ML12355A019
| ML12355A019 | |
| Person / Time | |
|---|---|
| Site: | University of Missouri-Columbia |
| Issue date: | 09/30/2010 |
| From: | Rhonda Butler Univ of Missouri - Columbia |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Wertz, Geoffrey 301-415-0893 | |
| References | |
| TAC ME1580 | |
| Download: ML12355A019 (24) | |
Text
UNIVERSITY OF MISSOURI - COLUMBIA RESEARCH REACTOR LICENSE No. R-103 DOCKET No. 50-186 RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION RELATING TO LICENSE RENEWAL REDACTED VERSION*
SECURITY-RELATED INFORMATION REMOVED
- REDACTED TEXT AND FIGURES ARE BLACKED OUT OR DENOTED BY BRACKETS
UNIVERSITY of MISSOURI RESEARCH REACTOR CENTER September 30,2010 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station PI-37 Washington, DC 20555-0001
Reference:
Docket 50-186 University of Missouri-Columbia Research Reactor Amended Facility License R-I03 Enclosed you will find the University of Missouri-Columbia Research Reactor's responses to the U.S. Nuclear Regulatory Commission's (NRC) request for additional information, dated June I,
- 2010, regarding our renewal request for Amended Facility Operating License R-I 03, which was submitted to the NRC on August 31, 2006, as supplemented.
If you have any questions, please contact Leslie P. Foyto, the facility Reactor Manager, at (573) 882-5276 or foytol@missouri.edu.
Sincerely,
~~
Ralph A. Butler, P.E.
Director RAB/djr Enclosures Aoao L\\ (l((
1513 Research Park Drive Columbia, MO 65211 Phone: 573-882-4211 Fax: 573-882-6360 Web: http://web.missouri.edul-murrwww Fighting Cancer with Tomorrow's Technology
UNIVERSITY of MISSOURI RESEARCH REACTOR CENTER September 30, 2010 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station Pl-37 Washington, DC 20555-0001
REFERENCE:
Docket 50-186
SUBJECT:
University of Missouri - Columbia Research Reactor Amended Facility License R-103 Written communication as specified by 10 CFR 50.4(b)(1) regarding responses to the "University of Missouri at Columbia - Request for Additional Information Re: License Renewal, Safety Analysis Report, 45-Day Response Questions (TAC No. MD3034),"
dated June 1, 2010 On August 31, 2006, the University of Missouri-Columbia Research Reactor (MURR) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) to renew Amended Facility Operating License R-103.
On June 1, 2010, the NRC requested additional information and clarification regarding the renewal request in the form of one hundred and sixty-seven (167) questions. By letter dated July 16, 2010, MURR responded to forty-seven (47) of those questions.
On July 14, 2010, via electronic mail (email), MURR requested additional time to respond to the remaining one hundred and twenty (120) questions. By letter dated August 4,2010, the NRC granted the request. By letter dated August 31, 2010, MURR responded to fifty-three (53) of the questions.
On September 29, 2010, via email, MURR requested additional time to respond to the remaining sixty-seven (67) questions and this request is pending approval by the NRC. MURR's responses to sixteen (16) of the remaining questions are attached.
If there are questions regarding this response, please contact me at (573) 882-5276 or foytoICdlmissouri.edu. I declare under penalty of perjury that the foregoing is true and correct.
ENDORSEMENT:
Sincerely, Reviewed and Approved, Reactor Manager 1513 Research Park Drive Columbia, MO 65211 Phone: 573-882-4211 Fax: 573-882-6360 Web: http://web.missouri.edul-murrwww Fighting Cancer with Tomorrow's Technology
Enclosed: : Relevant portion ofMURR internal report titled, "MURR Upgrade Safety Limit Analysis." : SAR Figure 4.13, "MURR Safety Limit Curves (Pressurizer at 75 Psia)" -
with lines drawn for added clarity xc:
Reactor Advisory Committee Reactor Safety Subcommittee Dr. Robert Duncan, Vice Chancellor for Research Mr. Craig Basset, U.S. NRC Mr. Alexander Adams, U.S. NRC Page 2 of9
CHAPTER 4 4.18 Section 4.6.3, Safety Limit Analysis.
- a.
Page 4-57. Cobra3C is stated as being used to verify outdated BOLERO results. Discuss the Cobra3C analysis and results, and the verification methodology.
The MURR safety limits were verified using the more modem program COBRA, Version 3C.
This program was used to reconfirm the results and conclusions of the safety limit analysis that was performed in 1973 by the NUS Corporation and which forms the basis for the safety limit curves that are included in the SAR. The COBRA analysis confirmed that the MURR safety limits, as shown in Figures 4.12,4.13 and 4.14, are still valid and quite conservative. It should be noted that these safety limit curves were not changed or updated based on the newer analysis.
The details ofthe methodology used and results from the COBRA-3C analyses are documented in the MURR internal report titled, "MURR Upgrade Safety Limit Analysis." The relevant portion of that document is enclosed as Attachment 1.
