ML20039G423
| ML20039G423 | |
| Person / Time | |
|---|---|
| Site: | 05000514, 05000515 |
| Issue date: | 01/14/1975 |
| From: | Stello V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201180209 | |
| Download: ML20039G423 (6) | |
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50-514 50-515 V. A. Moorn, Assistant' Director for Light 'L' ster Reactors 2. L 11:121/.L QUESTIO::S POR PELELL sin 1!:CS P.1: e t l' rec:
Pebbic Springc Licenring Step,c:
CP Dochet 1:os. :
50-514 & 50-515 Rerponsibic Branch LUR 2-3 mi Project I:r. nager:
C. Stahic Technical Review Branch Involved:
Core Perforrnnce Branch Description of Review:
First Round Questions
- Ihe first round questions fro:s the Reactor Fuels Section and the Physics Section of the Core Performt.nce Branch concerning the Pebble Springs l'SAR are enclosed.
These questions relate to Chapters 4 (Section 4.2 and 4.3) and 15 of the PSAR.
l,j VictorLStello,
, Assistant Director for Reactor' Safet'y Directorate of Licensing
Enclosure:
First Round Questions cc: 'S. Hanauer F. Schroeder A. Cintubusso W. Mcdonald A. Schwencer D. F. Ross L. Rubenstein P. Check C. Draney W. Brooks W. Hodges P. Atherton C. Stahle L. Chandler V. Stello E. Ieins 4
T'J01180209 810403 PDR FOIA MADDEN 80-515 PDR A
240 -1 241.0 REACTOR TUELS SECTION - CORE PERFOUn' CE BRANCH 241.1 Provide the taris for the fuel henJ11ng and shipping design (4.2.1.1) loads that are presented in subsection 4.2.1.1.1.
To what extent have these design loads been confirmed experimentally?
241.2 Provide information on the folloding items:
(4.2.1.1)
'(1) The physical properties of the cladding as a function of tenperature and irradiation (t..bular or equation form).
The following infornation should be included:
(c) Modulus of clacticity (b) Poisson's Ratio (c) Coef ficient of thermal expannion (d) Yic3d strc:.s (c) Ultimate stress (f) Thermal ccnductivity (g) Uniform ultimate strain I
(h) Specific heat (2) The variation in melting point and fuel conductivity with burnup.
241.3 Provide justification for the stress-strain limits presented (4.2.1.1) in the paragraphs under' Section 4.2.1.1.
4 241.4 Provide the deflection design specifications and experimental (4.2.1.2) observations for the upper and lower plenum sp' rings.
Provide j
the maximum deficction, both expected and possible, for the lower plenum spring.
4 241.5 Under the paragraph entitled " Fuel Rod Fabrication," it is in-(4.2.1.3) dicated that the internal and external surfaces of the Zircaloy tubing are cicaned just before the fuel pellets are loaded.
Provide information in regard to the materials and the pro-I cedure used in this c1 caning operation, k
241.6 Provide the design bases for Zr-4 irradiation growth and supply i
(4.2.1.3) supporting data or references.
241.7 Discuss in detail the surveillance, inspection and testing of (4.2.1.4) if radiated fuel rods.
An exampic of one post. irradiation examin-t (iLion plan is ASTM-E-453, " Examination of Fuel Element Cladding l
Including the Determination of Mechanical Properties."
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240-2 242.0 REACTORPliYSICSSECTION-COREPERFOPJ!AhCEBRANC11 242.1 The following definitions in Table 1.1-1 need clarification:
(1.1.3.1)
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(1)
Administrative Controls.
Administrative Controls are not defined.
(2) llot Shutdown Condition.
This definition is not cicar.
In keeping with currcat uork on technical specifications change the definition to "when reactor is cuberitical by at Icast 1% Ah /k and the reactor core-average coolant temperature is at or near hot zero power temperature (549'F)."
(3)
Reactivity.
This definition is lengthy and uncicar.
It should not incluJe direccnionn of g or pcm because they are not used in the PSAR.
(4)
Quadrant Power Tilt.
This definition is for maxhaum quadrant power tilt. Thedefinition should be changed as follows:
(a)' The term "the maximum upper excore detector" should be changed to "an upper excore detector".
(b)
The term "the maximum lower excore detector" should be changed to "the corresponding lower excore detector".
242.2 Specify the fuel enrichment for each of the fuel regions shown (4. 3. 2.1) in Figure 4.3-1.
Specify the poison content of the Burnable Poison Rod Assemblies (BPRA) shown in Figure 4.3-2, and discuss ~
any variation,with core location.
Specify the poison content of the Control Rod Assemblies (CRA) and the Axial Power Shaping Rod Assemblics (APSRA) used in the analyses of Sections 4.3 and 15.
242.3 Specify the plutonium content corresponding to 100% plutonium (4. 3. 2.1) in Figure 4.3-4, and plot plutonium - 239 content as'a function of burnup.
242.4 The text gives a Scff value of.00691 at BOL.
Figure 4.3-3 (4. 3. 2.1) should show this Scff at BOL and should show Scff as a function of burnup early in life.
242.5 The definition of Ifmiting power distribution is inconsistent (4. 3. 2. 2. 2) with the use of it to define offset limits.
Which axial power shaping rod maloperations cause limiting power distributions?
i Are these maloperations included in Tech Spec requirements?
