ML20113F640
| ML20113F640 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 09/30/1989 |
| From: | Waterman M EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20092F288 | List:
|
| References | |
| CON-IIT05-181-90, CON-IIT5-181-90, RTR-NUREG-1410 EGG-EAST-8681, NUDOCS 8909270212 | |
| Download: ML20113F640 (34) | |
Text
,
EGG EAST 8681 TECHfilCAL EVALUAT10f4 REPORT WCAP 11916
" LOSS OF RHRS COOLING WHILE THE RCS IS PARTIALLY FILLED" Michael E. Waterman Published September 1989 EG&G Idaho, Inc.
Idaho falls, Idaho 83415 Prepared for the U
S. Nuclear Regulatory Commission Washington, D. C. 20555 Under DOE Contract No. DE-AC07 761001570 FIN No. D6039
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ABSTRACT A technical evaluation of Westinghouse Report WCAP 11916
" Loss of RHRS
[ Residual Heat Removal System) Cooling while the RCS [ Reactor Coolant System) 1s Partially filled," was performed. WCAP-11916 addresses plant operations with the RCS partially filled; describes evaluations of air entrainment potential as a function of RCS level RHRS/ hot leg dimentional characteristics and RHRS flow rate; presents the results of thermal hydraulic evaluations of postulated losses of RHRS capability; and discusses the application of the analyses for developing operatir.g procedures for individual plants.
Overall, the broad scope of WCAP 11916 indicates that Westinghouse and the Westinghouse Owners Group recognize that the potential for loss of RHRS capability exists, that losses of RHRS capability can lead to serious complications in cooling the reactor core, and that this issue must be addressed in a timely manner.
The plant operations description section addresses existing plant prac-tices, with the emphasis on plant equipment availability.
However, some of the better techniques used by licensees to prevent losses of RHRS capability when the RCS is drained down were not included in this discussion.
The air entrainment experiments yielded correlations for the level at which air entrainment into a RHRS line could be expected as a function of RHRS flow.
These correlations should be used as a scarting point for specifying the minimum operating levei as a function of RHRS flow rate for mid-loop operations.
Some additional level should be added to the correlation values because nuclear plant test data have shown that higher operating levels in plants have resulted in RHRS air entrainment.
The thermal-hydraulic analyses address a broad range of scenarios for postulated loss of RHRS capability.
A scenario whereby the RCS could pressurize then blow down through a failed steam generator nozzle dam with a resulting rapid core uncovery needs additional evaluation.
The methodologies for applying the results of the thermal hydraulic analyses to individual plants are clearly written and well conceived.
r ti
SUMMARY
There have been numerous losses of decay heat removal capabilities at plants in the United States.
As a result of the Diablo Canyon event on April 12, 1987, the Nuclear Regulatory Commission (NRC) requested that each pressurized water reactor (PWR) licensee address several important issues.
One of these issues concerned the analytical basis of operating procedures for mid loop operations.
In response, the Westinghouse Owners Group (WOG) contracted the Westinghouse Electric Corporation to perform tests and analyses to address mid-loop operations.
The resulting report, WCAP 11916, is the subject of this technical evaluation.
WCAP-11916 summarizes current RHRS operating practices and equiptrent availability, but does not discuss many of the precedures used by severr.1 of the Westinghouse licensees with respect to loss of RHRS capability.
Soie applicable changes to Technical Specifications are provided, but the content is similar to technical specifications already used at some operating plants.
A program was initiated by the WOG to study air entrainment into Residual Heat Removal (RHR) systems.
This program consisted of (1) a literaturc search for information about air entrainment at pipe intakes, (2) review of pitnt test data, and (3) a series of experiments to determine the feasibility of specifying operating limits to prevent air entrainment during mid loop operating conditions.
Westinghouse performed experiments to simulate different RHRS/ hot leg configurations'to determine the effect of Reactor Coolant System (RCS) level, RHRS coolant flow, and RCS/RHRS system geometry on the initiation of air entrainment at the intake of typical Residual Heat Removal (RHR) systems.
Westinghouse considers the resulting experiment correlations to be conservative, but nuclear power plant test data do not fully support this conclusion.
Thermal-hydraulic analyses were performed by Westinghouse using sinplified models of Westinghouse 2-loop, 3 loop, and 4-loop PWRs.
The purpose of the analyses was to evaluate recovery strategies for different scenarios for loss-of RHRS-capability during mid-loop operations.
The scope of the scenarios was well-conceived.
However, recove y actions for a scenario in which the core rapidly uncovers as-a result of a primary system repressuri-zation followed by a failure of a cold leg steam generator nozzle dam may require additional analysis effort.
The thermal hydraulic model detail is too coarse to conclusively determine that a core recovery without fuel damage would occur using one safety injection pump with hot leg injection.
It is postulated that, if core damage occurs, the coolant flowing down through the core may transport fission products from damaged fuel rods out the steam generator manway into the containment.
