ML19326C404
| ML19326C404 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/03/1978 |
| From: | David Williams ARKANSAS POWER & LIGHT CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| 1-038-5, 1-38-5, NUDOCS 8004220951 | |
| Download: ML19326C404 (18) | |
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REGULATORY INFORMATION DISTRIBUTION SYSTEM ( R'I DS )
DISTRIBUTION FOR INCOMING MATERIAL 50-313 REC: REID R W ORG: WILLIANS D H DOCDATE: 03/03/78 NRC AR PWR R< LIGHT DATE RCVD: 03/17/78 DOCTYPE: LETTER NOTARIZED: NO COPIES RECEIVED LTR 1 ENCL 40
SUBJECT:
FURNISHING RESPONSES TO NRC"S LTR DTD 02/07/78 AND TELECOYP REC"D 02/17/78, RE QUESTIONS PERTAINING TO PROPOSED TECH SPEC FOR CYCLE 3 RELOAD REPT (BAW-1471).
PLANT NAME ARKANSAS - UNIT 1 REVIEWER INITIAL:
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DISTRIE ; TION OF THIS MATERI AL IS AS FOLLOWS ******************
GENERAL DISTRIBUTION FOR AFTER ISSU NCE OF OPERATING LICENSE.
(DISTRIBUTION CODE AOO1)
FOR ACTION:
DR CHIFF RFID**W/7 ENCL INTERNAL:
CT!E**W/ENr NRC PDR**W/ ENCL OELD**LTR ONLY AI P, F**W/z c.w_ m HANAUER**W/ ENCL CHECK **W/ ENCL EISENHUT**W/ ENCL SHAO**W/ ENCL BAER**W/ ENCL BUTLER **W/ ENCL GRIMES **W/ ENCL J COLLINS **W/ ENCL J.
MCGOUGH**W/ ENCL EXTERNAL:
LPDR'S RUSSELLVILLE, AR**W/ ENCL TIC **W/ ENCL NSIC**W/ ENCL ACRS CAT B**W/16 ENCL DISTRIBUTION:
LTR 40 ENCL 39 CONTROL NBR:
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ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 UTTLE ROCK ARKANSAS 72203 (501)371-4000 March 3, 1978 l
1-038-5 Director of Nuclear Reactor Regulation ATTN:
Mr. Robert W. Reid, Chief Operating React' ors Branch No. 4
.g U. S. Nuclear Regulatory Commission 1
^
1 Washington, D. C.
20555
Subject:
Arkansas Power & Light Company Arkansas Nuclear One-Unit 1 1
Docket No. 50-313 License No. DPR-51 Cycle 3 Reload Report Questions
- 43,
(File: 0242.5, 1511.1)
~~
Gentlemen:
Our December 28, 1977 submittal concerning proposed technical Specifi-cations for Cycle 3 operation of Arkansas Nuclear One - Unit 1 (ANO-1) provided to you our Cycle 3 Reload Report (BAW-1471) as supplemental information. This document was prepared to meet the requirements out-lined in the USNRC document, " Guidance for Proposed License Amendments Relating to Refueling," dated June 1975.
Your letter of February 7, 1978 and a telecopy we received on February 17, 1978 requested information in addition to that described in the above mentioned USNRC document. Please find enclosed our responses.
In ordar to distribute and implement Cycle 3 technical specifications on a schedule to support our current startup schedule (begin heatup on March 20, 1978), we request your immediate issuance of cycle 3 technical specifications. We appreciate your cooperation in our efforts not to delay Cycle 3 startup due to licensing.
Very.truly yours,
<L 46 1 *
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'cn Daniel H. Williams Manager, Licensing DHW:Je Enclosures
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I RESPONSES TO CYCLE 3 RELOAD QUESTIONS ARKANSAS NUCLEAR ONE - UNIT NO. 1 DOCKET No. 50-313 QUESTION A1.
Describe in detail the tests to be serformed to check for a misloaded assembly. What assurances are there that the core is as expected before going to powers > 5% rated power?
