ML17054B786
| ML17054B786 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 07/24/1985 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML17054B785 | List: |
| References | |
| NUDOCS 8508070363 | |
| Download: ML17054B786 (18) | |
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>>0~44 UNITEDSTATES NUCLEAR REGULATORY COMMISSION NIASHINGTON, D. C. 20555 Enclosure 3
SAFETY EVALlIATION NINE NILE POINT 1
PLANT EFFECTIVENESS OF EXISTING CnRE SPRAY SPARGER IN A STEAM ENVIRONMENT 1.0 Introduction and Background The following report presents the staff's evaluation of the effectiveness of the Nine Mile Point 1 core spray sparger in a steam environment.
The evaluation is based on a review of a topical report submitted by the licensee (Reference 1), responses to staff requests for additional information, and meetings between the staff and the licensee.
2.0 Discussion To address the problem of spray coolino effectiveness the licensee has pursued a program with General Electric Company which includes both tests and analysis.
The objectives of this program were to identify the minimum bundle spray flow rate which suppnrts the 10 CFR 50 I5ppendix K heat transfer coefficients, and to demonstrate that flow to all bundles durinq spray cooling would be greater than the minimum required flowrate.
An analytical model, qualified with experimental data was used to determine the required minimum bundle flow rate.
Results obtained with this model show that predicted heat transfer is equivalent to the Appendix K assump-tions at a bundle flowrate of 1.8 GPM.
General Electric's semi-emperical design methodology was used to determine the spray flow distribution in a steam environment.
This method combines plant specific si~e nozzle and full scale sparger test data with analytical results to pr'edict the minimum bundle flowrate in the core.
The methodology has been successfullv benchmarked aaainst 30'ector steam tests of a RWR/6 Core Sprav Sparaer (Reference 2).
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The minimum bundle flow rate predicted with the methodology including uncertainties, is The calculation of this result assumes sinale sparger (lower) operation in a
steam environment.
This value is above the required flowrate of On this basis the licensee concludes that the assumption of Appendix K heat transfer coefficients in the Nine Nile Point 1
ECCS evaluation model remains valid.
3.0 Evaluation The purpose of the evaluation is to determine whether the effects of steam environment on core spray distribution degrade the minimum bundle spray flowrate to the extent that the assumption of 10 CFR 50 Appendix K
heat transfer coefficients in LOCA analysis becomes invalid.
The bases for the evaluation are analyses submitted by the licensee and review of other tests and analyses germane to the issue.
The evaluation is divided into two parts.
The first section addresses the minimum bundle spray flowrate expected in a steam environment.
The second section addresses the question of how large a bundle flowrate is needed to achieve Appendix K heat transfer coefficients.
3.1 minimum Available Bundle Spray Flowrate in Steam The General Electric design methodology has been used tn determine the spray distribution in a steam environment.
The method treats condensation and drag effects as well as effects of interacting sprays.
The methodology was developed under the key assumption that the condensation (thermodynamic) effects.in.a steam environment can be handled independently from the hydro-dynamic effects of multiple nozzle interactions.
Under this assumption the'hermodynamic effects are evaluated from single nozzle spray distribution tests in a simulated steam environment.
The hydrodynamic effects of multiple nozzle interactions are determined from single nozzle and full scale sparger spray distribution tests in an air environment using simulator nozzles.
Simulator nozzles are specifically developed spray nozzles which produce spray patterns in an air environment similar to corresponding reactor nozzle spray patterns in a steam environment.
The methodology has been successfully benchmarked aaainst 30'ector steam tests of a BWR/6 core spray sparger (Reference 2); and approved by the staff (RSB) for use in RWR/6 design (Reference 3).
1n addition, model predictions have been made for the RWR/455-218 sparger design and compared with full scale 30'ector steam tests of that design '(Reference 4).
The comparison shows that as with BWR/6, predictions and test results agree very well within a certain error bond.
The BWR/4 comparison is also significant in that it confirms the method's ability of treat differences in nozzle and sparger design with variation in input.
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The licensee's methodology has been used successfully in evaluation of other sparoer designs and has been benchmarked with core sprav tests in a steam environment.
We therefore find it acceptable.
Cnre Spray Soar er Flowrates Sparger and individual nozzle flow rates assumed in the analyses were based on the following:
I.
Core Spray system performance data (pump characteristics).
2.
Spray distribution tests performed in 1968 with a full scale mock-up of the Nine Mile Point I sparger (nozzle flowrates).
3.