Consequently, additional analyses to evaluate core level response for this scenario using a more detailed thermal hydraulic model of the reactor core is recommended, iii
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f Hethodologies for extending the results of the suggested recovery strat-egies are clearly written and well-conceived and provide the licensees with more than one option that is applicable to their operating procedures.
Overall, the broad scope of WCAP llg16 indicates that Westinghouse and the Westinghouse Owners Group recognize that the potential for loss of RHRS capability exists, that losses of RHRS capability can lead to serious complications in cooling the reactor core, and that this issue must be addressed in a timely manner.
l-l l
iv
CONTENTS ABSTRACT................................................................
11
SUMMARY
................................................................. iii 1.
INTRODUCTION.......................................................
I 2.
PLANT OPERATION WITH LOOPS PARTIALLY FILLED........................
2 3.
AIR ENTRAINMENT EXPERIMENTS........................................
4 4.
THERMAL-HYDRAULIC ANALYSES.........................................
21 5.
CONCLUSIONS........................................................
24 6.
REFERENCES.........................................................
26 FIGURES 1.
Critical submergence level and plant operating level data for Category A.1 plants (0 degree orientation, RHR/HL 10 = 5.2/16.1 inches)..............................................................
9 2.
Critical submergence level for Category A.2 plants (0 aegree orien-tation, RHR/HL 10 8.5/27. Sinches)..................................
10 3.
Critical submergence level and plant operating level data for Category B.1 plants (45 ( jree orientation, RHR/HL 10 = 6.8/27.5 inches)..............................................................
11 4.~
Critical submergence level and plant operating level data for Category B.1 plants (45 degree orientation, RHR/HL 10 = 7.0/29.0 inch0s)..............................................................
12 5.
Critical submergence level and plant operating level data for Category B.2 plants (45 degree orientation, RHR/HL ID = 8.8/29.0 inches)..............................................................
13 6.
Critical submergence level and plant operating level data for Category B.3 plants (45 degree orientation, RHR/HL 10 = 10.5/29.0 inches)..............................................................
14
.7.
Critical submergence level and plant operating level data for Category 0.4 plants (45 degree orientation, RHR/HL ID = 11.2/29.0 inches)..............................................................
15 8.
Critical submergence level and plant operating level data for Category B.4 plants (45 degree orientation, RHR/HL ID = 11.5/29.0 inches)..............................................................
16 v
l
l 9.
Critical submergence level and plant operating level data for Category B.4 plants (45 degree orientation, RHR/HL ID - 11.8/29.0 inches)..............................................................
17
- 10. Critical submergence level for Category C plants (55 degree orien-tation, RHR/HL ID - 10.5/29.0 inches).............................
18
- 11. Critical submergence level for Category D plants (60 degree orien-tatica RHR/HL ID - 11.5/29.0).......................................
19
- 12. Critical submergence level and plant operating level data for Category E plants (90 degree orientation, RHR/HL ID 8.5/29.0 inches)..............................................................
20 TABLES 1.
AIR ENTRAINMENT TEST DATA REGRESSION COEFFICIENTS FOR CRITICAL SUBMERGENCE AS A FUNCTION OF RHRS INTAKE FROUDE NUMBER AND VISCOSITY CR11ER10N VALUES OBTAINED FROM EQUATION (2).........................
8
- vi e
e aw&
1.
INTRODUCTION There have been numerous losses of decay heat removal capabilities at Pressurized Water Reactor (PWR) plants in the United States. On April 10, 1987, the Diablo Canyon Power Plant, Unit 2, lost Residual Heat Removal System (RHRS) capability for approximately 90 minutes. Although thi plant could have withstood the loss of the RHRS for another 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without operator interven-tion, slightly different circumstances could have led to significant core As a result of the Diablo Canyon event, damage and release of radioactivity.
the Nuclear Regulatory Commission (NRC) issued Generic letter 87 12 (GL 87 12)3 to all licensees of PWRs.
GL 87-12 requested each licensee to respond to several important issues, one of which concerns the analytic bases used to develop proccoures for recosery from loss of RHRS capability.
The licensees of plante designed by Westinghouse deferred addressing the GL 8712 question of supporting analyses until they could perform detailed evaluations of operating conditions and recovery strategies. Westinghouse performed the analyses for the Westinghouse Owners Group (WOG), and reported the resylts in WCAP 11916, " Loss of RHRS Cooling while the RCS is Partially filled" Four major topics are addressed in WCAP 11916: (1) plant operations with the Reactor Coolant System (RCS) partially filled, (2) evaluations of air entrainment pottntial at various RCS levels and RHRS flowrates for different Residual Heat Removal (RHR) intake pipe / hot leg orientations and sizes, (3) thermal hydraulic evaluations of postulated losses of RHRS capability for various RCS conditions of operation with the loops partially filled, and (4) application of analyses results to individual plants.
The more significant resuitt of these evaluations are presented in the corresponding sections of this report.
w I
2.
PLANT OPERATION WITH LOOPS PARVIALLY FILLE 0 Section 1 of Westinghouse repott, WCAP 11916, states the purpose of the report and provides RHRS mid loop operations guidance to the WOG partici-pants.