RESPONSE
Following the completion of the refueling shuffle sequence, the core loading was verified by underwater television camera. The camera was positioned over each fuel assembly in sequence and two individuals independently read and reported the serial numbers. These numbers were written on a blank core map, and upon completion of the full core scan, the serial numbers were compared to the final core loading plan supplied by the core designers. This verification was recorded on video tape. Additionally, the following physics tests conducted at low power (<5%FP) give
ed assurance that the core is loaded properly. The all rods or soron concentration measurement which
. compares measured to predicted critical boron concentrations, the measurement of the regulating CRA group worths and comparison to predicted worths and the measurement of the worst case ejected CRA and its comparison to prediction all provide additional assurance that the core has not been misloaded.
QUESTION A2.
Describe the procedures for the control rod-trip test.
Include the acceptance criteria and the procedures to be followed if the acceptance criteria are not met.
RESPONSE
Initial RCS conditions are established at a temperature of approxi-mately 5320F, a pressure of 215%t 30 psig, all (4) reactor coolant pumps running, with Boron at refueling concentration. Control rod groups 1 through 7 are fully withdrawn and group 8(APSR's) are with-drawn approximately 25%. The control rod drive mechanisms are then tripped via the manual trip button. The insertion times for each control rod from its initial position to its 3/4 insertion point is measured by the plant computer Rod Drop Timer program. The printout of this program includes trip initiation time, initial position and trip insertion time for each control rod (excluding group 8).
The acceptance criterion'is that the measured time from trip initiation to 3/4 insertion shall not exceed 1.46 seconds at full Reactor Coolant Flow conditions.
If this acceptance criterion were not met, the reactor would not be taken critical until further analysis and study were performed which would justify operation. The NRC would be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and presented all the details known at the time and would be kept informed throughout. NRC approval of any findings which would justify l
continuation of the startup would be required prior to proceeding.
QUESTION A3.
Provide the details of the procedures for the critical boron concen-tration tests. Discuss how corrections are made to the measured data and how the measured data is compared to the predictions. What are the acceptance criteria and what are the procedures if the acceptance criteria are not met?
RESPONSE
The boron concentration for criticality with all rods out except Group 7 at approximately 85% wd is estimated. The control rods are withdrawn to the "all rods out" position and the estimatel amount of DI Water to achieve criticality is added using continuous feed and bleed. A 1/M plot (using source range instrumentation) versus boron concentration is maintained and the critical boron concentration is projected as the approach to criticality proceeds. When the boron concentration nears the projected critical concentration, the letdown flow rate is reduced and the boron sampling frequency is increased. When criticality is achieved, deboration is terminated and the control rods are withdrawn slightly to establish a positive startup rate.
Power is leveled off 10-9 amps on the intermediate range and the boron concentration is at allowed to come to equilibrium. Equilibrium boron concentration is verified by sampling the RCS, MU Tank and Pressurizer.
The remaining reactivity held in the inserted portion of Group 7 is then measured by withdrawal of Group 7 to its out limit and ' concurrent doubling time measurements. The doubling time is converted to reac-tivity and the reactivity to equivalent boron concentration change using the boron differential reactivity worth. The All Rods Out boron concentration is the sum of the measured boron concentration and the equivalent boron from the reactivity measurement during rod withdrawal with appropriate corrections for Xenon and Samarium concentrations at the time of the measurement.
The acceptance criterion placed on critical boron concentration is that the actual boron concentration must be within + 100 ppm of the value predicted by the core designer.
If the acceptance criterion were not met, the reactor would be taken at least 1% AK/K subcritical until review by the company safety connittees determined the proper course of action. The core designer would be asked for recommendations and probably further testing at hot zero power would be recommended to attempt to gather more infor-mation. NRC would be notified and kept informed. No operation above
- 5%FP would be attempted without resolution of the discrepancy.
QUESTION A4.
Describe in detail the procedures and methods used for the temperature reactivity coefficient tests. Also provide the acceptance criterion and the procedures to be followed if the acceptance criterion is not met.
RESPONSE.
The isothermal temperature coefficient is measured at approximately the all-rods-out configuration. The average coolant temperature is varied by first decreasing then increasing temperature by 5 F.
During the change in temperature, reactivity feedback is compensated by discrete change in rod motion, the change in reactivity is then cal-culated by the summation of reactivity (obtained from reactivity cal-culation on strip chart recorder) associated with the temperature change.