Additionally, as previously discussed in the introduction to Question 4.13, neutronics and single channel thermal-hydraulic analysis programs such as BOLD VENTURE and COBRA-3C were used during the reactor analyses that were performed in 1986 in support of the submittal to the NRC for approval of Amendment 20 to Amended Facility License R-103. Amendment 20, which was issued on August 1, 1990, authorized MURR to use Extended Life Aluminide Fuel (ELAF) in the reactor core in combination with the current highly-enriched uranium fuel element (Ref. 4.46). This amendment allowed the MURR to use a new fuel element design that would have significantly reduced the fuel cycle cost and reduced the amount of uranium-235 needed per MWD of energy produced. Nonetheless, as stated on page 4-29, the ELAF elements have not been fabricated and never will be, and hence, will never be used in MURR cores.
4.18 Section 4.6.3, Safety Limit Analysis.
- b. Page 4-51. The NUS study results in three sets of safety limit curves (figures 4.12, 4.13, and 4.14), which effectively define the conditions at which the Departure from Nucleate Boiling Ratio (DNBR) is equal to 1.2. However, this is not the typical DNBR at normal operating conditions (or transients). Discuss the thermal margin by showing the DNBR at normal operating conditions.
That is correct. Figures 4.12, 4.13 and 4.14 are the limiting power values obtained for the worst location within the reactor core and are based on the values listed in Table 4-16. The peaking values assumed in this worst (hot) channel are provided in Table 4-14. As described in the first six paragraphs (pages 4-26 to 4-29) of Section 4.5, Nuclear Design, the MURR steady-state safety limits are based on nucleonics analyses performed prior to July 1974 using the EXTERMINATOR-II code. Starting with the fourth paragraph on page 4-29, the SAR discusses how the BOLD VENTURE-IV code system was used in the mid-1980s to perform 3D nucleonics analyses, including fuel depletion, using diffusion theory methods.
This information up to Section 4.5.1, Normal Operating Conditions, was included to show that using this more sophisticated code system confirmed that the original analyses using the EXTERMINATOR-II code provided sufficiently conservative steady-state safety limits.
The safety limit assumed Peaking Factor On Heat Flux is 4.35 (see Table 4-14), which includes Engineering Hot Channel Factors on Flux of 1.030 and 1.150, so the power related Nuclear Peaking Factor is 3.676 [4.35 including the engineering peaking factors]. This is obtained from combining the peaking factors Page 3 of9
from three separate EXTERMINATOR-II code 2D models, viz., R-Z (for radial-2.220 and axial-1.432 peaking), 8-R all fresh fuel (for azimuthal within a single fuel element-1.040) and 8-R mixed burnup (for azimuthal between fuel elements with different burnups-1.112). The BOLD VENTURE 3D model showed the highest nuclear peaking factor calculated for the MURR gram mixed burnup cores was 3.146 [3.73 including the engineering peaking factors]. So the worst case peak heat flux is assumed to be 16.6% higher than what the actual worst case peak would be.
As shown in Table 4-14, with BOLERO the enthalpy rise is based on all the same nuclear peaking factors except for Axial. This is because BOLERO conservatively calculates a safety limit using the peak heat flux point, which is slightly below the axial center of the coolant channel, combined with the enthalpy rise at the end of the channel and the pressure at the reactor core exit. Therefore, the hot stripe is considered to have a power level 2.72 times the average and a flow rate 0.81 times the average. Under these extreme conditions, a DNBR of 1.2 will be reached when the reactor is operated at the limiting power value given by the safety limit tables for each combination of pressurizer pressure, core coolant flow rate, and core inlet temperature.
For our normal operating values: greater than 75 psia, 3,800 gpm and 120 of, the safety limit on power would be 18.985 MW. For these conservatively assumed peaking factors, the DNBR would be 2.28 (= 18.98511 0 x 1.2). The DNB would occur at a power level of 22.8 MW based on 60% of the Bernath correlation value as determined in the preliminary Advanced Test Reactor (ATR) testing (Ref. 4.32), which were performed on coolant channel lengths twice as long as the MURR channels. This provides conservative values based on Reference 4.33.
4.18 Section 4.6.3, Safety Limit Analysis.
- c.
Page 4-59. Explain the impact of measurement uncertainty and instrument response time on the results of Case One.
The attached safety limit curve (Attachment 2), which is the same as Figure 4.13 (page 4-55 of the SAR), but includes lines to help better understand the three cases discussed in Section 4.6.4.2, Bases. With three of the four Limiting Safety System Setting (LSSS) variables at their respective limit, the fourth LSSS variable has to be exceeded by a significant margin before a safety limit is reached, which corresponds to DNBR of 1.2. The attached figure can also be used to explain the impact of instrument and measurement uncertainties.
If a power uncertainty of 5% is assumed for the situation discussed in Case One, the high power scram would be initiated only after reaching a power level of 13.1 MW instead of 12.5 MW.