The report also presents (1) the necessary background information for understanding the evaluations presented in WCAP 11916 Sections 2 and 3 (2) a description of a typical, newer + vintage RHRS, (3) the Westinghouse Maintenance Instruction for draining the RCS, and (4) the Technical Specifications associated with RCS loops partially filled (mid loop) operations.
V5tle WCAP 11916 addresses only the newer plant designs, the recovery strategies discussed in WCAP 11916 are oriented toward end results, rather than specific valve alignments and systems configurations.
Consequently, the omission of descriptions related to older system designs does not appear to be a major deficiency in the scope of WCAP-11916.
Westinghouse recommends raising the RCS water level above the top of the hot and cold legs af ter installing nozzle dams.
This should help prevent loss of RHRS caused by air entrainment into the RHRS pump (s), and will give the operators n re time to respond to loss of RHRS capability.
The analyses in NCAP 11916 address RCS refill using the Chemical and Volume Control System (CVCS) charging pump (s).
Westinghouse points out that the CVCS capacity ma.y not be adequate for RCS level recovery in some scenarios. These McNrici, are discussed, and alternate methods of RCS level recovery (e.g., use of a safety injection pump) are adequately described.
Alternative methods of cooling the RCS are well thought out.
Implementa-tion of these recommendations may require some changes in plant Technical Specifications to eMure that the necessary backup systems are available during periods when Mme systems are typically taken out of service.
For example, one Safety injection pump is required to be available during Mode 5 operations, but two pumps may be required to recover from a prolonged loss of RHRS capability.
The " Standard Technical Specifications for Westinghouse Pressurized Water Reactors", NUREG 0452, Draft Revision 5, is the basis for the plant specific technical specifications for the typical, newer RHR system design.
Technical Specification 3/4.4.1.4.2 allows RHRS loops to be removed from operation for up to I hour during Mode 5 operations, if no operations are parformed that would cause dilution of the RC3 boron concentration and core outlet temperature is maintained at least 10F' below saturation temperature. 3This requirement is being addressed through Generic Letter 88 17 (GL 88 17) by the Westinghouse licensees.
Loss of RHRS flow results in loss of RCS temperature indication if the incore thermocouples have been disconnected in preparation for reactor vessel head removal in many plants.
While Westinghouse does not discuss in WCAP-11916 methods of monitoring the RCS temperature following a loss of both RHRS loops, this issue is addressed in GL 88 17.
Technical Specification 3/4.4.1.4.2 also specifies a 12-hour surveillance interval for the RHRS loops.
The Westinghouse analysr indicate that the core can uncover in less than one hour in some situations; however, since this 2
issueisaddressedinGL88It,itsomissionfromWCAP11916isnot significant.
s Section 1 of WCAP 11916 summarizes current operating practices, but does not adequately discuss many of the better procedures used by several of the Westinghouse licensees.
For example, procedures and actions such as those listed below are now being implemented by Westinghouse licensees as a result of GL 88 17.
1.
Replacing Tygon level instrumentation with permanent hardened level instrumentation.
2.
Revising operating procedures to insure backup RHRS availability.
3.
Continuously monitoring the RHRS to provide early indication of air entrainment into the RHRS pumps.
4.
Ensuring that core exit temperature indications are available during all periods of operation.
5.
Maintaining a high RCS water level during steam generator tube bundle draining operations.
3
\\
3.
AIR ENTRAINMLNT EXPERlHENTS The Westinghnuse 0,<ners Group (WOG) initiated a program to study air entrainment into RHR systems.
This program consisted of a series of tests to i
determine the feasibility of specifying operating limits that would prevent j
air entrainment during drained down conditions. A test facility was constructed and experiments were performed to determine the effect of RCS level, RHRS coolant flow, and system geometry on the initiation of air entrainment at the intake of typical RHR systems.
This section contains an evaluation of the Westinghouse findings and recommendations.
Westinghouse reviewed technical reports and journal articles concerning air entrainment phenomena at pipe inta(es, and perfortned dimensional analyses to determine which parameters se important in 11e design of an air entrainment test facility. Westinghouse concludes that viscosity effects and surface tension effects, which are reflected in the Reynolds Number (Re) and the Weber Number (Ws), respectively, could be ignored in the test facility scaling provided the Re is maintained at greater than 50,000 and the Ws is greater than 120.
A criterion developed in a journal article by Odgaard' was used to independently verify the Westinghouse assumption that viscosity effects may be ignored in a froude scaled experiment when the Re is less than 50,000.
The criterion in the Odgaard report indicates that viscosity effects may be ignored when c,Htyt
< l.26 x 10* *
- ml (1) 9 where the units are metric, and v
kinematic viscosity, ml/s H
critical submergence, m 0.15 JRe Re, vertfcal Reynolds number of the vortex g - 9.80665 m/sr Converting to British units, and making the appropriate algebraic substi-tutions ta eliminate the velocity term in Re yM6 W mWW9M W h g
- < 5.0 x 10'88 ft*
(2) 9 v
kinematic viscosity, ftr/s Sc critical submergence, ft ge 32.174 ft/sr Equation (2) was evaluated for each Westinghouse critical submergence correlation in WCAP 11916 using the corresponding upper froude Number (fr) limit as the limiting case and assuming that the test water temperature was 72'f.
