ML20031D018
| ML20031D018 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 10/01/1981 |
| From: | Davidson D CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8110090214 | |
| Download: ML20031D018 (41) | |
Text
' THE CI EVEl.AND ELECTRIC ILLUMIN ATING COMPANY P o Box 5000 m CLEVELAND. oHlo 44101 e TELEPHONE (216) 622-9800 e id UMi*' MING BLDG e 55 PUBLIC SOUARE Serving The Best Location in the Nation Dalwyn R. Davidson tflCF PRESIDENT SYSTEM ENGINEERING AND CONSTRUCYlON
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b october 1, 1981 (T
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k, Mr. Robert L. Tedesco Assistant Director for Licensing Division of Licensing W
U. S. Nuclear Regulatory Commission
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Washington, D. C.
20555 Perry Nuclear Power Plant Docket Nos. 50-440; 50-441 Response to Request for Additional Information -
Core Thermal Hydraulics
Dear Mr. Tedesco:
This letter and its attachment is submitted to provide draft responses to the concerns identified in your letter dated August 12, 1981, in regard to care thermal hydraulics.
It is our intention to incorporate ese responses in a subsequent amendment to our Final Safety Au.
3is Report.
Very truly yours,
/
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Dalwy'. Davidson Vice President System Engineering and Construction DRD: dip Attachment ec:
M. D. Houston G. Charnoff NRC Resident Inspector go !
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- 492.1
)j,Referencel'-(Section4.4)-documentNEDO-10958Adoesnotin-
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(4.4.2.2.1)' >t. clude -the 'BWR/6 Atlas Test data and other factors that just-.
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'ify.the GEXL correlation application to Perry.
Please pro--
vide the correct references.
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i Respons'e
~ Reference 1 has been modified in Amendment 3.
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~492.2 How'was the axial power distribution given in 'r ble 4.4-5 a
(4.4.2.4) selected as the basis to generate void and. quality distri-
-s butions for Perry? Was any sensitivity study performed to select this axial power distribution for use in the analysis?
Response
The axial power distributien of Table 4.4-5 was selected because it results in the highest expected core average and exit void fraction and qualfcy, given that the power shape is similar to that expected during normal operating conditions. This axial power ~ distribution sensitivity has been determined over the course of the total BWR design history.
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What are the values of the following for Perry?
(Table 4.4-1) a) Design basis maximum core support plate pressure drop
-(normal + upset);-
b)' Design basis maximum allowable channel wall pressure drop.
Response
The design basis for BWR fuel channels and vessel internal structures (including the core support plate) with respect to pressure differentials is that the component in question maintain structural integrity when subjected to certain load combinations. A pressure difference acting on a given component is merely considered as one of the loads in those combina tions. As such, it is not necessarily meaningful to specify a design basis maximum pressure drop without also specifying the other associated loads.
It can be stated, however, that pressure differences used in those calculations conservatively bound the calculated pressure loadings of Table 4.4-1.
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492.4 Have the core plate pressure drop measurements ever been (4.4.2.6) done for the operating BWRs with 8x8R or P8x8R fuel with two water rods?
If not, do you intend to do these measurements?
Response
No data exist which compare calculated with measured core plate pressure drop for operating plants with only two water rod fuels. As the methods of analyzing those cores with 8x8R or P8x8R fuel are no different than
'those for 7x7 or 8x8 fuels, it is anticipated that any such comparison I
would yield results nearly identical to-those given in Section 4.4.2.6 and in Table 4.4-7.
Furthermore, the plant operator has the capability, via the plant process computer, to monitor the core plate pressure drop at any time during operation, thus providing the means for such a comparison if it is so desired.
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492.5 Identify the. pressure drop data and the models used for
- (.4. 4. 2. 6) -
Perry by appropriate' references and provide perspective on the latest significant model changes and the previous plant design applications for the latest model.
Response
. Pressure drop data a'nd models used in the Perry design are as indicated in Section 4.4.2.6 and References 3 and 4 of Section 4.4.7.