With pressurizer pressure at 75 psia, core coolant flow rate at 3,200 gpm and core inlet temperature at 155 OF, the limiting power based on the safety limit curve is 14.89 MW.
Therefore, the high power scram would still occur at 1.79 MW less than the safety limit.
If a temperature uncertainty of 5 OF is assumed for the situation discussed in Case Two, the high temperature scram would be initiated only after reaching a temperature of 160 OF instead of the 155 OF LSSS. With core coolant flow rate at 3,200 gpm and pressurizer pressure at 75 psia, the safety limit curve gives a limiting power level of 14.53 MW. Therefore, the high power scram would still occur at 2.03 MW less than the safety limit.
If a flow uncertainty of 50 gpm is assumed for the situation discussed in Case Three, the low flow scram would be initiated only after reaching a core coolant flow rate of 3,150 gpm instead of the 3,200 gpm LSSS. With a core inlet temperature of 155 OF and pressurizer pressure at 75 psia, the Page 4 of9
limiting power level is 14.75 MW. Therefore, the high power scram would still occur at 2.25 MW less than the safety limit.
If a pressure uncertainty of 2 psi is assumed, the low pressure scram would be initiated only after reaching a pressurizer pressure of 73 psia instead of the 75 psia LSSS.
With a core inlet temperature of 155 OF and a core coolant flow rate at 3,200 gpm, the limiting power level is 14.64 MW. Therefore, the high power scram would still occur at 2.14 MW less than the safety limit.
Additionally, the power, temperature, flow and pressure uncertainties mentioned above are greater than the calibration tolerances of the instruments that monitor these variables, thus the high power scram would actually occur at a greater margin below the actual safety limit than those stated above.
Next, considering the instrument response times, the four variables that determine the safety limit curves can be divided into those that can cause a safety limit to be exceeded if they get too high -
power and temperature, and those that can cause a safety limit to be exceeded if they get too low
- flow and pressure. The flow and pressure sensors are very responsive instruments and have demonstrated their timely response in actual loss of flow and loss of pressure incidents and as indicated in the loss of flow accident analysis. The temperature can only rise due to a loss of flow or an increase in power; and due to the heat capacity of the system the power can respond faster than the temperature can rise. The reactor is protected from a fast power increase by the period scram due to its fast response time.
CHAPTER 12 12.7 Section 12.4, Reportable Events and Required Actions and TS 6.5.a., Reportable Events and Required Actions, Safety Limit Violation. Reporting events to the NRC under proposed TS 6.5 specifies notification of the NRC Project Manager for MURR rather than the 24-hour NRC Operations Center.
Propose changes to the TS to have NRC notification go to the NRC Operations Center.
TS 6.5.a (2) will be revised to read:
(2)
The safety limit violation shall be promptly reported to the NRC. Prompt reporting of the violation shall be made by MU, by telephone or email, to the NRC Operations Center no later than the following working day; TS 6.5.b (2) will be revised to read:
(2)
The release of radioactivity shall be promptly reported to the NRC. Prompt reporting of the violation shall be made by MU, by telephone or email, to the NRC Operations Center no later than the following working day; TS 6.5.c (1) will be revised to read:
(1)
The abnormal occurrence shall be promptly reported to the NRC. Prompt reporting of the violation shall be made by MU, by telephone or email, to the NRC Operations Center no later than the following working day; Page 5 of9
12.8 Section 12.4.3, Other Reportable Occurrences, TS 1.1, Abnormal Occurrences and TS 6.S.c, Other Reportable Occurrences.
- a.
The information in the SAR and proposed TS should, at a mlmmum, meet the recommendations of ANS 15.1.
The TS lists actions to take in the event of Abnormal Occurrences. However, the text does not require the return to a safe condition or reactor shutdown. Please address this discrepancy.
TS 6.S.c (3) will be revised to read:
(3)
The reactor shall be shutdown or placed in a safe condition and a return to normal reactor operations will not be allowed until authorized by the Reactor Manager.
12.8 Section 12.4.3, Other Reportable Occurrences, TS 1.1, Abnormal Occurrences and TS 6.S.c, Other Reportable Occurrences.
- b. Some of the proposed abnormal occurrence definitions are different than the recommendations of ANS 15.1. For example, TS 1.1 e only considers it an occurrence if radiation exposure limits are exceeded. The NRC needs to be informed promptly regarding any abnormal and significant degradation of the fuel, cladding, primary coolant boundary or containment boundary regardless of the radiological impact. TS 1.1 e. only considers the primary coolant boundary. Provide justification for not also considering the pool system coolant boundary.
Propose appropriate changes to the SAR and TS to meet the recommendations of ANS 15.1 or justifY your TS and SAR as proposed.
