSC agreed to consider including all supporting instrumentation in the upgraded TS.
Comments 3&4 - PSC supplied draft definitions of these terms in Enclosure 3.
Coments 587 - PSC agreed to relabel the curves to clarify that they apply for a range of helium densities. .
Coment 6 - PSC agreed to utilize an expanded scale in the range of 0 to 5 percent power, as an aid to reading the figures. The limits to the curves in Figure 4.1.9-3 would be explained.
i Coments 8&9 - PSC accepted these coments.
March 28, 1986 In addition, PSC noted that additional data on the basis for the 760*F bulk core temperature concept were submitted in their letter, P-86169. They noted that the supporting calculations for this evaluation would be provided by the reactor engineers, not the plant operators.
The staff and PSC agreed that LC0 4.1.9 should be submitted and initially approved in the current TS format. PSC agreed to submit a final draft by April 30, 1986 and an amendment request by June 15,-1986.
original signed by Kenneth L. Heitner, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B
Enclosures:
As stated cc w/ enclosures:
See next page DISTRIBUTION:
7DecRet;FHe.3
'NRC PDR'~
Local PDR SSPD Reading PNoonan KHeitner Olynch HBerkow OELD EJordan BGrimes ACRS (10)
NRC Participants cc: Licensee and Plant Service List
. 'N d
In addition, PSC noted that additional data on the basis for the 760*F bulk core temperature concept were submitted in their letter, P-86169. They noted that the supporting calculations for this evaluation would be provided by the reactor engineers, not the plant operators.
The staff and PSC agreed that LC0 4.1.9 should be submitted and initially approved in the current TS format. PSC agreed to submit a final draft by April 30, 1986 and an amendment request by June 15, 1986.
=~ }- ,
Kenneth L. Heitner, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B
Enclosures:
As stated cc w/ enclosures:
See next page 9
m
. Mr. R. F. Walker .
Public Service Company of Colorado Fort St. Vrain cc:
Mr. D. W. Warembourg, Manager Albert J. Hazle, Director Nuclear Engineering Division Radiation Control Division Public Service Company Department of Health of Colorado 4210 East lith Avenue P. O. Box 840 Denver, Colorado 80220 Denver, Colorado 80201 Mr. David Alberstein, 14/159A Mr. J. W. Gahm, Manager GA Technologies, Inc. Nuclear Production Division Post Office Box 85608 Public Service Company of Colorado San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. H. L. Brey, Manager Nuclear Licensing and Fuel Division Mr. L. W. Singleton, Manager Public Service Company of Colorado Quality Assurance Division P. O. Box 840 Fort St. Vrain Nuclear Station Denver, Colorado 80201 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. 0. Box 640 Platteville, Colorado 80651 Kelley, Stansfield &'0'Donnell Public Service Company Building Room 900 550 15th Street Denver, Colorado 80202 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 l
Arlington, Texas 76011 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 l
Regional Representative Radiation Programs Environmental Protection Agency 1800 Lincoln Street
< Denver, Colorado 80203
~
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Enclosure 1 Attendees Meeting PSC - NRC on March 13, 1986 in Arlington, Texas Name Organization Ken Heitner NRC/NRR/PBSS M. H. Holmes Nuclear Licensing, PSC Syd Ball ORNL Chuck Fuller PSC, FSV Station Manager Rick Kapernick GA Technologies Jack Levin Grove Engineering D. Alberstein GA Technologies R. E. Ireland NRC-RIV D. R. Hunter NRC-RIV, RSB J. P. Jaudon NRC-RIV J. E. Gagliardo NRC-RIV o
4
'. i- Enclosure 2 DRAFT - MARCH 13,1986 REACTOR CORE AND REACTIVITY CONTROL LCO 4.1.9 CORE INLET ORIFICE VALVES / MINIMUM HELIUM FLOW and MAXIMUM CORE REGION TEMPERATURE RISE LIMITING CONDITION FOR OPERATION The total reactor helium coolant flow or the helium coolant temperature rise for all core regions shall be maintained within the limits given in Table 4.1.9-1.
APPLICABILITY: Power levels below 25%, including shutdown with decay heat.*
ACTION: a. Wi' i any of the above limits exceeded, either:
- 1. Increase the region helium coolant flow or correct the out-of-limit condition within 15 minutes, or
- 2. Be in at least REACTOR SHUTDOWN within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with the inlet orifice valves adjusted for equal region
. coolant flows within the following 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
SURVEILLANCE REQUIREMENTS The total reactor coolant flow or the helium coolant temperature rise through all core regions shall be determined to be within the above limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
l l
i
- With the calculated bulk core temperature greater than 760 l
degrees F.