Analyses for all BWR/6 plants, including Perry, are based on the same pressure data and r-models; no unique changes have been made for the case of Perry.
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492.6 The following equations in the Perry FSAR are in error and (4.4.2.6.2) should be corrected as described below:
2 W
K 2
0 L" 2gpf A
TPL 2
2 The term "A " should be "A n 2
f (l ~ "}
+P g AL 11 APE" O
g This equation should be as:
APE" NE ( ~ "} + E8" E
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=
Pm aa (1 - a) pg g
The term "Pm" should be "p,"
Response
The response to this question is provided in revised Section 4.4.2.
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. Local Pressure Drop The' local pressure drop is. defined as the irreversible pressure loss associated with an area changeisuch as the orifice, lower tie plates, and spacers of a fuel assembly.
The general-local pressure drop model is similar to the friction pressure drop and is:
2 w
K 2
AP =
2 TPL g
2gpf A
n as where:
= Local Pressure Drop, psi L
K
= Local Pressure Drop Loss Coefficient A
= Reference Area for Local Loss Coefficient, and 2
$TPL = Two-Phase Local Multiplier p
= Density c_
O and w, g, and p are defined the same as for friction.
This basic model is similar to that used throughout the nuclear power industry.
The formulation for the two phase multiplier is similar to that reported in the open literature (Reference 4) with the addition of empirical constants to adjust the results to fit data taken by General Electric Company for the specific designs of the BWR fuel assembly. Tests are performed in single phase water to calibrate the orifice in the lower tie plate, and ir. both single-and two phase flow to arrive at best-fit design values for spacer and upper tie plate pressure drop. The range of test variables is specified to include the range of interest to boiling water reactors.
New data are taken whenever there is a significant design change to ensure the most applicable methods are in use at all times.
4.4-10
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'4.4.2.6.3-Elevation Pressure Drop The' elevation pressure drop is based on the well known relation:
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- P E
=[pg (1-a) + p a] GAL d
g e
- where:
= Elevation Pressure Drop, psi E
= Incremental Length
.p
= Average Water Density a
- = Average Void Fraction Over the Length L p,p = Saturated Water and Vapor Density, respectively
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= Acceleration of Gravity 4.4.2.6.4 Acceleration Pressure Drop A reversible pressure change occurs when an area change is encountered, and an irreversible loss occurs when the fl5id is accelerated through the boiling process. The basic formulation for the reversible pressure change resulting from a flow area change is given by:
2 A
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ACC 2'
2gpA A
2 7
where:
= Acceleration pressure drop, ACC A
= Final flow area, 2
A
= Initial flow area 1
and other terms are as previously defined. The basic formulation for the acceleration pressure change due to density change is:
4.4-11
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AP ACC.
2 gA E
P ch M out M in where:
1 x
(1-x) e
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pa M Mg d
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.pg = momentum density x = steam quality and other terms are as previously defined. The total acceleration pressure drop in boiling' water reactors is on the order of a few percent of the total pressure drop.
4.4.2.7 Correlation and Physical Data General Electric has obtained substantial amounts of physical data in support of the pressure drop and thermal hydraulic loads discussed in Section 4.4.2.6.
Correlations have been developed to' lit these data to the formulations discussed.
4.4.2.7.1 Pressure Drop Correlations General Electric Company has taken significant amounts of friction prersure drop data in multirod geometries representative of modern BWR plant fuel bundles and correlated both the friction factor and two phase multipliers on a best fit basis using the pressure drop formulations reported in Sections 4.4.2.6.1 and 4.4.2.6.2.
Checks against more recent data are being made on a continuous basis to ensure the best models are used over the full range of interest to boiling water reactors.
Tests are performed in single phase water to calibrate the orifice and the lower tie plate, and in both single-and two phase flow to arrive at best fit design values for spacer and upper tie plate pressure drop.
The range of test 4.4-12 i
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492.7.
--The power / flow operating map of Figure 4.4-2 in.the Perry
=(4.4;3.3),
,FSAR is incomplete.