The ANS lS.1 recommendation in 6.7.2 (l)(v) states "abnormal and significant degradation in the reactor fuel or cladding, or both, coolant boundary, or containment boundary (excluding minor leaks) where applicable." MURR added specific limits of radiological consequences to better define what was meant by the word "significant" since this is such a subjective term. The words "which could result in exceeding prescribed prescribed radiation exposure limits of personnel or environment, or both" can be deleted if the NRC feels it adds nothing to the definition.
The pool coolant system should not be included in Technical Specification 1.1.e. The pool coolant system serves to remove heat from in-pool components and as shielding from direct core radiation. The pool being an open water system does not serve as one of the three levels of containment - cladding, primary coolant system and the reactor containment building - and is not assumed as a boundary in any accident analysis.
APPENDIX A, TECHNICAL SPECIFICATIONS A.1 Definitions, General. The information in the SAR and proposed TS should, at a minimum, meet the recommendations of ANS 15.1. Definitions for channel calibration, channel check, measured value, operating, protective action, reactivity worth of experiment, reference core condition, research reactor, research reactor facility, shall, should and may, and unscheduled shutdown are
. missing from the proposed TSs. Please justifY not needing these definitions or add them to the proposed TSs.
MURR feels that no additional definitions are necessary in the Technical Specifications (TS).
The definition of calibration is referred to in the definition of calibration or testing interval (TS 1.2) as "normal checks for accuracy or operability of a system or component."
Page 6 of9
The term operating has no special significance in this document and is commonly defined as "of, relating, or used for or in operation."
The terms research reactor, and shall should and may, have been defined in other licensing-based documents, specifically the MURR Emergency Plan.
MURR uses the term "Operations Boundary" instead of research reactor facility as stated in ANSIIANS-15.1-2007. Operations boundary is also defined in the MURR Emergency Plan.
The terms channel check, measured value, protective action, reference core condition and unscheduled shutdown are not used in this document and therefore were excluded from this document.
MURR administrative procedure AP-RO-IlO, "Conduct of Operations," defines unscheduled shutdown as "Any unplanned automatic shutdown of the reactor, that occurs after all 'Blade Full-In Lights' have cleared, that is the result of an equipment actuation, Operator error, or equipment malfunction, or any unplanned manual shutdown that is an immediate response to conditions that could adversely affect safe operation."
A.23 TS 3.4 a., Reactor Instrumentation.
- a.
The TSs require three radiation monitors be operable.
However, the SAR discusses additional monitors such as beamport floor, south, west, east and north wall, fuel storage vault, mechanical equipment room, fuel element failure monitoring system, and secondary coolant monitoring system. Please proposed TS requirements for these radiation monitors or explain why these monitors need not be TS requirements.
With the exception of the fuel element failure monitoring system, MURR feels that Technical Specification (TS) requirements are not needed for the additional area radiation monitors. The fuel element failure monitoring system is required by TS 3.9.b.
As stated in TS 3.4, the objective of this specification is to provide sufficient reliable information to assure safe operation of the reactor. The three radiation monitors already listed - reactor bridge, reactor containment building exhaust plenum and off-gas -
provide that type of information and they are sufficient for that purpose. The additional monitors mentioned in the SAR provide useful information about conditions in other areas of the facility but are not directly related to the safe operation of the reactor.
A.23 TS 3.4 a., Reactor Instrumentation.
- b. The SAR refers to two bridge monitors while the TS require one. ClarifY which monitor is required by the TSs and justifY why the other monitor isn't required.
A second bridge radiation monitor is installed for added operational reliability. Either monitor can satisfy the requirements of Technical Specification 3.4.a.
This is unchanged from the currently approved Technical Specifications.
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A.23 TS 3.4 a., Reactor Instrumentation.
- c.
The TS require one exhaust plenum monitor while the SAR refers to two monitors. Clarify which monitor is required by the TSs and justify why the other monitor isn't required.
A second exhaust plenum radiation monitor is installed for added operational reliability. Either monitor can satisfy the requirements of Technical Specification 3.4.a. This is unchanged from the currently approved Technical Specifications.
A.23 TS 3.4 a., Reactor Instrumentation.
- d.
The TS refer to the off-gas radiation monitor while the SAR discusses the off-gas radiation monitor system which contains three-channels.
Please clarify if this is the same instrumentation.
Yes, the off-gas radiation monitor listed under Technical Specification 3.4.a is the same monitor as described in Section 7.9.5, Off-Gas Radiation Monitoring System. The monitor, or system, consists of three channels - particulate, iodine and noble gas - in a single, self-contained unit.
A.23 TS 3.4 a., Reactor Instrumentation.
- e.
Please provide maximum alarm set points and the bases for any TS required radiation monitor not covered in TS 3. 7 or justify why TSs are not needed.
The following two Technical Specification (TS) required radiation monitors are not covered by TS 3.7 - the reactor bridge radiation monitor and the reactor containment building exhaust plenum radiation monitor, which are both required by TS 3.4.a. The maximum alarm set points are established in TS 3.5.b, which reads as follows: "While containment integrity is required, the reactor containment building shall be automatically isolated if the activity in the ventilation exhaust plenum or at the reactor bridge indicates an increase of 10 times above previously established levels at the same operating condition." See the answer to Question A.28 for the bases.