The acceptance criterion states that the measured value shall not differ from the value predicted by the core designer by more than + 0.4 x 10-4 aK/K F.
If th more than + 0.4 x 10 g measured value exceeds the predicted value by AK/K F, an alditional evaluation will be per-formed. Moderator coefficient of t ? activity is calculated in conjunc-tion with temperature coefficient measurement. After temperature coefficient is measured, a predicted value of fuel doppler coefficient of reactivity is added to obtain moderator coefficient. The measured moderator temperature coefficient is extrapolated to 95% FP per TechnicalSpecification3.1.7andthisvalueisgerifiedtobeless than the acceptance criteria limit of +0.5 x 10-O AK/K F (the value used in the FSAR for accident analysis).
If the extrapolated value should exceed +0.5 x 10-4 F, the reactor would be taken to at least one percent subcritical and additional review and analysis performed. The NRC would be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and NRC approval of any findings which would justify continued operation would be needed prior to continuing with the power ascension.
QUESTION A5.
Provide the details of the regulating control rod group reactivity worth tests. Give the predicted worth of each group to be measured, and the stuck rod worth and the predicted total worth for all rods.
RESPONSE
The procedure for measuring control rod group worths is to deborate CRA groups 7, 6 and 5 into the core from an all rods out condition while maintaining criticality. Reactivity is measured with an on-line Reactivity Calculator during discrete changes in CRA group movements.
The summation of these changes in reactivity are calculated to obtain integral rod worths.
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W The predicted worths for control rod groups to be measured, the stuck rod worth and the predicted total worth for all rods is listed below:
Control Rods Predicted Worth
@ hot zero power %AK/K Group 5 1.02 Group 6 0.97 Group 7 0.70 Stuck rod 2.50 Total 1-7 8.83 Total 1-8 9.23 QUESTION A6.
Describe in detail the procedures for the ejected control rod reactivity worth test.
State the methods used to compare the measurements with predictions and the acceptance criteria. Also, include procedures if the acceptance criteria are not met.
RESPONSE
Starting with the Reactor critical with control rod groups 5, 6 and 7 inserted, the worst case ejected control rod is borated out of the The reactor is adjusted to just critical by motion of Control core.
Rod Group 5.
The worth of the ejected rod is calculated using the change in boron concentration, the differential boron worth and the change in Group 5 position.
The measured worth of the ejected control rod is corrected for any deviation in the Group 5 position for 0% withdrawn following the boration of the ejected rod. This correcticn is performed by cal-culating the worst case ejected rod worth, adjusted for the inserted worth of control rods. The calculated worst case ejected rod worth is a function of the worth of Group 5 withdrawn from the core at the start of the test and the measured worth of the worst case ejected rod.
This value is then compared to the acceptance criteria which is that the deviation between the predicted and measured ejected rod worth does not exceed +20%.
Next, we calculate the error adjusted worst case ejected rod worth by considering the uncertainty factor associated with the use of predicted rod worth data.and the uncertainty factor associated with the use of the boron swap method.
~This value is then compared to the safety analysis limit of 1% AK/K.
If this value were exceeded, the reactor would be immediately taken to greater than 1% AK/K suberitical. No further critical operation would be conducted until further review and resolution of the discre-pancy were made. NRC would be nutified and NRC approval of any findings which would justify continued operation would be required prior to resumption of the unit startup.
- QUESTION A7.
AND-1 had a quadrant eilt at the beginning of Cycle 2.
How did this tilt change during the cycle? How was the presence of this tilt used in the predictions of the power distributions for Cycle 37
RESPONSE
At 13.8 EFPD into Cycle 2, ANO-1 had quadrant power tilts of.10, 1.04, 480, and -1.93% in quadrants 1 through 4 respectively. Up until approxi-mately 35 EFPD, the tilt in quadrant 4 decreased slightly in absolute value while the tilts in the other quadrants remained about the same so that at 33.5 EFPD the tilts were. 11,.99,.70, and -1.80.
At some time between 33.5 EFPD and 40.3 EFFD, the quadrant tilts changed; at 40.3 EFFD the tilts in quadrants 1 through 4 were 1.50,.55,
.50, and
-1.55 respectively. The tilts remained quite constant for the remainder of the cycle; for example, the tilts at 254.6 EFPD were 1.61,.61,
.71, and -1.52.