. The following are not provided for Perry (Figure 4.4-2)
FSAR.
i APRM scram line 11 APRM rod block line 111 105 percent rod line iv 70 percent rod line Where and when will this information b2 availabic? If this infctmation will not be included in Perry 78AR, provide j ustification.
Response
A revised Figure 4.4-2 is attached. The revised figure includes the required information.
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492.8 What is the value of. calculated nominal bypass flow fraction (4. 4.' 4. 5. 2 )
for Perry.
Reference any measurement data used to confirm the calculated bypass flow and what is its uncertainty?.
Discuss and justify assumptions used, such as axial power distribution, friction loss coefficients, etc., in the
-calculation of the bypass fle'.
Response
.The expected nominal bypass flow fraction for Perry is 11.0% with a 10 uncertainty of 2.5%; the actual bypaas flow fraction at any time during operation will depend on core conditions including axial power distribution, time during fuel lifetime, etc.
Significant data concerning measurement of bypass flow fraction is given in Reference 1 (Reference 10 of Section 4.4.7).
Reference:
1.
" Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibration," NEDE-21156, Class III, January 1976.
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492.9
.What fraction of the fuel bundle flow is " water rod flow"?
1 (4.'4. 4. 5. 2 )
Did'you verify your. calculations with previous measurement data?
Response
The nominal water-rod flow fraction for Perry at rated conditions is 1.35%
of the total core flow; models used in the calculation of this fraction were derived from. experimental data.
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492.10 What is the value of-the fraction'of total reactor power de-(4.' 4. 4. 5. 3)
~. posited in the bypass' region for Perry?--
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Response
The cominal fraction of total reactor power deposited in the bypass region ' for -Perry is 2.0%.
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^492 11 You have not cited the name, version, or reference of the (4. 4. 4. 5') _
computer program used in this sub-section.
. Letter from N. W. Curtis (Pennsylvania Power and Light Company) to B. J. Youngblood (NRC), " Response to NRC questions on Susquehanna FSAR," dated March 25, 1981, states that name of the computer program is "ISC R" and reference is " General Electric Document NEDO-20953, May 1976, Chapter 4."
Please confirm ISC$R has been used for Perry.
What version number of ISCOR is the latest version? Has this version been applied to Perry?
If the reference of this version is different from GE Document NED0-20953, provide the document or the reference. Also describe any e
significant changes of this version of ISC0R code'over the previous version of ISC R.
/
Response
.The computer program cited in Section 4.4.4.5 is named ISC0R. The ISC0R computer program and another GE program PANACEA (3 dimentional BWR core
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simulator) use the same steady state thermal hydraulic mathematical module described in NED0-20954 dated January 1977. The program ISCOR and the calculations used for Perry are consistent with the technical content of NEDO-20954 dated January 1977.
The details of ISC$R and its associated proprietary documentation are available for review at GE in San Jose.
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492.12 Section 4.4 contains no discussion of crud and its effect
-(4.4) on CPR :nd core pressure drop.
Provide the assumptions u,ed for amount of crud in design calculations and the sensitivity of CPR and core pressure drop to variations in the amount of crud present. Also reference or provide data supporting the assumption on crud thickness and dis-l cuss how crud build-up in the core would be detected Provide a descriptics of the instrumentation available and the surveillance requirements and procedures available which would alert the reactor operator to an abnormal core flow or core pressure drop during steady state operation.
Also describe any corrective action which would be taken.
Response
In general, the CPR is not affected as crud accumulates on fuel rods (References 1 and 2).
Therefore, no modifications to CEXL are made to account for crud deposition.
For pressure drop considerations, the amount of. crud assumed to be deposited on the fuel rods and fuel rod spacers is greater than is actually expected at any point in the fuel lifetime.
This is reflected in a decreased flow area, increased friction factors, and increased spacer loss coefficients.
The effect of this crud deposition is to increase the core pressure drop by approximately 1.7 i.i; this increase could be detected in monitoring the core plate pressure drop.