Discuss the reactor conditions (e.g., design basis power man-cuver, mispositioned CRA, etc.) which cause limiting power distributions.
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242.6 The reference to topical' report EAW-10061 should be deleted (4.3.2.2.5) becauce this report has been withdrawn by BLV.
242.7 Tbc applicant states that there is a maneuvering range for (4.3.2.2.7) the APSRAs.
Does this mean that power distribution analyses
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I were not made for the APSRAs out of the maneuvering range?
Is the APSRA caneuvering range included in the Tech Specs, and are alarms or inhibits used to keep the APSRA's in their maneuvering range?
^'?.C D:pirir uhy the noderator tcmporature coefficient for LOL is
?.2.3)
Jer rocitive for Pcbb3c Springs than Oconce 2 uhen the ppm of reluble boron is cuch higher in Pebble Springs than Oconce 2.
_..z.9 Are the uncertaintics in the reactivity coefficients included
(!.3.2.3) in the values uced in the ccfety, analyses?
242.10 Uhat power 1cvel, ppm soluble boron, and CRA positions were (4.3.2.3) used for the derivation of Figure 4.3-23?
242.11 What ppm soluble poison and amount of rod insertion were used (4.3.2.3) for the derivation of Figure 4.3-24?
242.12 In Table 4.3-9, the moderator temperature deficit for the tem-(4.3.2.3) perature change from 549 to 602*F appears inconsistent with the moderator temperature coefficients indicated by Figure 4.3-23.
Explain.
242.13 In Table 4.3-10, the moderator deficits for the temperature (4.3.2.3) change from 549 to 602*F at BOL and EOL appear inconsistent with the moderator temperature coefficients indicated by Figure 4.3-23.
Explain the inconsistency.
Also, provide the moderator temperature coefficients used in the calculation of the moderator deficit (549 to 602*F) for both BOL and EOL a
for the equilibrium cycle.
242.14 The BOL boron icvels calculated for all CRA inserted and for one (4.3.2.3)
CRA stuck out in the 70*F condition with a 0.99 kett in Table 4.3-11 do not agree with the total rod worth and stuck CRA wcrth values of Table 4.3-10.
Please explain the differences.
242.15 Present an EOL control rod worth tabic similar to Table 4.3-12.
f (4.3.2.6)
F 242.16 Discuss the difference between the HFP sequential worths of j
(4.3.2.6) the Group No. 2 safety rods in Tabic 4.3-12 of the Pebble Springs l
and Ecliefonte plants.
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1 240 -4 242.17 Diccuss the difference betueen the total inter:cdiate tceperature (4.3.2.6)
(300*F) sequentini worths in Tabic 4.3-12 of the Pebble Springs and the Bellefonte plants.
s 242.18 Why are the moderator coefficient values in Table 4.3-13 (4.3.2.6) different from those in Figure 4.3-237 1
242.19 Do the fuel asscnbly configurations discursed in this -section (4. 3. 2. 7) and Tabic 4.3-8 centain any poicon rods fren CR*'r !. PSP.A's, or EPRA's?
If they do, prencnt the 1:cgg values for the assenbly configurations eithout any of the poison ar.senb]ies.
242.20 Show that the c:. -ecre dctectern enn provide pret s ction t;.. inst (4. 3. 2. 8.1) exceeding safety liuits for anf al nenon oscillations stcrted in the ccntral part rf !!.e n neter cere.
Alen. prcvide,irrtification that the eight APSRA's are capable-of suppressing all axial xenon oscillations.-
242.21 What is the calculated fast flux at the locations of the material (4.3.2.9) surveillance specimens as shown in Figure 4.2-47 Show that these specimen locations nect the requirencnts of part II.C.2 of 10CFR50 Appendix H.
242.22 Provide detailed descriptions of ths B&W computer programs LIFET, (4.3.3)
RIP, IAME, and AMOP.
Include comparisons with measurc:ents and other codes.
242.23 The accident analyses should have been calculated for an initial (15.0) power 1cvel of 3672 FBCE per Regulatory Guide 1.49.
The accidents must be reanalyzed for that power level.
242.24 The nominal moderator and Doppler coefficient values in Table 15.0-2 (15.0) used for the accidents do not agree with the values given in Tables 4.3-4 and 4.3-13 and Figure 4.3-23.
Explain why the Table 15.0-2 values were used for safety analyses. Also, were coef-
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ficient uncertainties included in a conservative fashion?
l 242.25 Analyze and show in figures the values of the minimum DNBR and (15.3) the maximum fael centerline temp,erature for the control rod misoperation accident.
242.26 The analysis of possibic misloaded fuel accidents is incomplete.
(15.15)
All analyses for this accident should show the effects of the accident on the reactor power distribution and reactor instrumen-tation. Misloadings of BPRA's, CRA's, and APSRA's should be analyzed including the loading of a BPRA vith the wrong number of burnabic poison rods or the wrong poison content in the number of burnable poison rods.
Analyses should also be made of loading 4
Is 240-5 242.26 (C;nt) fuel ansemblies, EPRA's,' CRA's, or APSRA's, with crpty Zircaloy (15.15) tubes or with missing pellets.
linalyses should be made for interchanges of fuel assenblics with and without burnable poison rods.
Show that any loading errors which can cause fuel limits to be exceeded will be detected by reactor instrumenta-tion and the reactor tripped before significant fuel damage occurs.
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