The cycle 2 tilts were not included in the prediction of the cycle 3 power distributions. The cycle 3 maneuvering analysis accounts for the presence of tilts much larger than those encountered in cycle 2; further, a symmetric rod worth test performed near the end of cycle 2 indicated the quadrant-to-quadrant reactivity difference to be very small indicating that no appreciable quadrant tilt existed at the end of Cycle 2.
QUESTION A8.
Provide the details of the core power distribution tests. Describe in detail the methods used to predict the assembly by assembly power as well as the analyses of the data obtained during the measurements.
What are the assembly by assembly acceptance criteria? How are tilts accounted for in the analysis of the data? If a 1/4 or 1/8 core map is the result of the measurement, what method is used to determine the assembly power for those assemblies having their symmetric assemblies instrumented? For example, are the measured assembly powers averaged, or is only one of the symmetric measurements used?
RESPONSE
Core power distribution data will be obtained at various power levels and compared with predicted data to assure compliance with operating limits and Technical Specifications. Power Imbalance, Quadrant Power Tilt, Linear Heat Rate, DNBR, and Power Peaking Factors will be analyzed.
For this test, 40% will mean 40% + or - 2%FP, 75% will mean 75% + or
- 2%FP and 100% will mean highest attainable power without exceeding 100%FP. Equilibrium Xenon will be defined as a condition where reac-tivity change is less than.01% AK/K per hour with all other reactivity change contributions stable. Equilibrium Xenon will not be required for the 40% tests. Control rod index is established at a position correspond-ing to the rod positions where core power distribution predictions were calculated.
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The acceptance criteria are as follows:
i)
The maximum linear heat rate in the core is less than the LOCA limit per Technical Specifications for the axial location of the peak. When testing at a power level below rated power, the maxi-mum LHR when extrapolated to rated power must also meet this criteria.
- 11) The minimum DNBR must be greater than 1.30 at rated power conditions and when extrapolated to rated power conditions from a lesser test plateau.
iii) The quadrant power tilt must not exceed the value allowed in the Technical Specifications.
iv) The highest measured radial and total power peaking factors shall not exceed the highest predicted peaks by more than 5% and 7.5%
respectively at the 75% and 100% power plateaus.
These acceptance criteria are established to verify that core nuclear and thermal hydraulic calculational models are conservative with respect to measured conditions thereby verifying the acceptability of data i
from these models for input to safety analysis. The acceptance criteria also serve to verify safe operating conditions at each test plateau and eventually at rated power conditions.
Predictions for the radial and total peaks at 40, 75, and 100% FP are calculated using the FLAME-3 with thermal-hydraulic feedback code
'BAW-10124). Radial peaks are calculated from the predicted power oc'put for each assembly in a 1/8 core. Total peaks are calculated fru, the predicted power output of the maximum segment for each assembly in a 1/8 core.
Assembly and segment power representations are calculated by the on-line computer based on current-signal outputs from the 52 incore detector strings. Any tilt which exists in the core is inherent in the measure-I ment of neutron flux by the incore detector system. Only instrumented assemblies are utilized in the analysis of the data to calculate measured radial and total peaks for comparison to predicted radial and total peaks. Symmetric instrumented locations are averaged to provide a single value for the assembly or segment power in the 1/8 core location.
Radial and total peaks are then calculated. As previously stated, the eaximum measured radial and total peaks are compared to maximum pre-dicted radial and total peaks. There are no criteria for comparisons 4
on an assembly by assembly basis.
Tilt effects are accounted for in the calculation of DNBR and linear heat rate.
If a tilt does exist, a routine in the on-line computer adjusts the segment power representations of an instrumented assembly in order to provide segment power representations of a symmetric, non-instrumented assembly. DNBR and linear heat rate are calculated by the on-line computer for the maximum assembly in each of the four core flow
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n regions. These values are then compared to acceptanca criteria previously discussed.
In addition, a hand calculation of linear heat rate is per-formed in order to obtain values for comparison with LOCA acceptance criteria which are level dependent.
