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- Docket Nos. 50-424/425 MEMORANDUM FOR:
Elinor Adensam, Chief, Licensing Branch No. 4 Division of Licensing FROM:
Olan D. Parr, Chief, Auxiliary Systems Branch, Division of Systems Integration
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - V0GTLE, UNITS 1 AND 2 The enclosed request for additional infonnation for Vogtle, Units 1 and 2 is a result of recent ACRS concerns involving the handling of heavy loads (SRP Section 9.1.5).
We request that the applicant respond to this request for information in a time frame consistent with the present SER schedule.
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. Parr, Chief L
Auxiliary Systems Branch Division of Systems Integration
Enclosure:
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As Stated cc w/ enclosure:
L. Rubenstein T. Novak M. Miller J. Wilson W. LeFave
Contact:
W. LeFave X29470 4
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o REQUEST FOR ADDITIONAL INFORMATION V0GTLE, UNITS 1 AND 2 AUXILIARY SYSTEMS BRANCH 410.70 (Section9.1.5) As a result of recently identified ACRS concerns, provide a response to the following request for infomation regarding the handling of heavy loads:
Describe the means provided to assiIre the integrity of a.
concrete hatch covers lifting eye, and any other concrete heavy loads so that they will.not fall apart while being handled during refueling should the lifting eye fail or the load impact other structures.
b.
Alternatively, describe the consequences of failure of the concrete hatch covers or other concrete heavy loads
.during handling. This, evaluation should confirm that unacceptable fuel damage or damage to safety related equipment will not occur.
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BNL COMMENTS ON THE V0GTLE AFWS RELIABILITY ANALYSIS The following is a list of comments from Brookhaven National Laboratory (BNL)
The comments on the VEGP Auxiliary Feedwater System Reliability Analysis.
were drawn from the Technical Evaluation Report (TER) entitled " Review of the Vogtle Units 1 and 2 Auxiliary Feedwater System Reliability Analysis",
which was prepared by BNL.
This TER was attached to a January 10, 1985, letter from Elinor G. Adensam to Donald O. Foster.
Comments:
m.ssmJ pp. 8, 29 - Pump testing procedure requires further g"iscu"ssio*n by the (1)
W applicant. h-tL%F(M @@O'd--L^ MJdD n.3)
MCd.dDf,G.cMd (2) pp.10, 20 - No preaccident operator errors were assumed for the manual valves in the applicant's report. This omission has a significant impact on the quantitative results.
(3) p.10 - Applicant assumed that the probability of a motor-operated valve failing is 5.0E-3/ demand.
BNL assumed a 0.2 recovery factor resulting in a motor-operated valve failure rate of 1.0E-3/ demand.
(4) pp.11, 23 - Westinghous'e Technical Specifications allow for two However BNL and the applicant assumed that inoperable AFWS pumps.
only one AFWS train can be in maintenance at a time.
(5) p.12 - Applicant did act assume maintenance of the diesel generatorC p.12 - Table on p.12 shows discrepancies between applicant's da (6) 3 g g b y J A.<_c 3 1,v d e e w BNL data.
(7) p.16 - The check valves on the pump suctions have had their flappers removed. _This could present operational problems.
p.17 - Emergency procedures for transferring AFWS suction from one l
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CST to the other CST have not been provided by the ap'plicant.
(9) p.17 - Possible comon mode f ailures are discussed in section 9.1.6.
(10) p.18 - Emergency procedures for operation of the AFWS must be provided by the applicant.
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p.19 - Applicant states that unavailability due to testing and common (11) cause human error during testing and maintenance were considered in the tp find these aspects in the fault Howeve BNL was unab
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BNL Comments on the Vogtle AFWS Reliability Analysis Page 2 (13)
- p. 20 - Applicant has maintenance data for DC power, but no data for failure on demand.
Further, there is no event for DC power maintenance in the fault tree, but there is an event for random DC power failure.
(14) pp. 22 In order to perform their own assessment, BNL modified the applicant's fault trees.
One modification included modeling the possibility of maintenance on the steam generator intake check
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valves and stop check valves (Note: This ~ appears to disagree with BNL statement on p.12 that maintenance was not assumed for valves other than motor-operated valves).
Another modification models the operator failing to close the recirculation valve in the condensate system return line.
There were also other minor modifications.to the fault tree.