L
D R A FT - M A R C H 13,19 8 6 Page 2 of 8 Table 4.1.9-1 REGION ORIFICE REACIOR PRESSURE LIMITING CONDITION EDR prr! TION HELI (M DENSITY OPERATION All regions set Greater than 50 he total core helium coolant for equal region psia with flow shall be greater than or coolant flow *** helium density
- equal to the minimum EXCEPT greater than allowable value shown on Up to 5 regions 60%. Figure 4.1.9-1.
may have their orifices further open.
As above. Greater than 50 n e total core helium coolant psia with helium flow shall be greater than or density
- less equal to the minimum than or equal to allowable value shown on 60%. Figure 4.1.9-2 All regions set Isss than or De helium coolant tenperature for equal region equal to 50 psia. rise ** through any core region coolant flow.*** shall not exceed 600 degrees F.
Orifice valves Greater than he helium coolant tenperature at any position 50 psia rise ** through any core region (Mjusting for shall not exceed the limit equal region shown in Figure 4.1.9-3.
outlet tenpera-ture).
Orifice valves Isss than or De helium coolant tenperature at any position equal to 50 psia rise ** through any core region (Mjusting for shall not exceed 350 degrees equal region F.
outlet tenpera-ture).
Percent helium density equals:
175.12 x Reactor Pressure (psia)
(Circulator inlet tenperature (degrees F) plus 460) . .
Helium coolant tenperature rise equal INDIVIDUAL REFUELING REGION OLTITEP 11MPERATURE minus CORE AVERAGE INIfr TFMPERATURE
- Equal region coolant flow with orifice valves set between 8% and 20% open.
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I DRAFT - M A RC H 13,1986 BASIS FOR SPECIFICATION LCO 4.1.9 The minimum reactor helium coolant flow or the maximum core region helium coolant temperature rise as a function of calculated reactor 6
thermal power (including power for decay hat) have been specified to prevent very low helium coolant flow rates through any coolant channel. Very low helium coolant flow rates may result in laminar flow conditions with resultant high friction factors, low heat transfer film coefficients, and potential for possible local helium flow stagnation or reverse flow, which could result in excessive fuel temperatures.
This Specification addresses minimum flow requirements for all coolant channels. Since low coolant flows exist at lower reactor powers, its applicability is limited to less than approximately 25%RkTEDTHERMALPOWER. Since thermal power is continuously generated by decay heat even after the reactor is shutdown, the
~
flow requirements are also applicable in the REACTOR SHUTDOWN mode.
This specification is not applicable when the CALCULATED BULK CORE TEMPERATURE is less than 760 degrees P. If the active core is below this temperature, which corresponds to the design maximum core inlet temperature, then the design core inlet temperature could not be exceeded and there is no
~ '
possibility of damage to fuel or PCRV internal _ components regardless of the amount, or even total absence of, primary coolant helium flow. The applicability of this Specification is Iso limited to the range of power level indicated in
~.
9 DRAPT - M A R C 11 13,1986 Figures 4.1.9-1, 4.1.9-2 and 4.1.9-3. Above the power levels for which limits are shown in the Figures, the Reactor Core Safety Limit, Specification 3.1, governs. In addition to this Specification, fuel integrity is ensured for power levels from 0 to 100% by limiting the INDIVIDUAL REFUELING REGION OUTLET TEMPERATURES to values given in Specification 4.1.7.
The CALCULATED BULK CORE TEMPERATURE is the predicted, time dependent, 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 usca several conservative assumptions including: 1) that the core heatup rate remains constant during the heatup at the initial value which is calculated based on the decay heat from most restrictive recent power history or at an empirically determined, 2) that the composite specific heat, volume, and density of the core remain constant during the heatup at the initial values, and 3) that all decay heat generated af ter the assumed loss of forced circulation is retained in the active core with no heat transfer to the reflector, PCRV intutoals or primary coolant.
The limits have been developed based upon a number of conservative assumptions. For the limits in Figures 4.1.9-1 and 4.1.9-3, it was assumed that the primary system was pressurized to full inventory (107.5 percent of design helium density was used in t'he analysis).