QUESTION A9.
Provide details of the' power Doppler reactivity coefficient and tempera-ture reactivity coefficient measurements near full power. What methods are used to compare measured values with predictions? What are the acceptance criteria for these tests and what procedures are followed if acceptance criteria are not met?
RESPONSE
The moderator temperature coefficient at power operating conditions is measured by varying Tave using Tave setpoint controller on the Reactor i
Demand Station and maintaining constant power with the ICS in full auto.
The corresponding control rod metion is related to the reactivity change to determine the temperature coefficient. The power doppler coefficient is measured by varying Reactor Power using the ICS Unit Load Demand Station and recording the corresponding control rod motion.
The reactivity change due to the control rod motion is determined and from that the power doppler coefficient is determined.
The measured values are compared directly to the acceptance criteria placed on this test which are:
1)
That the moderator coefficient be non-positive at power levels greater than 95%FP. This is conservative with respect to the value used by safety analysis of 0.0 at 100% FP.
ii) That ghe power doppler coefficient be more negative than -0.55 x 10- AK/K%FP. This value is based on a minimum conservative value derived from the minimum doppler coefficient and fuel temperature versus power relationship used in the safety analysis.
The accident analyses which are power doppler coefficient dependent are power increasing transients such as rod ejection.
If these acceptance criteria were not met, the reactor would be returned to at least 1% AK/K subcritical and additional review and analysis per-formed. The NRC would be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and NRC approval
- of any findings which would justify continued operation would be needed prior to continuing with the power ascension.
QUESTION A10.
Discuss the changes to be made to the plant's computer prior to Cycle 3 operation. This should describe:
a.
What elements change from Cycle 2 to Cycle 3: coefficients, constants, correlations, etc., and why they change'.
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c.
What quality assurance procedures (testing) are used at the site to verify-that changes have been correctly made.
RESPONSE
The following elements in the plants computer sof tware will be changed prior to cycle 3 operation due to-the loading of a new fuel enrichment and the shuffling of the remaining fuel assemblies -
1.
Signal to power conversion factors for the fuel.
2.
Boron, Xen'on, and Fuel and Samarium worths for the core.
i 3.
Core Macroscopic Fission and Xenon Absorption Cross Sections and Xenon and Iodine Yields.'
4.
Isotopic concentrations versus burnup for the fresh fuel.
5.
Constants for new incore detectors.
6.
Integrated quantities related to cycle ht; tory are initialized and shuffled to account for changes in the core loading.,
Items 1 through 3 above were calculated using the same PDQ models used in designing and analyzing cycle 2.
Item 4 was calculated using the NULIF cell code.
Item 5 is based on as-built incore detector data and experimentally determined factors.
All cycle 3 sof tware changes are implemented 'while the reactor is in cold shutdown. The as implemented sof tware is tested with test case data which has been previously run on an independent off-line computer system which has the same software implemented. The outputs of the test cases run on the plant computer and on the independent off-line computer a
are compared for agreement. Any differences found are resolved prior to plant startup.
QUESTION Bl.
What is the fill gas pressure for Batch 57 Compare to fill gas pressures of previous batches. Explain any changes from batch to batch.
RESPONSE
The initial fill gas pressure for Batch 5 is
- psi. This was the same for 'all previous batches except batch 1, which had an initial fill gas pressure of
- psi. The higher pressure assures resistance to~ creep coll..pse for ANO-1 design conditions.
- Proprietary information to be supplied separately.
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QUESTION B2.
What are the predicted peaking factors for Cycle 3 and their associated uncertainties? This includes radial, axial, local and local peaking factors. Provide the comparable Cycle 2 predicted and actual BOC, mid-cycle, and EOC peaking factors.
TESPONSE The predicted peaking factors for cycle 3 are:
Radial 1.359 Axial 1.213 Radial Local 1.105 Axial Local 1.026 at 4 EFPD nominal full power conditions.
The associated uncertainties are + 5% on the radial and i 7.5% on the total peak.
The cycle 2 predicted and actual peaking factors are:
Predicted Measured Radial - BOC 1.343 1.394 Total * - BOC 1.661 1.704 Radial - MOC 1.252 1.296 Total * - MOC 1.507 1.538 i
Radial - EOC 1.226 1.260 Total * - EOC 1.433 1.468 The cycle 2 values are given in terms of radial and total peaking factors since these give the most straightforward comparison to measured values.