A.23 TS 3.4 a., Reactor Instrumentation.
- f.
The off-gas radiation monitor may be out of service for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. In the unlikely event of a release, discuss what other radiation monitoring equipment would alert the staff to the condition.
In the event that the off-gas radiation monitor is out of service, the reactor containment building exhaust plenum radiation monitor identified in Technical Specification 3.4.a would still be in service and would alert operators to increased radiation levels in the reactor containment building ventilation exhaust system.
Some glove boxes and hot cells containing experiments with a potential for off-gas release have individual monitors on their exhaust lines. To cover other potential releases, operating procedure OP-RO-720, "Radiation Monitoring-Stack Monitor Operational Check" contains the following Precaution and Limitation: "Monitors may be taken out of service for up to two hours for calibration and maintenance. During this out-of-service time no experiment or maintenance activities will be conducted which could likely result in the release of unknown quantities of airborne radioactivity." This same precaution and limitation is included as Note (3) in Technical Specification 3.4.a.
Page 8 of9
Historical records show that unexpected releases occur on an extremely infrequent basis at MURR. If operations or experiments are planned that might possibly cause a potentially unusual release of materials into the facility ventilation exhaust system, these operations or experiments would not be allowed or would be tenninated.
A.28 TS 3.5 b., Reactor Containment Building. Please provide a technical basis for containment isolation at 10 times above previously established levels.
Isolation of the reactor containment building at 10 times the normal previously established radiation levels is necessary to allow for sample handling within the reactor pool or when removing samples from the pool. Normal pool surface radiation levels are around 20 mrem per hour while those at the containment building exhaust plenum are around 0.15 mrem per hour.
Operational experience has demonstrated that the 10 times factor provides sufficient margin to minimize inadvertent reactor scrams without allowing for the potential of unacceptable exposure rates to personnel in containment. Ten times the routine dose rates equates to 200 mrem at the bridge monitor and 1.5 mrem at the exhaust plenum. Dose rates at this level do not constitute an unreasonable risk and could not go unidentified for any significant period of time. Radiation monitor indications are recorded at set intervals in the reactor log book and any increase above normal would be identified by and responded to by Reactor Operations.
A.30 TS 3.6 c., Experiments.
This TS controls occupational exposures.
Please discuss limits on exposure for members of the public from experiment failure.
The Reactor Utilization Request (RUR) process as described in Section 10.4, Experiment Review, includes an analysis of the potential for release of radioactive material to the atmosphere.
Potential releases to the general public are limited to the values found in 10 CFR 20, Appendix B, Table I.
A.54 TS 6.1, Organization.
ANS 15.1, Section 6.1.3 (2), Staffing, recommends TSs contain a requirement for control room contact iriformation. Propose appropriate TS wording to include requirements similar to these in the TSs or justifY why they are not needed.
MURR's existing NRC-approved Emergency Plan contains a requirement that contact information for not only operators, but also any other personnel considered part of the Facility Emergency Organization (FEO), be maintained in the Emergency Plan Implementing Procedures, specifically form FM -104, "Emergency Call List." MURR feels that this existing requirement meets the intent of ANS 15.1, Section 6.1.3(2), Staffing, and that no additional Technical Specification is required.
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ATTACHMENT 1 Table S shows the power ratio distribution for this case, which is plotted on Fig. 4.
Two valleys near the bottom and the top of the fuel are due to the high thermal flux peaking at the water reflector region.
- 3.
Safety Limit Analysis MURR limiting power calculations were originally performed using the BOLERO computer code(2) by NUS Corporation in 1973.
Since BOLERO is proprie-tory and rather 01 d program, most recent thenna1-hydrau1 i c code had been searched and in 1983, COBRA-3C/RERTR(3) was purchased from ANL for the MURR upgrade safety limit analysis. COBRA-3C/RERTR is a modified version of COBRA-3C/MIT(4), a thermal-hydraulic subchannel analysis code.
The purpose for modifying the MIT version was to make the code more suitable for research and test reactors which are operated at low pressure and low temperature, and which may use plate type fuel elements like those in the MURR core. It is to be noted that in the Reduced Enrichment Research and Test Reactor (RERTR) program(S), COBRA-3C/RERTR has b~en extensively used fo~steady state'
- thennal-hydrau1 ic analysis for the various research reactors.
After acquiring COBRA-3C/RERTR, extensive efforts have been made by pre-vious rese~rchers(6,7) to incorporate the iterative procedure in limiting power calculations and to improve COBRA thennal-hydrau1ic model for the MURR analysis. In the MURR version of COBRA, an iterative scheme to determine the core power level that would cause DNB or flow instability was incorporated.