Calculated system pressure drop characteristics of the deliverv pipina includina the ring spargers.
We find the licensee's approach in determinino system and nozzle flnwrates acceptable since it is generally based on test data.
The total single core spray sparger flowrate in 30 psia steam has been determined to be 5020 GPH based on tests conducted at the plant.
This flow rate is considerably larger than the original single sparger desion flowrate of 3400 GPM.
The flow difference is the result of different assumed reactor system pressures (i.e.
30 psia backpressure vs.
125 psia for rated flow).
We note that actual sparger flowrate determined at the plant must be shown to bound that assumed in plant safety analyses as per plant Technical Specification 3.4.
We have reviewed the licensee's surveillance test procedure.for the Core Spray System and conclude that it properly ensures the availability of the required system flowrate.
Uncertainty Anal sis In deriving the minimum bundle flow rate the licensee has accounted for the uncertainties in the calculational and experimental steps of the desion methodology.
To evaluate the uncertainties we have compared the overall uncertainty factor of with the difference between predictions and tests documented for BMR/4, BWR/5 and BWR/6 systems (Reference 3 and 4). The comparison shows that the analysis methodology generally overpredicts bundle flowrates and that the differences between predicted and measured bundle flowrate generally falls in the range between
?5X and 50Ã of measured flow.
Based on this comparison we conclude that the licensee's uncertainty factor is acceptable for application to results obtained with the desiqn methndoloav discussed above.
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Effects of Hi her System Pressure Based on the uncertainty in spray flow at high system pressure the licensee The Reactor Systems Branch in the Oivision of Systems Integration has reviewed the methods and similar analyses performed for Oyster Creek and found them both acceptable (Reference 9).
Due to the similarities between Nine Mile Point I and Oyster Creek in design and response to a
LOCA it follows that the conclusions regarding Oyster Creek also apply to Nine Mile Point l.
3.2 Minimum Required Bundle Spray Flowrate
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RANGE OF PARAMETEPS IN GE AND SWEDISH SPRAY COOLING TESTS PARAMETER GE Tests RANGE Swedis ests initial Rod Temperature Bundle Power Spray Flow Rate Pressure In addition to the tests performed by General Electric and Riso National Laboratory (Swedish) other independent tests in this country and in Japan support the CORECOOL result.
Tests by Exxon Nuclear Company (Reference
- 7) on a
flowrate as being sufficient to provide Appendix K heat transfer coefficients.
.laoanese tests (Reference 8) on an 8
X 8 full scale RWR bundle show that Appendix K heat transfer coefficients are obtainable at spray flowrates between 1.0 and 1.8 GPM/bundle depending on rod location.
We conclude that the licensee's determination of required minimum bundle flowrate is acceptable because it is well supported by experimental data.
4.0 Conclusion and Findin s
Based on the analyses presented by the licensee and our review of those analyses we conclude that core spray distribution in the Nine Mile Point 1
plant is aFfected by the presence of a steam environment.
The anal.vses presented by the licensee indicate the minimum bundle flowrate including uncertainties is based on a single core spray sparger flowrate nf 5020 GPM.
Original full scale tests in air showed a minimum bundle flowrate of 2.45 GPM based on a single sparger flowrate of 3400 GPM.
Specific conclusions and maior findings of the staff review are given below.
1.
The minimum bundle flowrate of predicted in the licensee's analysis was arrived at using the General Electric design methodology for determining core spray distribution.
We have reviewed the licensee's analysis and find it acceptable.
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In order to demonstrate acceptable conseauences for certain small break loss-of-coolant accident scenarios, The staff (RSBIDSI) has reviewed the analyses as part of its review of the Oyster Creek Cycle 10 reload core analysis and found them acceptable.
3.
The licensee has determined that the minimum required bundle spray flowrate needed tn achieve Appendix K heat transfer coefficients is We accept this determination because it is well supported by test dat'a.
4.
We have reviewed the uncertainties applied in the determination of minimum bundle spray flowrate and find them acceptable based on comparison of test results with analytical results.
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4 5.0 References 1.
Letter and Enclosures from Mr. C. V. Mangan (Niagra Mohawk)
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to Mr. Domenic B. Vassallo (NRC) Seotember 13, 1~83.
2.
NED0-24712, "Core Spray Design Methodology Confirmation Tests,"
August
- 1979, General Electric Topical Report 3.
NRC memorandum, P.
S.
Check to R. L. Tedesco September 8,
1980 4.