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DEC 191983
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MEMORANDUM FOR: Elinor Adensam. Chief Licensing Branch M. DL FROM:
C. H. Berlinger. Chief Core Performance Branch. CSI l
SUBJECT:
qutSTIONS FOR V0GTLE UNITS 1 AND 2 Plant Name:
Vogtle Units 1 and 2 Docket Numbers:
50-424 and 50-425 Responsible Branch:
Licensing Branch No. 4 Licensing Stage:
0L Project Manager:
M. Miller Review Branch Involved: Core Performance Branch Description of Review:
Questions Enclosed are questions fbr the Vogtle Units 1 and 2 FSAR review from the Core Performance Branch.
tb!? nalsi ned bu, i
C. H. Berlinger. Chief Core Performance Branch. DSI
Enclosure:
As stated cc:
L. Rubenstein T. Novak M. Miller
Contact:
H. Richings; DSI:CPB X-2M18 93
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e W 0 4 By, HEMORANDUM FOR:
E. Adensam, Cnief Licensing Branch, d4, DL FR04:
C. Berlinger, Chief Core Perfonnance Branch, DSI SU6 JECT.
REQUEST FOR INFORMATI0H FOR V0GTLE UNITS 1 AND 2 Plant N.vae:
Vogtle Units 1 and 2 Docket Nuiabers:
50-424/425 Licensing Stage:
Operating License Responsible Branch:
Licensing Branch #4 Project flanager.
H. Hiller (X-24259)
DSI Review Brancn:
Core Perfomanc'e Branch Paquested Cuapletion Date:
April 25,1984 Heview Status:
Incompiete - Additional Infomation Required The Core Perfonaance Branch has prepared the attached questions on Section 4.4, "Themal-Hydraulic Design," of the Vogtle FSAR. The Physics questions were sent previously. The Fuels questions are being prepared with the aid of a contractor and should be done by June, j
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.a prisinal signeoyt
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y r1 H. Berlinger, Chief Core Perfomance Branch, DSI e
Enclosure:
DISTRIBUTIO!!:
As stated DOCKET FILE:50-424/425 CPS:r/f C. BEP.LI?!GEP.
cc:
R. Mattson L. PHILLIPS D. Eisenhut R. LOBEL L. Rubenstein d'
H. BALUr.JIAN CJ, R. Capra G. Lainas B. Sharon M. Hiller y
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Contact:
H. Balukjian, CPS:DSIA X-i"t43G:,
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.i-o QUESTIONS FOR V0GTLE UNITS 1 AND 2 4 91.1 The discussion given in Section 4.3.2.2.6 relating to (4.3.2.2.6) the total peaking factor is the standard material given (SRP 4.3.II.1) in previous Westinghouse submittals. However, the value of the total peaking factor given is 2.30 rather than the usual 2.32.
Please explain this change from standard practice. If credit is being taken for a change in some aspect of the methodology not previously explicitly presented, reviewed and approved, please
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4 present a detailed justification and indicate relevance and bounds over first and subsequent cycle operation.
4 91.2 For the control rod withdrawal at zero power event (15.4.1.2) please provide quantitative values used in this (SRP 15.4.1.III) analysis for the moderator reactivity coefficient, for the radial and axial power peaking factor and for the axial power shape. Also indicate if these peaking factors are used in figures 15.4.1-2 and 15.4.1-3.
Does the axial peaking and shape bound those extreme top peaked zero power shapes which may exist at end of cycle or in reloads as a result of burnup?
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Comments and Questions V0GTLE UNITS 1 AND 2 j
492.1 Georgia Power Company (GPC) has provided infomation on the loose parts monitoring systen (LPMS) for Vogtle Units 1 and 2 which is called a Metal Impact Monitoring System (MIMS). However, the responses are not complete. Also, Section 4.4.6.4 of the FSAR states that confomance with Regulatory Guide 1.133 is discussed in Section 1.9 of the FSAR.
Section 1.9 of the FSAR states that Westinghouse (W), with GPC concurrence, has taken a position which takes exception to any need for regulatory guidance relative to a LPMS. Also, GPC has taken exception to some items in Regulatory Guide 1.133.
These itens relate to seismic qualification, redundancy, separation, and in-containment calibration. Also, Item C.5.b. Section D and some of the technical requirements are not agreed to. However, the licensee has evaluated the require-ments against an early draft version of Regulatory Guide 1.133 for which some of the requirements have been modified in the final version, Revision 1, May 1981.
The licensee has not provided justification for these exceptions other than arguments with the Regulatory Guide 1.133 criteria.
Since these criteria have been used for licensing for several years and since the cited version of Regulatory Guide 1.133 was issued with due consideration for i
industry comments, the justification provided is unacceptable.
We will require the licensee to provide a LPMS consistent with the provisions of Regulatory Guide 1.133 as has been provided for other licensed reactors and to commit to provide, prior to power operation, a final design report which contains the following:
1.
k evaluation of the LPMS for confomance to Regulatory Eside 1.133.