As shown in Table 1, the criterion given by Equation (2) is satisfied 4
m I
for all tests. Consequently, the Westinghouse assumption that viscous effects may be ignored is justified.
The Westinghouse assumption that Ws values greater than 120-indicate that i
water surface tension has a negligible effect on vortex formation for the water conditions in the air entrainment tests is also supported by Odgaard and the applicable references in WCAP 11916.
Based on the above assumptions, the test model size was based on the capability of obtaining ratios of inertia force to gravity forces (as repre-sented by Fr) that would be in the same rar t as those in PWRs.
To ensure the l
best representation of all the Westinghouse design PWRs, the-plants were categorized by RHRS intake orientation relative to the horizontal plane, and by j
the ratio of the RHRS-intake pipe inside diameter to the hot leg inside diameter.
This resulted in nine plant categories, A.1, A.2, 8.1, 8.2, 8.3, B.4, C D, and E.
The Category A and B plants were subdivided according to RHRS intake pipe diameters. The correspondence of the nine plant categories to the test configurations is shown in Table 1.
Using information obtained from technical literature and data obtained from their own experiments, Westinghouse concludes that the critical submergence ratio, Sc/d, may be expressed as a function of Fr.
Sc/d = afr (3) b where critical submergence depth above the highest elevation of RHRS nozzle
-Sc
=
inlet RHRS nozzle inlet diameter d
correlation coefficients a,b
=
and froude Number at the nozzle-inlet Fr
=
v
"/( d)
- l g
V = velocity of coolant at the RHRS nozzle inlet The form of the above relationship is supported by both the test data and technical literature.
To verify the Westinghouse correlations, power series regressions of Sc/d
- as a function of-Fr were performed using the data from Figures 2 2 through 2 5 i
of WCAP-11916. The-two sets of regression coefficients are compared in Table 1. -
Westinghouse does not include their correlations for the 1-1/4 inch-
' diameter data at the O' orientation (Test 1.1), the 2 1/4 inch diameter data at 3
the 45' orientation (Test 2.2), and the 2 3/4 inch diameter data at the 90' orientation--(Test 4.3)-in WCAP 11916, although-the data are available (as indicated in Figure 2 2, figure 2-4, and Figure 2-5, respective y.
Instead -
l)
Westinghouse uses the regression coefficients from Tests 1.2, 2.3 and 4.2 for these geometries.
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4 The two sets of correlation coefficients in Table 1 are in agreement for Tests 1.2, 2.1, 3.2, and 4.1.
In Tests 2.3, 3.1, and 4.2, Westinghouse performed the regressions over a partial set of data and then extrapolated over the entire Fr range, even though the test data indicate that there could be limits of fr over which different correlations are applicable.
These limits are also presented in Table 1.
Critical submergence levels relative to the hot leg centerline for the nine categories of plants were calculateu as a function of RHRS intake flow rate using the correlation coefficients listed in Table 1.
The Westinghouse correlations (represented as the WCAP-11916 correlations in the figures), and the results of the power series regressions of the digitized WCAP 11916 data (represented by the EGG EAST-8681 correlations in the figures) are compared in Figures 1 through 12. The plant RHRS mid leop operating level data provided in WCAP 11916 are also included in these figures as a means of illustrating the adequacy of the Westinghouse Sc/d correlations.
(Operating data for the Cate-gory A.2. Category C, and Category D plants are not provided in WCAP 11916).
Operating levels above a Sc curve indicate a lower probability of air entrainment into the RHRS, while operating levels below the Sc curve imply a higher probability of air entrainment.
The level discontinuities in Figures 4 through 10 and figure 12 are the result of a change in the correlation coefficients obtained from the digitized Westinghouse data.
These discontinuities occur in the data presented in WCAP-11916 figures 2-2 through 2 4 as sudden changes in the slope of the Sc/d functions at specific values of Fr.
These Fr limits may be determined by equating the corresponding Sc/d _ regression correlations and then solving for Fr.
The Fr limits are presented in Table 1.
Differences between the Westinghouse correlation coefficients and the coefficients that are obtained by digitizing the Westinghouse data in WCAP 11916 are shown in figures 3 through 10 and Figure 12.
The Westinghouse correlations yield higher critical submergence levels at the higher flow rates. The difference between the WCAP 11916 correlations and the correlations derived in the review of WCAP 11916 is not significant.
As shown in Figure 4 (Category B.1 plants), Figure 6 (Category B.3 plants),
figures 7 and 8 (Category B.4 plants with RHRS inside diameters other than 11.8 inches), and figure 12 (Category E plants), some of the plant operating levels are at or below the Westinghouse Se correlation limits for air entrainment; yet, these plants apparently have not experienced significant air entrainment into their RHRS pumps.