When inita1 guessed power level is input to the code, the limiting power level is calculated by increasing or decreasing the power level until the specified constraint is reached. Therefore the unique feature of the MURR version of COBRA, COBRA-3C/MURR, is that it can calculate limiting power based on 1) Departure of Nuclear Boiling (DNB), 2) Onset of Nuclear Boiling, and 3) Onset of Flow Instability.
12 ATTACHMENT 1
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I The pressure drop, dP, from the pressurizer to the MURR core inlet and/or outlet was calculated using the computer program, EXPRESS(7) (see Appendix B).
Using the reference pressure drop data in Table 6(8), individual dP components are corrected to the new flow rate, temperature~ and other core condition.
Fornon_-frictional components, was used where the subcript 0 denotes the reference conditions as given in Table 6.
For the friction loss components, the pressure drop was assumed to be given by the Blasius equation, i.e.,
I p (T) ~ o. 8 f Q \\1. 8( II (T),\\0. 2 dP =dPo ' -: I-! ! -
\\
1-p(T or
'- Qo; l ll(T 0).'
where p, Q~ and II are water density, flow rate, and viscosity, respectively.
The new dP components were then summed and the result was subtracted from the specified pressurizer operating pressure to get the absolute pressure at r'
core inlet and/or outlet.
In COBRA limiting power calculations, core inlet pressure is-usually u~ed as input and the pressure drop in the core is cal-culated in the code.
3.1 Safety Limit Criteria The objective of the safety anslysis is to determine the maximum allowable reactor power level which avoids fuel burnout(or DNB) and flow instability. Also exit coolant temperature at the core exit should be below thf! saturation temperature at the core exit pressure to prevent any bul k boiling. The basis for specifing criteria(9) is described as follows.
The MURR fuel geometry is simil ar to the Avdanced Test Reactor(ATR) fuel elements.
In 1963, the ATR fuel tests were performed at Argonne on 15
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Tabl e 6 Reference Pressure Drop Data
- Component **
6Po (PSI)
Qo(GPM)
To(F)
Frictional In Core 1,2,3 3.259
- 1800 155
- Yes, No 4
0.2689 1800 155
-No No 5,6",,10 4.08 3600 155 Yes No 11 0.1977 3600 155 Yes No 12 0.8980 3600 155 No No 13 12.35 3600 165 Yes Yes
- Data from reference (8)
- Component description using notation of reference (8)
- 1.
Across pressurizer surge line to pressurizer outlet
- 2.
Across 5 feet of 8 inch pipe
- 3.
Across 8 inch Y strainer
- 4.
Across 8 inch/12 inch expansion
- 5.
Across 80 feet of 12 inch pipe
- 6.
Across four 12 inch 90 degree elbows
- 7.
Across three 12 inch 45 degree elbows
- 8.
Across one 12 inch butterfly valve (5078)
- 9.
Across one 12 inch swing check valve (502)
- 10.
Across entrance to annular pressure vessel
- 11.
Across 6 feet of annular pressure vessel
- 12.
Across entrance to fuel element plates
- 13. Across core *** 25.5 inches of fuel element plates ** to core exit 16
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Ii II II II II II II II Ii II I~
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thr~e channel thicknesses (0.054 11,0.072", 0.094"), and it was found that for the two thinnest channels, the burnouts were due to hydraulic instabil-ity when the coolant reached saturation at the exit. Subcooled burnout occurred for the 0.094" channel before the coolant reached the saturation conditions at channel exit. Thesubcooled burnout heat flux data obtained in these tests were 0.6 of the burnout heat flux predicted by the Bernath cor-relation. Full-scale ATR testing at Battelle Northwest with a channel thickness of 0.07" confirmed earlier test results; namely, that burnout induced by hydraulic instability was the limiting factor for ATR.
In addi-tion, it was established that the hydraulic instability condition did not correspond to initiation of local boiling, but to the beginning of bulk boiling at the channel exit in the region where the coolant enthalpy was highest. Test results also indicated that lateral mixing(in the channel) was quite small.
The MURR fuel channel length is about one-half that of ATR and the investigators have shown higher or equal burnout heat flux levels for shorter channel length. Similarly, the shorter channel length are less susceptible to the hydraulic instabilities related to incipient bulk boil-ing. Therefore, the use of the ATR test results will provide conservatism for MURR.
One additional criterion applied in this analysis is the Ledinegg type of flow instability(3). This phenomenon can occur when system pressure drops increase with decreasing flow rate.
For example, boiling increases channel pressure losses and therefore, when a constant heat input is supplied to the system with a small decrease in flow rate, further boiling will occur w{th further increase in pressure loss. As a consequence, the flow rate con-tinues to decrease as the pressure loss increases until a region of unstable operation or burnout occurs.
In COBRA, The Ledinegg flow instability is
. 17
monitored by the Flow Instability Ra"tio(FIR), which is defined as the ratio of the surface heat flux(qi) at which the flow instability will occur, to the average nominal heat flux(q).