At lower densities, higher region temperature rises and lower core coolant flow are acceptable. Since startup operations can proceed
DRAFT -
M A .R C H 13,1986 4
with lower helium densities, af ter the reactor has been pressurized to greater -than 100 psia, which corresponds to about 30 percent helium inventory at 200 degrees F, flow requirements were '
calculated for 60% helium density and are given in Figure 4.1.9-2.
Percent helium density equals:
8 175.12 x Reactor Pressure (psia)
(Circulator inlet temperature (degrees F) plus 460)
The core inlet helium temperature used in the analysis cover the range of 100-400 degrees F between 0 and 5% RATED THERMAL POWER and 100-700 degrees F between 5 and 25% RATED THERMAL POWER. These are reasonable assumptions for low power operation.
In the analysis to determine the limits, the ef fects of heat conduction between columris in a region, or between regions, were conser'vatively neglected. Envelope values of RPF/ Intra Region Petking Factors (3.0/1.25 and 1.6/1.61) were used to anticipate
- worst case conditions considering all future fuel cycles.
Consistently conservative nominal values and uncertainties were used for bypass flows and measured parameters throughout the cnalysis. For the condition with orifice valves at any position, the allowable region delta T is based upon a region peaking factor cqual to 0.4. For regions with high:. .ser dennities, higher region delta T's are acceptable. .
The analysis also accounts for maximum uncertainty in the instrumentation used to measure the necessary input parameters. Thus, the Specification limitations can be
o
- se D R A FT - MARCH 13,1986 cpplied directly to measured values without further consideration of instrument calibration accuracy and no instrumentation surveillance beyond routine calibration is required.
Besides the minimen flow requirement curves with the orifices set for equal region flows in rigures 4.1.9-1 and 4.1.9-2, flow r6quirements are provided with a number of orifice valves positioned further open. These curves allow for a minimum number of orifices stuck open as well as assisting in the transition between equal region flows and equal region outlet temperatures.
Cy monitoring the total reactor coolant flow when the orifices are cdjusted for equal region coolant flows, minimum flow through each region at the appropriate power can be assured. When the orifice valves are adjusted to dif ferent positions, mir.imum coolant flows can be' assured for each region by monitoring the helium coolant temperature rise in that region.
For depressurized operations, limits area also specified to prevent very low helium coolant flow rates through any coolant channel.
These limits have been establisted based upon a 50 psia reactor pressure.
To ensure that flow stagnation in a fuel column or region does not persist. an action time of only 15 minutes is allowed to correct the out of limit condition. ,
The requirement to be in PEACTOR SilDTDOWN within I hour with the crifices set for equal flown in an additional 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is realistic because it takes from 4-6 hours to set the orifice valves from
. DRAFT -
M A RCH 13,1986 cqual temperatures to equal region flows. This is considered cceeptable since there is sufficient primary coolant flow from the circulators which are driven by steam generated from residual heat in the system following REACTOR SHUTDOWN.
In performance of the surveillance, the total reactor helium coolant flow is determined by calculation based on measured circulator inlet nozzle low range delta P, temperature and pressure. This is consistent with the method to determine the rcquired flow in the analysis. Procedures require that the flow rcte be mcnitored whenever the power level is being changed to cnsure that the requirements of this Specification are satisfied, but the surveillance is required once per shift (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) and is consistent with other Specifications.
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- Enclosure 3
- DEFINITIONS CAT.CULATED BULA CORE TEMPERATURE r
?
The CALCULATED BULK CORE TEMPERATUP.E shall be the calculated t average temperature of the core, including graphite and fuel, but not the reflector, assuming a loss of all forced circulation f7 of primary coelant flow.
I CORE AVERAGE ThMPEII ATURE I
- a. buring SHUTDOWN and REFUELING, CORE AVERAGE TEMPEkATURE
- - be the arithmetic average of CORE AVERAGE INLET !
shall
, TEMP 2hATURE and the CORE AVERAGE OUTLET TEMPERATORE.
i
% pu'cing STARTUP, LOW POWER, and POWER, COF.E AVEkAGE l TEMPERATURE shall De therre0 dynamically calculated basec! on CORE AVERAGE INL>;T and CORE AVERAGE O'JTLET TEMPERATURES, PRIMAkY COOLANT FLOW, and renctor power.
i l .
- QRE AVERAGE INLET TF.MPERATURE
,' The CORE AVERAGE INLET TEMPERATURE shall be the arithmetic ave rage of the operating circulator inlet temperatures, adjusted for circulator power input, steam generator regenerative heat loads, and PCRV liner cooling system heat losses.
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