Also, cycle 2 values do not contain local effects since these are not applicable in the case of measured peaking factors.
- Segment average. total peaking factor consistent with the method of measurement. The discrete value may be somewhat higher.
QUESTION B3.
Provide added information as to how the variable low pressure trip and the flux / flow trip setpoints provide the thermal margin for the 11.2%
DNBR rod bow penalty.
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RESPONSE
The rod bow DNBR penalty has been incorporated directly into the thermal-hydraulic analyses used to define RPS limits and setpoints. The penalty is incorporated by increasing the required MDNBR from the 95/95 correlation limit (1.30) by the bow penalty less any credits which were included in Se analysis which may be used as a direct offset against rod bow.
B&W st..erically includes a flow area reduction factor in subchannel and iso-lated hot channel thermal-hydraulic analyses. B&W has claimed, and the NRC staff has approved, a 1% rod bow DNBR credit for all thermal-hydraulic analyses which include the flow area reduction factor.
The flux / flow analysis for ANO-1, cycle 3 included the flow area reduction factor and a credit of 1% was taken against the 11.2% bow penalty. The minimum required DNBR used to establish the flux / flow setpoint was increased from 1.300 to 1.433 (10.2% increase). This resulted in a maxi-mum allowable, or thermal-hydraulic limit, flux / flow trip setpoint of 1.091.
This limit value corresponds to the maximum setpoint which can be used without violating the DNBR criteria. Corrections were then applied to the thermal-hydraulic limit to account for flow measurement errors and flow signal noise. The maximum allowable Technical Specification flux /
flow setpoint of 1.060 covers the 11.2% DNBR rod bow penalty and accounts for measurement errors, flow signal noise and the maximum delay time on control rod motion after a trip signal.
The analysis to determine the variable-low-pressure trip function included the flow area reduction factor (1% DNBR credit) and also in-cluded a 4% DNBR penalty resulting f, rom a densificant power spike. B&W no longer considers the effects of a densification power spike in DNBR analyses and the 4% penalty was taken as a credit against the 11.2%
rod bow penalty. This same credit was also approve 3 in a December 5, 1977 letter from Steven A. Varga, NRC to James H. Taylor, B&W concerning updates to the " Fuel Densification Report," BAW-10055. The remaining DNBR rod bow penalty was incorporated directly into pressure-temperature limit calculations by increasing the required MDNBR by 6.2%.
The current cycle 2 variable-low-pressure trip function covers the 11.2% rod bow penalty and no changes are being made for cycle 3 operation.
QUESTION B4.
What 100% core flow value is assumed in your analysis?
RESPONSE
The 100% core flow value assumed for thermal-hydraulic analysis is the cycle 1 design flowrate of 88,000 gpm per pump. Current analyses in support of ANO-1, cycles 2 and 3 use 106.5% of the design flowrate. As previously discuased in revisions to BAW-1433 (ANO-1, cycle 2 Reload Report), the measured reactor coolant system flow is in excess of 109.7%
of design flow. While this excess flow provides a real thermal margin for plant operation, no credit is claimed in thermal-hydraulic evaluations.
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QUESTION B5.
' Describe how the fission products of the five Batch I fuel assemblies being utilized in Cycle 3 have been accounted for in the nuclear cal-culations. Discuss. how the concentrations from these assemblies were accounted for in the nuclear calculations. Discuss also the effects of the uncertainties in concentration distribution in these assemblies,
, as well as other batch concentration uncertainties, on power distributions 4
and peaking factors.
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RESPONSE
j Nuclide concentrations for the five Batch I fuel assemblies were recover-ed from the end of cycle 1 fuel cycle design calculations for ANO-1.
The fission products I-135 and Xe-135 are set to zero because of their j
short half-lives. The Pm-149 is decayed to Sm-149 and added to the Sm-t 149 present at the end of cycle. The remaining lumped fission pro-ducts are considered stable (as in the ENDF representation) so their concentrations are-those calculated at the end of the cycle. This technique is the same method as used in the design of Three Mile Island-Unit I cycle 3 which used thirteen reinsarted fuel assemblies.