Consequently, it could be concluded from this data that these correlations yield conservative estimates of the level at which air entrainment can be expected to occur, or that some plants may operate their RHRS with some air ingestion.
The plant operating data provided in WCAP 11916 indicate that several licen-sees should consider modifying their procedures if their operating levels fall below the corresponding Westinghouse recommendeo limit. However, in WCAP 11916, Section 2.5, (" Conclusions and Recommendations") Westinghouse stated that the recommended operating limits are not intended to replace operating experience.
This could be interpreted by the licensees to mean that if a plant has operated below the recommended operating level and has not lost RHRS capability due to air entrainment, the plant should ignore the results of the air entrainment tests.
6 l
i In addition to using the test data from the air entrainment test facility, Westinghouse evaluated air entrainment test data from two PWRs, Indian Point Station. Unit 2, (a Category B.4 plant) and Seabrook Nucicar Station, Unit 1, (a Category B.3 plant).
In explaining the differences between the Indian Point Station, Unit 2 tests and the correlation for Category B.4 plants, Westing-house postulates that air n.ay have been trapped in the short horizontal section of the Indian Point Station Unit 2. RHRS piping and was then
- gulped" into the RHRS pump during the next test.
This could be the case if the tests were conducted by decreasing RHRS flow from an initially high RHRS inlet pipe flow rate. Otherwise, based upon the conclusions from the Hitsubishi Heavy Industries tests (discussed in Section 2.4.4 of WCAP 11916), entrained air is carried along with the water when the flow velocity is such that the fr is greater than 1, which occurred in the Indian Point, Unit 2, tests whon the RHRS flow rate was greater than 1680 GPM.
Consequently, if the Indian Point Station, Unit 2, tests were conducted by increasing the RHRS flow rate from an initially low flow to a higher flow, then air gulping may not have been the cause of the dif ference between the Westinghouse correlation and the plant test
- data, in comparing the results of the Westinghoust 5ts with the Seabrook Nuclear Station Unit 1, tests, Westinghouse conciodes that their Sc correla-tion is conservative. However, the Westinghouse correlation underpredicts the critical submergence level in four of the seven Seabrook Huclear Station, Unit 1, tests.
Assuming that the Seabrook Nuclear Station plant would use the same instrumentation for day to day drained down operations as that used in the plant tests, applying the Westinghouse correlation as the aperating limit does not appear to be a prudent guideline, since several of the S6abrook Nuclear Station, Unit 1, plant tests resulted in air entrainment when the level was above the minimum level obtained by applying the Westinghouse correlation for Category B.3 plants given in WCAP 11916.
Nevertheless, the technical basis for the Westinghouse correlations of critical submergence level as a function of fr is sound.
it must be emphasized, however, that (based on PWR plant test data) the Westinghouse correlations may not be
- conservative" correlations, in that they oo not incorporate additional f actors that could prevent air entrainment from the RCS into the RHRS.
This lack of conservatism is not adequately addressed in WCAP 11916. Westinghouse should provide to the Westinghouse licensees some caveat regarding the nonconservative nature of the Westinghouse etr entrainment test data vis a vis its applicability to operating plants.
That is, based on the air entrainment test data from two PWRt (Indian Point Station, Unit 2 and Seabrook Nuclear Station), the Westinghouse correlations should be used only as a starting point for RCS hot leg level, to which some conservative factor could be applied to ensure the prevention of air entrainment during operations with the RCS partially filled.
7 i
1
TABLE 1.
AIR EN1RAINMENT TEST DATA REGRES$10N COEfflCIENTS FOR CRITICAL SUBMERGENCE AS A FUNCTION Of RHRS INTAKE FROUDE NUMBER, AND VISCOSITY CRITER10N VALUES OBTAlHED FROM EQUATION (2).
Plant WCAP 11916 DATA RANGE v8Sc/g Test Cateaory 0
ti/0 a
b a
b to rr Hi Fr fl0ir) 0.56 0.15 1.70 2.85 1.1 0
0.18 0.34 0.61 2.85 5.70 1.70 5.70 0.35 0.31 0.69 1.2 A.),A.2 0
0.32 0.31 0.69 0.31 0.69 1.00 2.80 0.39 2.1 45 0.18 1.60 0.12 1.60 0.12 1.70 5.70 0.67 0.38 1.40 1.00 1.41 2.2 B.1,B.2, 45 0.32 0.50 0.50 1.41 2.80 B.3 1.00 2.80 0.53 0.45 0.64 0.49 1.97 0.60 0.94 2.3 8.4 45 0.39 0.45 0.64 0.94 1.73 0.30 1.70 0.47 0.45 0.64
+
0.81 2.43 0.66 1.08 3.1 C
60 0.32 0.96 0.10 0.96 0.10 1.08 2.80 0.65 3.2 0
60 0.39 0.68 0.17 0.68 0.17 0.88 1.70 0.56 4.1 90 0.18 2.30 0.12 2.30 0.12 2.12 8.80 1.01 1.26 0.68 0.95 1.38 4.2 E
90 0.32 1.50 0.14 1.38 2.80 1.00 2.80 1.06 1.50 0.i4 4.3 90 0.39 1 20 0.14 1.18 0,67 0.60 1.64 1.20 8
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4.