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A description of the systen hardware, operation an'd implementation of the loose parts detection program, including plans. for start-up testing, acquisition of baseline data and alam settings.
3.
A description and evaluation of diagnostic procedures used to confirm the presence of a loose part.
- 4.. A description of the operator training program.
A sample table of contents of the LPMS description is provided in.
492.2 Standard Fomat and Content of Safety Analysis Reports, Regulatory Guide 1.70, states that in Chapter 4 of the SAR
...the applicant should provide an evaluation and supporting infomation to establish the capability of the reactor to perfom its safety functions throughout its design lifetime under all nomal operation modes..."
Are the analyses presented in Section 4.4 representative of_ the initial core only or have future cycles been enalyzed? Provide a discussion of how power distributions for future cycles are considered in the FSAR analyses.
Is there any assurance that the Vogtle Units can operate at the licensed power level without i
excessive DNB trips throughout future cycles? Will revisions to the design methodelogy be required in order to maintain j
i sufficient themal margin?
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. 492.3 The SRP for 4.4, Thermal and Hydraulic Design, in Section II.9 states that infomation should be provided in response to NUREG-0737. Item II.F.2, " Instrumentation for Detection of Inadequate Core Cooling." Therefore the staff will require the applicant to provide the documentation itemized in Item II.F.2 of NUREG-0737.
492.4 The effects of fuel rod bowing must be included in the themal-hydraulic design.
The predicted extent of rod bow (gap closure) versus exposure and the effect of rod bowing on DNBR must be addressed.
Provide the maximtzn projected assembly burnup and the gap closure for the rod bow penalty. Also, provide a table of rod bow penalty vs burnup (MWD /MTU).
492.5 Operating experience on two pressurized water reactors (not of the Westinghouse design) indicate that significant reduction in core flow rate can occur over a relatively short period of time as a result of crud deposition on the fuel rods.
In establishing the Technical Specifications for Vogtle we will require provisions to assure that the flow rates are not lower than the minimum design flow allowed. Therefore, provide a description of the flow measurenents capability for Vogtle as well as a description of the procedures to measure flow and the actions to be taken in the event of an indication of lower thr.n design flow.
492.6 Please state your intent regarding the use of the Westinghouse optimized fuel assembly in your plant.
If the use of this design is being considered, provide a discussion of the status and schedule for any revised subnittals.
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' 492.7 Please state your intent regarding the use of the Westinghouse
" Improved Thermal Design Procedure" described in WCAP-8567, dated July,1975.. If you intend to use these methods, responses to the following questions will be required:
(a) Provide a block diagram depicting sensor, process equipment, computer, and readout devises for each parameter channel used in the' uncertainty analysis.
Within each element of the block diagram, identify the accuracy, drift, range, span, operating limits and setpoints.
Identify the overall accuracy of each channel transmitter to final output and
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specify the minimum acceptable accuracy for use with the l
new procedure. Also identify the overall accuracy of the j
cutput'value and maximum accuracy requirements for each l
input channel of this final output device.
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(b)
Discuss the method (s) for incorporating environmental a
effects (e.g., noise, ~DtI) on instrument channels into i
the uncertainty analysis.
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(c) Provide data to verify that the plant instruments will perfonn with a high degree of confidence, within their l
design accuracies. This information may be obtained from operating history of identical ins'.ruments installed in i
other plants. This request pertains to the instruments affecting the uncertainties in the Cesign procedures (as identified in question 1 above), the overtemperature AT trip, the high flow trip, the low pressure trip and the i
pump voltage trip.
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. (d) Provide the ranges of applicability of sensitivity' factors.
(e) Demonstrate that the linearity assumption of equation 3-8 in WCAP-8567 is valid when the WRB-1 correlation is used.
492.8 The following items relate to the Technical Specifications which should include:
1.
A declaration that prohibits N-1 loop operation unless it is adequately justified in plant-specific analysis.
2.
Appropriate surveillance to ensure acceptable flow rates and to recognize crud buildup.
3.
A discussion in the basis of the Technical Specification of any generic or plant-specific margins that have been used to offset the reduction in departure to nucleate boiling ratio (DNBR) due to rod bowing.
492.9 In Section 15.1.5 of the Vogtle FSAR the Steam Line Break (SLB) accident two cases are presented:
Case 1, which is at an initial no-load condition with offsite power available; and Case 2, which corresponds to Case 1 with additional loss of offsite power at the time the SI signal is generated.
From the figures presented the minimum RCS pressure appears to be approximately 500 psia for Case 1 (Figure 15.1.5-2) and approximately 950 psia for Case 2 (Figure 15.1.5-5). The following infomation is requested:
1.