Comparisons of measurements and predictions for the thirteen re-inserted 4 -
Batch I fuel assemblies at Three Mile Island-Unit I cycle 3 have dis-played a Root Mean Square percent difference less than 1%.
Therefore, the uncertainty associated with the above method of treating re-inserted assemblies is small and is well within the uncertainty associated with design models. Similarly, the effects of uncertainty associated with the other batch concentrations on power distributions and peaking factors have been shown to be acceptable by comparisons of measured and predicted power distributions at the startup of each reload cycle. Power distri-bution verifications are planned at 40%, 75%, and 100% full power during the startup of cycle 3 at ANO-1.
i The questions raised in this section have been the subject of audits of j
the Region IV NRC I&E inspector and the results of his audit are docu-i mented in Report No. 50-313/77-25. However, brief details of the fuel inspection process are provided in answers below:
QUESTION C.
Provide a description of the on-site QA/QC functions relating to the handling of nuclear fuel. This description should not be limited to, but should include in detail, a discussion of the following.
1.
Inspections to be performed on new fuel assemblies on-site and the acceptance criteria.
2.
Inspections to be performed on the five (5) Cycle One (1) fuel assemblies which will be reloaded during Cycle Three (3), and the acceptance criteria.
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Inspections to be performed during the Cycle Three (3) fuel shuffle and the acceptance criteria.
(This snould include spent and recycled fuel assemblies).
4.
QA/QC documentatf be available on-site.
5.
Inspector qualifications and training.
RESPONSES C1.
All new fuel assemblies are 100% visually inspected for defects found as a result of shipping / handling. The integrity of all shipping containers is verified and evidence of mishandling is looked for. The fuel cladding, spacer grids and end fittings are inspected for scratches, dings or perforations.
C2.
The five fuel assemblies which were removed after cycle 1 and are to be used again in cycle 3 have been 100% visually inspected with under-water television equipment for evidence of damage as a result of fuel handling operations or resulting from the burnup they received in cycle 1.
t During these inspections, each side of each fuel assembly was inspected
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using the underwater video system. The bottoms of these essemblies were inspected after being discharged from the reactor after Cycle 1.
The tops of the fuel assemblies were observed during semiannual inventory.
No visual evidence of damage to the five assemblies was observed. The inspections included examination for grid damage, fuel rod damage, fuel rod bowing, unusual crud patterns, signs of abnormal heat flux and lodged foreign materials.
C3.
Visual inspections of all assemblies as loaded in the core for verification of proper orientation, axial elevation and control component placement have been made. One fuel assembly was removed during the shuffle and-visually inspected completely to check for damage as a result I
of handling problems encountered.
It was found completely acceptable and returned to the core. Following core loading, 10% of discharge fuel was visually inspected in the spent fuel pool with underwater video equipment. The assemblies to be inspacted were chosen based on previous i
handling experiences, and to obtain a spectrum of core locations radially.
During this inspection, the sides of all fuel assemblies were viewed and 4
inspected for spacer grid damage, fuel rod perforations, fuel rod bowing, signs of abnormal heat flux or unusual crud patterns. No such findings were encountered. The fuel appeared to be in excellent physical condition.
Visual examinations were recorded on video tape.
C4.
The quality assurance equipment certificates for all fuel assemblies have been received, reviewed and are stored on-site. The records of on-
- site inspections of new and' irradiated fuel are also documented and stored on-site.
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C5.
All fuel inspections performed in preparation for, Cycle 3 have been per-formed by personnel from the Plant Nuclear Engineering' Staff. All four' per-sonnel involved have nuclear sngineering degrees (2 of which have Masters de-grees and 3 of which are registered Professional Engineers). Three of the four personnel involved have previously visited fuel fabrication facilities and are familiar with fuel desigr. and QC requirements. All personnel involved reviewed the fuel inspection procedures prior to performing inspections and all had. previously inspected fuel assemblics at Arkansas Nuclear One under the supervision of previously experienced personnel.
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RESPONSES TO 3/17/78 TELECOPY QUESTIONS QUESTION 1.