The qi is calculated using the Bowring equation:(3) qi -= 1/4 RU (T_s-T i) p C Dh/Lh '
FIR = qi/q R = II (1 +nDh/Lh) where U, Ts, Ti, p, C, Dh, Lh, De, G are respectively coolant velocity, saturation temperature, inlet temperature, coolant density, specific heat, diameter based on heated perimeter, heated length, diameter based on wetted perimeter, and coolaht flow rate. Several options are available for nand it was known(7) that the correlation(n=25) by Whittle and Forgan was the best fit for the operating condition of MURR.
Based on the above, in the MURR safety analysis, the following three safety limit criteria were used, i.e.,
a) The local heat flux' at any point in the core shall be less than 0.5 of the DNB heat flux obtained from the Bernath correlation at that
- spot, b) The Ledinegg flow instability must be avoided even if DNB criterion is
- met, c) The coolant exit temperature from the ho~ channel shall be less than the saturation temperature at the core exit pressure.
The above burnout heat flux limitation is adopted to provide additional d~sign safety margin by a reduction of the correlated ATR test data by the factor of 0.5/0.6 relative to th.e original Bernath correlation. The bulk boiling limitation is adopted to exclude occurrence of the in-core hydraulic 18
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instabilities related to incipient bulk boiling. In the limiting power cal-culations, the ONBR was set to 2.0 for the Bernath correlation and the FIR was set to 1.0 for the Ledinegg flow instability.
3.2 COBRA Runs. for the NUS Study To compare the previous BOLERO results for the current 6.2 Kg core, axial power distribution (Fig. 5) used in the NUS study was input to COBRA and the safety limit calculations were repeated. Table 7 presents a sum-mary of hot channel factors used in the NUS analysis. Since the COBRA input format was a quite different from the BOLERO input, a sl i ght mod ifi cati on was necessary.
For the flow related factors, overall product was calculated to be 1.00 x 1.00 x 1.00 x 1./1.08 x 1./1.05 (=0.882).
Channel factor bn the effective flow area was taken care of in the COBRA input.
BOLERO differentiates between hot channel factors on enthalopy rise and those on heat flux.
The overall product on enthalopy rise is 2.72, where the oVerall product on heat flux is 4.35.
The difference is the axial nuclear ~eaking factor and the greater effect fuel thickness/width variation has on local heat flux (1.15) compared to enthalpy rise (1.03).
In the NUS study, BOLERO was used,to' get the power 'level corresponding to a DNBR=2.0=
critical heat flux/"BOLERO" hot spot heat flux.
For COBRA the hot channel factor of 2.72 and the axial power distribution can be given in the input.
But the heat flux then has the fuel thickness/width variation engineering factor of 1.03 for the hot channel instead of the local hot spot engineering factor of 1.15. Therefore, the "BOLERO" hot spot heat flux = (1.15/1.03)
"C,OBRA" hot spot heat flux. This shows for COBRA, the local hot spot engi-neering factor for fuel thickness/width variation can be taken into consid-eration by the ONBR, 19
lttlttl II t! Hjt 12 IIUHI II HI I rtHUJ
~ammnlHmn tl IH Figure 5 Power Distribution USed in the NUS Study Ul1tJ
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-H1I~lIHHtJ f/'HttHtu immHtiU L8 I. lnchea from top of fueled.~tion t!l 14
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Table 7 Summary of MURR Hot Channel Factors Used in the NUS Study On Enthalpy Rise **********
Power-related Factors -
Nuclear Peaking Factors
]adial ************************
Local (Circumferenctial) * * * *
- Non-uniform Burnup ******************
Axial ************************
Engineering Hot Channel Factors on Enthalpy Rise Fuel Content Variation ****************
Fuel Thickness/Width Variation ************
Overall Product *********************
Flow-Related Factors Core/Loop Flow Fraction ********
Assembly Minimum/Average Flow Fraction *****
Channel Minimum/Average Flow Fraction Inlet Variation ****************
Width Variation *******************
Thickness Variation * * * * *
- Within Channel Minimum/Average Flow Fraction Thickness Variation *****************
Effective Flow Area *****************
Overall Product. * * * *
- On Heat Flux **********
Power-Rel ated Factors' Nuclear Peaking Factors Radial * * * * * *
- Local (Circumferenctial) ********
Non-uniform Burnup *********
Ax; a 1 *************
Engineering Hot Channel Factors on Flux Fuel Content Variation ****************
Fuel Thickness/Width Variation **********
Overall Product **********************
Energy Fraction Generated in Fuel Plate '0 * *......
21 2.220 1.040 1.112 1.000 1.030 1.030 1.000 1.000 1.000 1.000 1./1.080 1./1.050 0.3231/0.3505 0.81 2.220 1.040 1.ll2 1.432 1.030 1.150 0.930 2.72 4.35
II 11 11 II II 11 J
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1.15 Critical Heat Flux ONBR = 2.0 --- = 2.23 = ------------.--
1.03 COBRA Hot Spot Given in Table 8 and plotted in Fig. 6 and 7 are the llmiting power levels calculated by BOLERO for the pressurizer at 60 and 75 psia. The underscoTed table entries in Table 8 are the power limits established by the criterion of avoiding any flow instability of the coolant. Coolant flow rate ranges from 400 to 4000 gpm and the coolant temperature varies; from 120°F and to 200°F.