There are no surveillance requirements or procedures with respect to abnormal core flow or pressure drop; an instance of abnormal crud build-up would first be detected in the monitorlag of reactor water chemistry.
Further discussion of crud (service-induced variation) and its uncertainty is. contained in Section III of Reference 3 (Reference 1 in Section 4.4.7).
v q-Pf7erences:
1.
- McBeth, R.V., R. Trenberth, and R. W. Wood. "An Invt.itigation Into the Ef fects of Crud Deposits on Surface Temperature, Dry-Out, and Pressure Drop, with Forced Convection Boiling of Water at 69 Bar 3
in an Annular Test Section", AEEW-R-705,1971.
"Th r al and Hydraulic 2.
- Creen, S.J.,
B.W. LeTourneau, A.C. Peterson, em
- ffects of Crud Deposited on Electrically Heated Rod Bundles",
WAPD-Di-918, Sept. 1970.
3.
" General Electric Thermal Analysis Basis (':ETAR): Data, Correlation, and Design Application", General Electric Coepany, January 1977, (NEDO-10958A).
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.r 492.13 Do you have any docketed reference for the analysis of fuel (4.4.2.9) cladding integrity safety limit value of 1.06 for the first core and 1.07 for the reload core?
Response
There is no docketed reference for the analysis of the fuel cladding integrity 4
safaty limit values for Perry. These values result from the statistical rod boiling transition analysis (Section 4.4.2.9 and Questions 492.01 and 492.14).
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492.14 Have you done any bounding BWR statistical analysis for (4.4.2.9)
BWR/6 fuel? Your Reference 1 does not mention this analysis done for BWR/6 fuel.
Please provide a correct reference for this.
Response
Reference 1 has been modified in Amendment 3.
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r 492.15 Relative bundle power histograms should be similar for all (4. 4. 2. 9) the BWR/6 plants.
If this is true, why is Perry histogram different than Grand Gulf.
Response
The relative bundle power distributions used in the statistical analyses for Perry and Grand Gulf are identical.
Histograms shown in Grand Gulf FSAR (Figures 4.4-10 and 4.4-11) are CPR histograms, as opposed to Figure 4.4-1 of the Perry FSAR, which is a relative bundle power histogram.
These histograms can be found in NEDO 10958.
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492.16 You have not cited the name, version, and reference of the (4.4) co > wide transient analysis code (i.e., ODYN or REDY) and
~
for the GETAB-MCPR evaluation of the transients.
Please provide name, version, and reference of these two codes used for Perry.
Response
The REDY code, as documented in NEDO-10802, " Analytical Methods of Plant Transient Evaluations for the General Electric Eoiling Water Reactor," was used for the core wide transient analysis as shown in Chapter 15.
Limiting pressurization events evaluated with the ODYN cede will be provided in the near future. All the GETAB-MCPR evaluation of the transients was performed with the SCAT code as documented in NEDO-20566, " General Electric Coi pany Analytical Model for Loss-of-Coolant Analysis in accordance with 10CFR50, Appendix K."
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492.17-Provide the name, version, and reference description of the (4.4.4.6.4) following models cited in this sub-section:
i the hydrodynamics model; ii.
the core model; and 111 the plant model.
. Cite our previous approvals of these models.
Response
The hydrodynamics model and the core model are described in References 23 through 28 of Section 4.4.
The plant model is described in NEDO 24154 and NEDE 24154-P on the computer program ODYN, which was accepted by NRC.
See R. Tedesco's letter to G. G. Sherwood dated February 4, 1981.
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492.18 In the discussion of the models used in the stability anal-
' (4. 4. 4. 6).
yses, you state that "As new experimental or reactor oper-ating data are cbtained.the model is refined-to improve its capability and accuracy." Ara the comparisons of the models with the data, as given in Figure 4.4-6, based on the same versions of the models as were used for Perry?
If not, provide comparisons using Perry models.
In addition, pro-vide a description of the models and the references.