THERMAL HYDRAULIC ANALYSES Westinghouse performed thermal-hydraulic analyses to predict the time to boiling, the RCS pressurization rate, and the-time to core uncovery for a broad range of postulated scenarios of loss of RHRS capability during mid-loop operations.
The thermal-hydraulic analyses were performed using TREAT-NC, an inter-active engineering simulation program based upon a single component two-phase fluid with a noncondensible gas component.
TREAT-NC is adaquate for thermal hydraulic calculations involving subcooled and two ph.1se water conditions with a noncondensible gas compone4.
The thermal-hydraulic analyses were performed using three plant models.
The plants were categorized by their power to heatup volume ratios (P/V) so that the results of the analyses could be adjusted to account for differences in the various plants. This is an appropriate cirterion for this class of analyses.
The P/V for the R. E. Ginna Nuclear Power Plant (2.38) was selected to represent the 2-loop plants.
The average P/V for the 2-loop plants is 2.48 10.12. Although the Ginna P/V is lower than the average P/V, the differ-ence is not significant with respect to predicted plant response. The 3-loop P/V of the modeled Surry Nuclear Station plant is 2.58, the average P/V for the 3-loop plants is 2.74 10.25 (without including the P/V for SONGS 1, which is 1.51); but 8 of the _14 3-loop plants have a P/V greater than 2.80.
A higher P/V for the 3-loop plant model would have -yielded more representative results; but the 3-loop analyses are still valid, given the other conserva-tisms used in the models. The average P/V for the 4-loop plants is 2.68 (without Indian Point Station Units 2 and 3), while the P/V for the modeled Diablo Canyon Power Plant is 2.94; consequently the 4-loop P/V is conserva-tively high.
The assumptions used in developing the plant models are not included in WCAP-11916; therefore, the individual licensees will have to do some developmental work if they choose to repeat the calculations with their own computer models and programs.
Since most licensees already perform their own analyses and have already developed their respective thermal-hydraulic models, this will probably be only a minor inconvenience.
The Westinghouse analyses for loss of RHRS capability during mid-loop operation were divided into six categories to cover various postulated conditions in the RCS during a loss of RHRS capability event.
The analyses categories consist of the following:
1.
Base case heatup and pressurization analyses 2.
Large vent analyses 3.
Steam Generator (SG) nozzle dam sensitivity analyses 4.
V3por vent sensitivity analyses 5.
Cold leg opening sensitivity analyses 6.
SG condensation analyses-21
p i
Additionally, Westinghouse performed a set of recovery analyses to assist licensees in the development or improvement of their plant specific procedures for recovery from a loss of RHRS capability.
The times to core damage were not determined for the analysis categories listed above; rather, the time to core uncovery was used te estimate the time required to mitigate an event.
Since the results of the thermal-hydraulic analyses are to be used by licensees to develop or modify their recovery procedures, specifying the time to core uncovery as the limiting time to mitigate an accident is appropriate.
WCAP-11916 clearly describes how the results of the analyses can be adapted to accomodate different plant designs and calculational techniques for recovering from a loss of RhRS.
The scope of the analyses discussed in WCAP-11916 address recovery of core cooling through the use of coolant systems that could operate under elevated temperatures and pressures. The importance of mitigating the effect of overpressurizing the primary coolant system through the use of injection techniques is addressed in sufficient detail. While other recovery strategies for preventing overpressurization, such as maintaining a large vent in the pressurizer, are not discussed in WCAP-11916, these strategies are discussed in GL-88-17.
Analyses showing the effect of typical vent path openings were performed, and the results are discussed clearly in the report. Westinghouse did not perform a sufficiently broad spectrum of analyses to determine the' vent size that would prevent overpressurizing the RCS following a loss of RHRS capability. A broader spectrum of analyses could have been used to prescribe specific recovery strategies involving RCS venting; specifically, those strategies that would prevent pressurization of the RCS following a loss nf RilRS capability.
-Westinghouse recommends recovery strategies for the different classes of events using the results of the thermal-hydraulic analyses.
With one exception, iiscussed below, these strategies are viable provided the licensee commits to keeping the necessary equipment available during reduced inventory operations.
For certain scenarios, the only acceptab strategies are either to align the Migh Pressure Safety injection System for hot leg injection within approximately 10 minutes after loss of RHRS cooling, or to not allow the RCS to be configured such that these scenarios could occur.
In the thermal-hydraulic analyses the reactor core is modeled as a single comoonent.
Consequently, through numerical diffusion, any fluid entering the volume is uniformly distributed throughout the volume.
This is an adequate assumption when analyzing ti.e effect of coolant additions to a covered core.