The BOLERO values are compared to COBRA results in Table 9, and plotted in Fig. 8 and 9.
It was found that at the MURR operat-ing condition, there is a good (agreement between the COBRA and the BOLERO results. However, most of the COBRA values came ~ut to be less than those of the BOLERO result; which indicates that the COBRA results provide conserva~
tism to the current safety limit curves.
When compared to the BOLERO results, more cases in the COBRA results are 1 imited by the Ledi negg flow instability criterion. -This indicates that even if the coolant temperature at the core exit is below the saturation temperature, there is a possibility that the Ledinegg flow instability can occur.
3.3 COBRA Runs for the Mixed Fuel Combinations Using the axial power distributions obtained from the 30 BOLO VENTURE runs, limiting power calculations for the three fuel combinations were per-formed. Uncertainty factors considered for the COBRA runs were:(7) a) U-235 content variation in UAlx; 4.4%
b) U-235 weight variation in each fuel plate; 1.0%
c) Fuel thickness/width variation; 3.0%
d) Power variation in circumferential direction; 4.0%
22
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Temperature I
Oeg F 400.
120.
2.40 I
140.
160.
180.
.2:12 1.82 I.52 I
200.
1.26 I
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I Oeg F 120.
400.
2.60 140.
2.32 I
160.
180.
2.02
- 1. 74 200.
1.46 I
Table 9 COBRA Safety Limits for the NUS Power Shape MURR Safety Limits for the Pressurizer at 60 PSIA r Flow Rate (GPM) 800.
1200.
1600. 2000.
2400.
2800.
3200.
4.70 6.66 8.50 10.30
' 12.02 13.68 15.20 4.22 6.12 7.84 9.48 11.06 12.56 14.00
. 3.62 5.42 7.10 8.60 10.00 11.30 12.58 3.06 5.~6 6.04 7.48 8.86 10.04 11.28 2.50*
- 3. 72
~.84 5.98 7.08 8.12 9.10 MURR Safety Limits for the Pressurizer at 75 PSIA Flow Rate (GPM) 800.
1200. 1600.
2000.
2400.
2800.
3200.
5.08 7.20 9.22 11.18 13.04 14.86
' 16.64 4.64
'6.68 8.56 10.36 12.12 13.82 15.38 4.06 6.06 7.80 9.48 11.00 12.54 14.00 3.48 5.20 6.90 8.56 10.00 11.38 12.70 2.00 4.34 5.68 7.04 8.36 9.64 10.86 I NOTE; Underlined power levels are limited by Ledinegg flow instability.
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I 26 3600.
4000 *.
16.68 18.08 15.36 16.64 13.78 14.88 12.34 13.32 10.00 10.82 3600.
4000.
18.34 19.96 16.94 18.42 15.40 16.74 13.96 15.16 12.04 13.14
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Fig. 8 COBRA RESULT FOR THE NUS POWER SHAPE AT 60 PSIA MURR SAFETY LIMIT CURVES FOR PRESSURIZER AT 60 PSIA 26 25 24
( NUS POWER SHAPE USED )
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22 21 M 20 A
X 19 1
M 18 U
M 17 A 16 L
'0 W 14 A
B 13 L
E 12 C 11 a
R 10 E
9 p a 8
w E
7 R
0.0 0.4 0.8 1.2 1.6 2.0 2.4 2.8 3.2 3.6 4.0 CORE FLOW RATE.lOOO GPM LEGEND& P
- 2
... +... 3
--- 4
-.+-- 5
...... 6 27
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I Fig~ 9 COBRA RESULT FOR THE NUS POWER SHAPE AT 15PSIA MURR SAFETY LIMIT CURVES FOR PRESSURIZER AT 75 PSIA 26 25 24 23 22 21 M 20 A
X 19 1 M 18 U
M 17 A 16 L
L 15 a
W 14 A
8 13 L E 12 C 11 o
R 10 E
9 P o 8 W
E 7
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6 M
W 5 4
3 2
( NUS POWER SHAPE USED )
0.0 0.4 0.8 1.2 1.6 2.0 2.4 2.8 3.2 3.6 4.0 CORE FLOW RATE.1000 GPM LEGENDI P
- *
- 2
... 3
-.-- 4
5
...... 6 28
ATTACHMENT <2
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16 r FIQJRE H.l 15
~1URR DN13 CURve
~DDEI OPER4TION 14 PRESSURIZER AT 75 PSIA lSr 12 L I
I lU-I Inlet Water loL Temperature, OF
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Core Flow Rate, 1000 gpm 0.4 O.S 1.2 1.6 2.0 2.4 2.8 3.2 3.6 4.0 ATTACHMENT 2