What DNBR correlation is used in the analysis for SLB7 l
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Provide infomation on the applicable range of the parameters for the DNB correlation and cmpare with the range experienced (especially pressure) in the SLB accident.
3.
Provide information on the DNBR margin available over the design limit for the SLB.
492.10 Section 4.4.2 of the Vogtle FSAR refers to model tests for obtaining core pressure drop using correlations from one-seventh scale model hydraulic test data of the San Onofre and Connecticut-Yankee reactor models.
Provide infomation on the similarity of these model reactors to the design of the Vogtle reactor and how any design differences were addressed in utilizing the results of the tests.
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ATTACHMENT 1 5 AMPLE TABLE OF CONTENTS LOOSE PART DETECTION PROGRAM DESCRIPTION I.
System Description
A.
Scale piping diagram showing NH sensor locations B.
Sensor specifications (type, manufacturer, sensitivity, temperature rating,etc.)
C.
Sensor mounting details (drawing and procedure)
D.
Preamplifier or line driver (type, annufacturer, location and specifications)
E.
Functional description of LPMS 1.
Theory of operation, detection logic, alarm display 2.
Data recorder specifications (No. of channels, length of is initiated) quency range, and conditions under which recording recording, fre II. Operatio'nal Procedures A.
System Calibration Procedures and Results 1.
Initial and subsequent calibrations 2.
Functional check, as defined in Regulatory Guide 1.133 3.
Channel check, as defined in Regulatory Guide 1.133 B.
Plant Operator Instructions for Use of LFMS 1.
Proceduras for routine operation 2.
Procedures to be used following indication of a loose part a.
Method to confirm existence of loose part b.
Method to diagnose a loose part (size and location)
III. Evaluation for Conformance to Regulatory Guide 1.133 and Justification for any Deviations l
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fl. S.'Ddnenfelo Core Performance Branch U.S. Nuclear Regulatory Comission Attached are the questions (Milestone 1 of Task 8 of FIN No. B2544) concerning tne FSAR for Vogtle-1 and -2 (Docket Nos. 50-424 and 50-425, respectively).
W. J. Bailey and C. E. Beyer Pacific Northwest Laboratory cc:
R. Lobel
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Vogtle Reactor Fuels Section. Core Performance Branch 490.0 The reference in Section 4.2.1.3 for fuel rod models does n Please confirm that it should be Reference 490.1 appear to be correct.
6 rather than Reference S.
Several fuel perfbrmance models, i.e., for rod bowing, fuel and cladding materials properties, creep collapse, the PAD code, and 490.2 rod internal pressure, referenced in Section 4.2 of the Vogtle FSAR are not approved models.
versions (if different from the unapproved versions) of theseIf the resu models been performed for the Vogtle plant?
changed using the, approved models.what are the new results?
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f4ethods and criteria for evaluating fuel cladding 490.3 Have an analysis for the Vogtle plant have not been presented.If so, what are such analyses been perforr.ed for Vcgtle?
results of these analyses?
The analysis of combined seismic and LOCA loads has referenced an unapproved model that has subsequently been a 490.4 If not, applicable to the methods defined in the approved versio'17 has this analysis been perforned using the approved methods and, if so, what are the results?
The use of the CVCS letdown monitor for detecting fuel rod failuresI 490.5 has been explained in the Vogtle FSAR. commitment and fuel f ailures, as per SRP Section 4.27
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Does the analysis of the fuel handling accident (Section 15.7.4 of that the peak pellet burnup of 1
490.6 the FSAR) take into account shown on p. 4.2-2 approximately 50,000 mwd / tonne cf uraniut.
exceeds the value (i.e., approximately 45,000 mwd /t) stated in Footnote a, 3 of Table 15.7.4-2 (Sheet 13 of 13)7 Tables 4.1-1 and 4A-1 show the nominal coolant pressure as 2250 Does the use of 2280 psia for reactor coolant pressure in 490.7 the ECCS analysis (see Table 15.6.5-1) provide more conservative psia.
results?
Cn p. 4.1-2 of the FSAR,itsatts hafnium or silver-indiun cadmium absorber rods are to be used. The NRC staff (
490.8 of representative rods should be carried out,at the first two plants (expected to be Callaway-1 and Comanche Peak-1) to h new hafnium control rods. If for any reason the startup of one of those plants should be delayed beyond the startup of Vogtle and Vogtle 4 to use hafnium absorber rods, then th be innlemented at Vogtie.
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Referenc_e s.
1.
L. S. Rubenstein (NRC), Me:norandum for R. L. Tedesco (NRC), "SSER Input for.Callaway Concerning Hafniun Red Surveillance", June 30, 1922.
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