In a worst-case scenarion involving an uncovered core, it cannot be assumed that adding water to the hot leg will result in radially uniform fluid distribution of the coolant across the top of the core when it enters the reactor vessel through a single hot leg nozzle. The injected hot leg coolant will cascade out of the hot leg nozzle into the upper plenum, where it will be dispersed primarily by the control rod guide tubes and other upper plenum structures. This dispersion will limit the amount of water actually distributed across the core at the upper core support plate.
Limiting the amount of wa.er distributed across the entire core could result in a longer period of fuel rod uncovery for some portions of the core.
22
r-i i
e s
I For example, a steady 400 GPM of hot leg injection flow into a 29-inch hot leg with a 1-inch deep stream of water on the' bottom of_ the hot leg.will produce a 1-foot wide flow stream moving at about: 24-feet /second when it enters the reactor vessel upper plenum, Without obstructions-(such as-the control-rod guide tubes), this; fluid velocity would carry the water along a-i 1-foot wide axis approximately 6 feet before it reached the upper core support plate. A typical core diameter is 12 feet; consequently, even without i
obstructions in the upper plenum, the half of the core on.the side' of the reactor-vessel farthest from the hot leg. nozzle would not receive any of the injected coolant at the upper core support plate.
l Westinghouse assumes that a 50F'/ min heatup rate is a conservative value i
to use for fuel rod cladding temperature response during a core uncovery, i-This may not apply for-fuel rods with high burnup (which is one of the _
assumptions _used in the Westinghouse analyses)._There is significant fuel relocaction into the pellet-to-cladding gap of highly irradiated fuel rods, which effectively eliminates the gap between the fuel pellets and the inner i
surface of the fuel rod cladding.
This will increase the heat transfer rate from the fuel to-the cladding because the form of heat transfer becomes i
conductive rather than convective, When the core is uncovered, this higher L_
heat transfer rate can significantly increase the fuel rod cladding heatup
~
rate.-
i Westinghouse states _that. severe fuel damage does not occur until-the n
cladding temperature exceeds 1800'F.
Fuel rod behavior tests conducted at the Idaho National Engineering Laboratory have shown that fuel rod failures can
- occur when the fuel rod cladding temperature is in the high a temperature range (-lE00*F).-. Using the Westinghouse 50F'/ min heatup-rate, and an initial fuel-rod temperature of 750'F (as derived from Section 3.7.4 of WCAP-11916),
fuel _ rods uncovered for 15 minutes would heat up to 1500'F, which is in the f
- high-a temperature range. This reduces the available response time for recovering the core coolant ~ inventory as estimated by Westinghouse by approximately 5-minutes.
If hot' leg injection does not prevent fuel rod leakage or. failure, the coolant _ flowing down through the core could ultimately transport. fission l;
products:through an open SG manway into the containment-building. While not
' addressed!in WCAP-11916, the issue of open SG manways providing a pathway for L
contaminated. fluid and steam from the RCS into the containment building-is
[
addressed in GL 88-17.
- The' thermal-hydraulic analyses-address the use of nozzle dams and the best
' sequence of installation to ensure the minimum impact on core level following a loss of RHRS capability. Westinghouse recommends that cold leg steam generator noztle~ dams be_ installed first to preclude rapid core uncovery caused by RCS: pressurization following-a loss of RHRS capability. The.
recommendation for.-this installation sequence is sound and is supported by the results of-the ' analyses. The fact that? Westinghouse addresses this topic is an indication of:the thoroughness of the Westinghouse analysis effort.
The analyses conducted by Westinghouse cover a broad range of postulated scenarios. However, analyses.to determine a vent size to prevent RCS pressurization, and detailed analyses of core reflooding phenomena are recommended.-
23
r 5.
CONCLUSIONS The alternative methods of cooling the RCS following a loss of RHRS capabil-ity, as discussed in WCAP 11916, are well thought out.
Implementation of these recommendations may require some changes in plant lechnical Specifications to-ensure that the necessary backup systems are available during periods when some systems may be out of service, for example, one Safety injection pump is usually required to be available during Mode 5 operations, but two pumps may be required to recover from a loss of RHRS capability in a timely manner.
WCAP-11916 summarizes current operating practices, but does not adequately discuss many of the procedures used by severa' Westinghouse PWR owners.
For example, Westinghouse does not recommend replacing Tygon level instrumentation with permanent, hardened level instrumentation; revising operating procedures to ensure backup RHRS availability; stationing personnel near the RHRS pumps to provide early indication of RHRS pump cavitation; injecting nitrogen into the steam generator plena to ensure that the SG tube bundles drain smoothly; or methods to ensure that core exit temperature indications are availa'le during o
all periods of operation.
Nevertheless, these items are addressed in GL 8S 17; consequently, their omission from WCAP Il916 is not significant.
Westinghouse reviewed air entrainment reports and performed dimensional analyses to determine the important design parameters for an air entrainment test facility.
The body of technical literature supports the Westinghouse assumption that viscosity effects and surface tension effects may be ignored in their air entrainment test facility.
Buoyancy effects on free-surface flow (which are represented by Fr) are the dominant scaling factor in the Westing-house air entrainment tests.