Please provide information on the status of the B&W Setpoint Methodology topical report. It is important that this report be submitted to provide the data and information necessary to interpret the Technical Specification changes of plant setpoints.
RESPONSE
B&W plans to submit topical reports, BAW-10121, 'RPS Limits and Setpoints' and BAW-10122, ' Normal Operating Controls' during the first quarter of 1978. These topicals will provide information which will permit inter-pretation of Technical Specification changes associated with plant setpoints.
QUESTION 2.
Please explain the increase in quadrant tilt allowed in the proposed Technical Specifications (from 3.4% to 4.92%). You state that the increase in allowable tilt is a result of an increase in calculated margin. Please explain how the allowable tilt is calculated from the margin available and identify how tradeoffs in other core parameters are made to arrive at the allowable tilt.
RESPONSE
The additional allowable peaking corresponding to the higher quadrant tilt is compensated for by statistically combining the nuclear uncer-tainty factor, the hot channel factor and the rod bow peaking penalty.
In addition, the power spike penalty due to fuel densification was not used. Also, by specifying APSR position limits for this cycle, addi-tional control of power peaking (and thus margin) is provided. References and further documentation are available on page 24 of BAW-1471.
QUESTION 3.
Please justify allowing your plant to operate with up to a 25% tilt before the plant must be placed in hot shutdown (3.5.2.4.3 on pg. 47 of proposed Technical Specifications). This appears to be an excessively large tilt and it seems that shutdown should occur well before the tilt reaches this magnitude.
RESPON"E The allowable 25% tilt limit must be viewed in the proper context of the Tech Spec Operat'on with a tilt in excess of 4.92% but !.ess than 25% is only allowed when the following required actions are implemented:
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a a) the core power must be reduced to below the power level cutoff (92.0% full power).
b) the power must be further reduced 2% for each 1% tilt in excess of 4.92% tilt.
c) less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> have elapsed
- since the tilt has exceeded 4.92%.
As an example, if a 10% tilt occurred, the allowable core power would be reduced to below 92-(10-4.92) (2) = 81.84% full power
- If more than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> have elapsed, reductions in RPS, rod insertion and operating imbalance limits must be implemented.
Further, if a 20% tilt occurred, the allowable core power would be reduced to below 92-(20-4.92) (2) = 61.84% full power At this red' iced power (and at other allowable combinations of tilt and power) the initial conditions for accident analyses are preserved.
QUESTION 4.
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The proposed Technical Specifications state that above 15% power, the quadrant tilt will be monitored at a minimum frequency of once every two hours. Justify not making the rate of tilt surveillance a function of the magnitude of the tilt.
It seems prudent that once the tilt exceeds the allowable tilt prior to power reductions, more frequent monitoring of the quadrant tilt should take place.
In the case at hand, Arkansas 1 is allowed a tilt of 4.92% prior to any power reductions and must shutdown only when the tilt reaches 25%. Between these two tilts it appears that the tilt surveillance should increase.
If you disagree please justify your selection of monitoring frequency.
RESPONSE
The frequency of monitoring quadrant tilt must be considered along with the Tech Spec action requirements. Between the tilt values of 4.92% and 25%, the required actions are:
a) the core power must be reduced to below the power level cutoff (92.0% full power) b)
the power must be further reduced 2% for each 1% tilt in excess of 4.92% tilt.
c) less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> have elapsed
- since the tilt has exceeded 4.92%.
- If more than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> have elapsed, reductions in the reactor protection system, rod insertion, and operating imbalance limits must be implemented.
With these added restrictions (required actions) it is not deemed necessary to require more frequent monitoring of quadrant tilt.
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'-* 5' QUESTION 5.
f Paragraph 4.1 of your reload submittal stated that "... improved test
. methods (dynamic impact testing) show~that the spacer grids have a higher seismic. capability..." What are these improved test methods and where are they documented?
RESPONSE
The spacer grid impact tests'were-performed to determine the spacer grid stiffness, damping, and maximum allowable impact force. The
-dynamic test methods for the Mark B (15x15) spacer grids is the same as described in the " Mark-C Fuel Assembly Topical Report on LOCA - Seismic Analyses," BAW-10133, October 1977, Section 3, which is currently under NRC review, f
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