NUREG/CR-1707, "BWR Refill - Reflood Program Task 4.2-Core Sprav Distribution Final Report," March 1981.
5.
Letter and Enclosure from R. L. Gridley (General Flectric) to M. Caruso (NRC); August 16, 1983.
6.
L. Nilsson, L. Gustafson, R. Hariu, "Experimental Investigation of Cooling by Top Spray and Bottom Flooding of a Simulated 64 Rod Bundle for a BWR," STUDSVIK/RL - 78/59 June 1978.
7.
XN-NF-78-12 "Spray Cooling Heat Transfer Test Results Phase II Facility and Test Data ENC 8 X 8 BWR Fuel," Exxon Nuclear Company Proprietary Report; May 1979.
8.
M. Naitoh, "Heat Removal by Top Sprav Emergency Core Cooling,"
presented at Second Two-Phase Heat and Mass Transfer Symposium Miami, FL; April 1979.
9.
NRC Memorandum from Brian Sheron to Carl Berlinger, Subiect:
Oyster Creek LOCA Analysis for Reload Fuel Application, June 5, 1984.
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BASES FOR 3.1.4 AND 4.1.4 CORE SPRAY SYSTEH The core spray system consists of two automatically actuated, independent, double-capacjty systems capable of coolina reactor fuel for a ranae of loss-of-coolant accidents.
Each core spray system has 100 percent cooling capacity from each spray header and each supply pump set.
Thus, specifying both systems to be fully operational will assure to a high degree core cooling capability if the core spray system is required.
Allowable outages are specified to account for components that become inoperable in both systems and for more than one component in a system.
k Both core spray systems contain redundant supply pump sets and blocking valves.
Operation of one pump set and blocking valve is sufficient to establish required delivery rate and flow path.
Therefore, even with the loss of one of the redundant components, a system is still capable of performing its intended function.
If a redundant component is found to have failed, corrective maintenance will begin promptly.
Nearly all maintenance can be completed within a few days.
Infrequently, however, major maintenance might be required.
Replacement of principal system components could necessitate outages in excess of those specified.
In spite of the best efforts of the operator to return equipment to service, some maintenance could require up to 6 months.
.In determining the operability of a core spray system the required performance capability of its various components shall be considered.
For example:
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Periodic tests>>ill establish a core spray pump set head/flow condition which will be measured against a
design basis pump curve to insure adequate core cooling is provided (Section VII)>> (HEOE-3024I).
2.
The pump shall be capable of automatic initiation from a low-low water level signal in the reactor vessel or a high containment pressure signal.
The blocking va'Ives shall be capable of automatically opening from either a low-low water signal or high containment pressure signal simultaneous with low reactor pressure permissive signal.
(Section VII)*
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BASES FOR 3.1.4 AND 4.1.4 CORE SPRAY SYSTEH 3.
Backup diesel-generator power shall be available to all motor-operated components.
Instrumentation has been installed to monitor the integrity of the core spray piping with1n the reactor pressure vessel.
Following installation of this instrumentation, the requirements stated in Specification 4.1.4 d shall be followed.
The testing spec1fied for each major refueling outage will demonstrate component response upon automatic system initiation.
For example, pump set starting (low-low level or high drywell pressure) and valve opening (low-low level or high drywell pressure and low reactor pressure) must function, under simulated conditions, 1'n the same manner as the systems are required to operate under actual conditions.
The only differences will be that demineralized water rather than suppression chanber water will be pumped to the reactor vessel and the reactor will be at atmospheric pressure.
The core spray systems are designed such that demineralized water is available to the suction of one set of pumps in each system.
. (Section VII-Figure VII-1)*
The system test interval between operating cycles results in a system failure probability of l.l x 10 (Fifth Supplement, page 115) and is consistent with practical considerations.
The more frequent compon-ent testing results in a more reliable system.
At quarterly intervals, startup of core spray pumps will demonstrate pump starting and operability.
Ho flow will take place to the reactor vessel due to the lack of a low-pressure permissive signal required for opening of the blocking valves.
Flow, instead will be re-cycled to the suopression chamber via a
test loop.
A flow restricting device has been provided in the test loop which will create a pressure loss for testing of the system.
In addition, the normally closed power operated blocking valves will be manually opened and. re-closed to demonstrate operability.
The intent of Spec1f1cat1on 3.1.4 f is to allow control rod drive maintenance and LPRH replacement at the t1me that the suppression chamber is unwatered and to perform normal fuel movement acti vi ties in the refuel node with an unwatered suppression chamber.
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