Response
The comparisons of the models with data as'given in Figure 4.4-6 are based on the same version of the model as that used for Perry.
The stability licensing topical report, NEDO-21506, provides a description of the analytical methods used in the code as well as model qualification through comparison with test data.
Reference:
Licensing Topical Report, " Stability and Dynamic Performance of the General Electric Boiling Water Reactor," January 1977 (NEDO-21506).
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U,?2.19-The-staff is performing a generic study of the hydrodynamic
- (4.4.4.6).
-- stability _ characteristics of LWRs under normal operation, 7enticipated transients, and accident conditions.
The results of this stt'dy will be applied to the staff review and acceptance of stability analyses and analytical methods now in use by the reactor vendors.
In the interim, the staff concludes that past operating experience, stability tests, and the inherent thermal-hydraulic characteristics of LWRs provide a basis for accepting the Perry stability evaluation for normal operation and anticipated transient events.- However, in order to provide additional margin to stability limits, natural circulation operation of. Perry will be prohibited until the staff review of these condi-tions is complete.
Any action resulting from the staff study will be applied to Perry.
Response
It is believed that the stability characteristics for Perry meet the ultimate stability limit even with the natural circulation operation.
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v5 492.20 The reactor core decay ratio for natural circulation, (4.4.4.6.1)'
105 percent rod pattern is given as 0.97.
Operation in a region of the-power-flow map with such a high decay ratio may not be permitted.
Discuss the uncertainty in the calculation of the decay ratio and discuss possible means of preventing operation in that region of the powar flow map, e.g., adjustment of rod block limits and APRM power-flow scram set points to preclude operation in that region of the power-fics map.
Response
The region of high calculatei decay ratio is generally in the area of natural circulation, which is a region in which BWRs do not normally operate.
The decay ratio of 0.97 was calculated using a bounding void coefficient which is considered characteristic of equilibriun cycle conditions (see Figure 4.3-23) and is therefore expected to cover several cycles of operation.
Although the decay ratio for the first cycle of operation will be less than 0.97, the 0.97 value is considered to be an acceptable value since it is well below the true safety limit (see attached Reference 1).
The acceptability of high decay ratios for BWRs has been demonstrated on licensed operating BWRs, many of which have decay ratios in the range of 0.8 to 1.0.
In addition, the model which is currently under generic review has been shown to be conservative relative to in-plant tests (i.e.,
Peach Bottom Cycles 2 and 3).
The most practical means of excluding planned operation in a certain region of the power flow map is to apply administrative controls, e.g.,
via technical specifications rather than adjustment of the rod block limit and the APRM power-flow scram lines. Modification of the rod block and APRM lines would affect operation over the entire region of expected power flow operation and would ultimately have a negative affect on plant availability and capacity factor.
Reference:
1.
Letter, R. Engel (CE) to D. Eisenhut (USNRC), " Boiling Water Reactor Stability Margins," April 4, 1977.
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i 492.21 Has the Vermont Yankee stability test data been compared (4.4.4.6)-
to the values predicted using the Perry stability model?
If so, when will this evaluation be available?
Response
Analysis of the Vermont Yankee test data is in process at this time. After the final test results are available, they will be compared to predictions made with the CE stability model which was used on Perry. The significance of the results will be discussed with the NRC and plant owners as soon as possible af ter completion of the work.
It is estimated that this will occur in the first half of 1982.
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~492.22 No analysis has been presented for MCPR limits or stability (4.4) characteristics for one loop operation.
One loop operation will not be permitted until supporting analyses are provided and are approved by the staff.
Response
Stability' analysis and transient analysis for operating CPR limits presented in the FSAR are for two loop operation.
Natural circulation conditions are the same in either case.
If the utility decides to operate the plant with one recirculation loop, additional support analyses will be provided and submitted to the NRC for approval.
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492.23 It is stated in Section 4.4.4.1 that the GEXL correlation (4. 4. 4. 5. 3) -
from Reference I was used for critical power ratio calculations.