Using the data from their air entrainment tests, Westinghouse developed correlations of critical submergence depth as a function of flow rate for various RHRS design parameters.
The technical basis for the Westinghouse correlations is sound.
It must be emphasized, however, that the Westinghouse correlations may not be " conservative" correlations thkt incorporate additional factors to prevent air entrainment at an RHRS inlet.
This lack of conservatism is not adequately addressed in WCAP-11916.
Based on the test data from two operating plants, the Westinghouse correlations should be used only as the base to which some additional conservative factor can be applied to ensure the-prevention of air entrainment during normal drained-down plant operations.
Westinghouse performed thermal-hydraulic analy:es to assist the olant licensees in developing or modifying their recovery procedures.
ine time to core uncovery was appropriately specified as the limiting time to mitigate a loss-of-RHRS-capability event.
The results of the analyses are clearly presented in such a manner that licensees can readily use the analyses to develop plant operating procedures.
Analyses showing the. effect of typical vent path openings were performed and discussed in the report; however, Westinghouse did not perform a sufficiently broad spectrum of analyses to determine the vent size that would prevent overpressurizing the RCS following a loss of RHRS capability.
24
The thermal-hydraulic analyses address the use of nozzle dams and the best sequence of nozzle dam installation to ensure'that the minimum impact on core level following a loss of RHRS capability. The recommendations for installation sequence are sound and are supported by the results of the analyses.
The TREAT NC thermal-hydraulic model detail was too coarse to conclusively determine that a ccmplete core recovery would occur in a timely manner using one Safety Injection pump with hot leg injection.
It is postulated that, if fuel rod leakage or failures occur, hot-leg-injected coolant flowing down through the core may transport fission products from leaking or damaged ruel rods out an open steam generator manway into the containment.
Because there may bc some potential for release of radioactivity into the containment, additional analyses using a more detailed thermal-hydraulic model may be required to evaluate core recovery for this event.
Overall, WCAP-11916 is very thorough and well written, and provides the licensees with valuable information regarding prevention o and recovery from r
-losses of RHRS capability.
y-25
r REFERENCES 1.
U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Generic letter 87-12. " Loss of Residual Heat Removal (RHR) while the Reacter Coolant System (RCS) is Partially Filled," July,1987.
2.
Westinghouse Electric. Corporation, Power Systems Division, WCAP 11916
" Loss of RHRS Cooling while the RCS is Partially filled," Rev. O, July 1988.
3.
U.S. Nuclear regulatory Commission, Office of Nuclear Reactor Regulation, Generic letter 88-17, " Loss of Decay Heat Removal," October, 1988.
4, A. J._ Odgaard, " Fluid Properties and their Scale Effects in Froude-Scaled Hydraulic Models", IAHR Synoosium on Scale Effects in Modellina Hydraulic Structures, 2": Edition, Edited by H. Kobus, Stuttgart University, Stuttgart, Federal Republic of Germany,1985.
t 26
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BIBUOGRAPHIC DATA SHEET
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Technical Evaluation Re:: ort, WCAP-1196 " Loss of RHRS 3.
mars.ne.r.vsus-as C:oling 'While the RCS is Partially Filled" i
- r-
...a September 1959
- 4. 8 s% C A G A A>si Nund s t e D5039
- 5. AviaC Aiss
(, 7Y P E 0 8 A t PO R T TER Michael E. 'iateman
.v taiccc:,sas:n..
o....,
May, 1989 to September, 1989
- s., s s : n w u z.: n a,w:a. 8
~ u s ~; a:: n n ss n, ~.e.,- o
. on a~1 ~~.
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k EGGG Idaho, Inc.
P.O. Ecx 1625 Idaho Falls, IG S3415
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..c o
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.u U. S. Nuclear Regulatory Cer=ission Washing n, D.C.
20555 lo. $VPP'.,lMENT AaY NOTES u.AasimA:T isoo A techelf41 evaluation of ben '*ghouse recort VCAP-11916.
- Loss of RHR5 Cooling =ntle the RO$
is Part tally Filled." was perfomed.
The plant operations description section did not include a discussion of same of the better techniques er cloyed by licensees to prevent lossas of RHR$ capability when the RCS 1s deatned ccwn.
The air entratrr,ent ex eriments yielded correlations of the level at which str entrainme' teto a RHR line could be espected as a function of RHR flow rate and orientation. These cort - t..n s should be used as the bases to which s:rie additional f actor can be added for specify!q,,he int.mrn operating le,el.
The thermal-bydraulic analyses covered a broad range of postulated losses of RHR$ cooling scenarios.
Methocciogies for acolytng the results of the thermal bycraulic analyses to individual plants are well-concetved. and clearly cescribed.
i.
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na s in a.s a==. s.amenve=.e a ans.eas sa. s
,..s Unlimited Loss of DHR cooling
,,,c,,,,c,,,,.,
Mid-loop operation Tr,.
Unclassified Non-power operation Nuclear power plant Unclassified Partially filled RCS u.a.v,.aaan*,4 cts Reduced Inventory Operation is. PRscs