Why does Section 4.4.4.5.3 reference a different correlation?
Response.
The response to this. question is provided in revised Section 4.4.4.5.
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The core power is divided into twn parts:
an active coolent power and a
. bypass flow power. The bypass flow is heated by neutron-slouing down and gamma heating in the water, and by heat transfer through the channel walls.
-Heat is also transferred to the bypass flow from structures and control elements which are themselves heated by gamma absorption and by q, a reaction in the control material. The fraction of total reactor power deposited in the bypass region is very nearly 2 percent.
A similar phenomenon occurs, with the
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fuel. bundle, to the active coolant and the water rod flows. The net effect is that 96 percent of the core power is conducted through the fuel cladding and appears as heat flux.
In design analyses, the power is allocated to the individual fuel bundles using a relative power factor. The power distribution along the length of the fuel bundle is specified with axial power factors which distribute the bundle's power among the 24 axial nodes.
A nodal local peaking factor is used to establish the peak heat flux at each nodal location.
The relative (radial) and axial power distributions when used with the bundle flow determine the axial coolant property distribution resulting in sufficient information to calculate the pressure drop ccmponents within each fuel assembly type.
Once the equal pressure drop criterion has been satisfied, the critical bundle power (the power which would result in critical quality existing at some point in the bundle using the correlation expressed in Reference 1) is determined by an iterative process for each fuel type.
In applying the above methods to core design, the number of bundles (for a specified core thermal power) and bundle geometry (8x8, rod diameter, etc.)
are selected based on power density and Linear Heat Generation Rate limits.
4.4.4.6 Thermal-Hydraulic Stability Analysis 4.4.4.6.1 Introduction There are many definitions of stability, but for feedback processes and system is stable if, control systems it can be defined as follows:
a following a disturbance, the transient settles to a steady, noncyclic etate.
4.4-24
492.24 How do you adjust operating limit MCPR values for the operation (4.4)-
at lower than 100 percent power and 100 percent flow conditions?
Response
h The operating limit minimum critical power ratio (MCPR) at off-rated operating l
states is determined from the MCPRg and MCPRp curves which are functions of core flow and power,<respectively.
These curves and the associated bases for these curves are part of the plant technical specifications; as such, further information regarding the off-rated operating limit MCPR will be Provided at the time of Perry tech specs submittal.
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492.25 Provide a detail discussion of the operator training program (4.4.6)-
for operation of the Loose Parts Monitoring System (LPMS),
planned operating procedures and record keeping procedures according to Regulatory Guide 1.133.
Res ponse The operator training program for the Loose Parts Monitoring System will be consistent with the level and quality of the training established for licensed operators at the Perry Plant.
The training will be a combination of formal classroom training, on-the-job training, and self-study. Oral and written exams that arc. periodically administered to ensure the adequacy of this training will include the Loose Parts Monitoring System.
The operator training program for the Loose Parts Monitoring System will include, but not be limited to, training in the following subject areas:
1.
Use of Perry Plant system operat,ing instructions to startup, operate, and shutdown the system including subsystems such as the tape recorder, loose part locater, spectrum analyzer, and spectrum analyzer x-y plotter.
2.
The basic theory of operation and purpose of the system.
3.
Operator actions and engineering review for the use of supplemental plant data to confirm the presence or possibility of a loose part and for determining the short and long-term safety implications of the loose part.
4.
Requirements for _;tification of the Nuclear Regulatory Commission which apply to the system.
5.
Technical Specifications which apply to the system.
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m The planned operating procedures for the Loose Parts Monitoring System will consist of:
1.
System Operating Instruction (s) 2.
Alarm Response Instruction (s) 3.
Periodic Test Instruction (s) 4.
Technical Specification Surveillance Instruction (s)
These instructions will be written in accordance with Perry Plant administrative procedures and will consider minimizing personnel time in high radiation areas to limit occupational exposure.
Together they will include, but not be limited-to, instructions concerning the following subjects:
1.
Startup, operation, and shutdown of the system and subsystems such as the tape recorder, loose part locator, spectrum analyzer, and spectrum analyzer x-y plotter.
2.
The use of supplemental plant data to confirm the presence or possibility of a loose part and for determining the short and long-term safety implications of the loose.part.
3.
Operator actions and required engineering review when the presence or possibility of a loose part is confirmed.
4.
Methods for determining the alert level and for identifying and making allowances for alert signals caused by plant maneuvers.
5.
Methods for performing the periodic tests, calibrations, and checks necessary to determine the presence or t.bsence of a loose part and for verifying proper system operation.
The record keeping procedures for the Loose Parts Monitoring System will be contained in an administrative instruction titled " Loose Parts Monitoring Program". Currently, this instruction is designated PAP-0210 and will be written in accordance with Perry Plant administrative procedures.
This instruction will contain, but not be limited to, instructions concerning the following subjects:
1.
Pr7paration of reports to the Nuclear Regulatory Commission when defining or permanently changing the alert level, when a loose part is confirmed to be present, or when the associated technical specification is violated; including followup summary reports.
2.
Retention of prior operating history data necessary for determining the safety.significa ce of a loose part.
3.
Retention of data acquired by the Loose Parts Monitoring system including loose parts locator printouts, magnetic tape recordings, and spectrum analyzer x-y plots.
4.
Retention of data documenting each alert including time and plant condition.
Retention of data obtained during technical specification survNillances 5.
and periodic tests of the system.
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492.26 Provide an evaluation of your LPMS for compliance with (4.4. 6)
Regulatory Guide 1.133.. Justify any deviations.
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Response
The response td this question is provided in revised Section 4.4.6.1.
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l 4.4.6 INSTRUMENTATION REQUIREMENTS The reactor vessel instrumentation monitors the key reactor vessel operating parameters during planned operations. This ensures sufficient control of the parameters. The following reactor vessel sensors are discussed in Sections 7.7.1 and 7.6.1.
1.
ReacLor Vessel Temperature 2.
Reactar Vessel Coolant Flow Rates and Differential Pressures 4.
1:eactor Vessel Internal Pressure 5.
Neutron Monitoring System 4.4.6.1 Loose Parts Monitoring The loose parts monitoring (LPM) program automatically detects, conveys y
wudible and visual signals to the operator and records vibration signals for 5
signature analysis when these signals differ from stored signals of normal ope ra tion.
It also includes administrative procedures for operation and reporting loose parts.
se 4.4.6.1.1 Power Generation Design Bases The LPM program is designed to provide early detection of loose parts in the primary system to avoid or mitigate safety-related damage to or malfunctions of primary system components.
4.4.6.1.2 Program Description Twelve sensing channels (6 channel pairs) are provided to detect a loose part e
that weighs f rom 0.25-30 lbs and impacts with kinetic energy of 0.5 f t-lb
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within 3 ft of each sensor.
A spring-loaded starrette punch standard is used to calibrate each channel.
Sensors are strategically located vith two sensors at each natural collection region.
The specific location of accelerometers has been determined based on manufacturer's recommendations and Regulatory Guide 1.133.
(See Figure 4.4-19) 4.4-32
s Channel pairs are separated.
Separate conduit runs to each sensor.
Channel pairs are then routed separately to one of~two drywell penetrations and 4,
terminate at the charg ' amplifiers in a steel cabinet near the containment a<
shell. This area of containment is accessible during. power operation.
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4.4-32a l'
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- The remainder of the equipment is in the miscellaneous electrical panel room e"
in the intermediate building. All cabling, pull boxes and cabinets are m
. seismically mounted. The electronics equipment cabinet is seismically
. designed.
The alarm setting for each sensor is determined after the system is installed, and will be sufficiently above the normal background noises to minimize spurious alarms yet low enough to meet the sensitivity requirements of the system. A disable signal is provided during control rod movement. The
. manufacturer is responsible for providing a discussion of anticipated major sources of internal and external noises along with procedures to minimize their effects.
When a signal that exceeds 0.5 ft-lbs is detected, the first-on channel alerts the control room and goes on audio and CRT. The magnetic tape records the first-on and the three matrixed channels associated with it.
In the manual mode, any of the twelve channels can be listened to, displayed on the spectrum analyzer and/or taped.
10 Sensors and hardline cable (5 feet)'are designed for 650' F, 100% RH, 10 rads, 70 psig and 100 G's.
Preamplifiers and softline cable are specified at 150* F, 100% RH, 10 rads, 70 psig and 10 G's.
The electronics equipment cabinet is seismically designed.
Solid state electronics and plug-in design are used throughout. These characteristics minimize repair time and g
consequent occupational radiation exposure.
In the automatic mode, when 0.5 ft-lb is exceeded, logic alerts the control room; the 4-channel magnetic tape recorder is started and audio and visual signals appear.
Pre-wired logic selects three additional channels for taping along with the tripped channel.
Recording continues for 6 minutes.
A microprccessor - controlled loose part locator is also activated in the auto mode.
It records sensors involved, order of arrival of signals, time lag
- between successive arrivals, calculated distance from source, peak energy and time code.
4.4-33
If an~ alert level is exceeded or weekly audio or quarterly measur'sents indicate presence of a loose part, diagnostic steps shall be taken in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to determine safety significance.
A technical specification for the loose part detection system shall be provided.
It will include location of the sensors, a limiting condition for operation that the system be operable during start-up and power operation and applicable surveillance requirements.
4.4.6.1.3 Safety Evaluation The LPM system is intended to be used for information, purposes only by the plant operator. The operator does not rely en the information provided by the LPM for the performance of any safety-related action.
Although the LPM is not classified as a safety-related system, it is designed to meet the seismic and l
environmental operability recommendations of Regulatory Guide 1.133 and requirements of IEEE Standard 344-1975, " Seismic Qualification cf Class 1E Equipment." In addition, the manufacturer ensures that the equipment will l
withstand, without loss of function, the normal vibrations expected.
4.4.6.1.4 Test and Inspection In the manual mode, on-line checks are made of each channel. Two channels of solenoid-operated pingers are provided for functional tests.
A calibrated starrette punch is used for of f-line calibration.
(
a Performance checks, functional tests and channel calibrations will be made in accordance with technical specifications.
The manufacturer will provide services of qualified personnel to provide technical guidance for installation, startup, and acceptance testing of the system.
In addition, the manuf:ctucer will provide the necessary training of plant personnel for proper system operation and maintenance and planned operating and record-keeping procedures.
4.4-13a
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- 4.4.6.1.5 Instrumentation Application The LPM system consists of accelerometer sensors, a central monitoring cabinet, a display panel, alarm lights both locally and in the main control room,-automatic turn-on in case of a signal that exceeds predetermined audio e
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4.4-33b t
O 492.27 We require that the LPMS be operational before the fuel loading.
(4.4.6)
Response
1 The LPMS will be operational before the fuel loading.
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o 492.28 The applicant takes the BWR Owner's Group position that no additional instrumentation is needed.
However, the Regulatory (II.F.2)
Guide 1.97 requires that the incore thermocouples be installed.
Therefore, we require that the applicant commit to install incore there> couples in accordance with Regulatory Guide 1.97 and to provide the documentation required by NUREG-0737 Section II.F.2 for staff review.
Response
The Cleveland Electric Illuminating Company supports the BWR Ow:7er's Group position that no additional instrumentation is needed to monitor inadequate core cooling at the Perry Nuclear Power Plant.
Further, we feel that incore thermocouples may provide the operator with ambiguous information in the event of an accident.
We are involved in funding and active participation in the BWR Owner's Group subcommittee on Regulatory Guide 1.97.
As part of this effort, alternative means of detection of inadequate core cooling which meet the intent of Regulatory Guide 1.97 are being inves,tigated.
Additionally, we are members of the Licensing Review Group II (LRG-II) which is pursuing a common resolution to this issue.