ML19241B994
| ML19241B994 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 06/14/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19241B993 | List: |
| References | |
| SER-790614, NUDOCS 7907260096 | |
| Download: ML19241B994 (19) | |
Text
sya nuco UNITED STATES Q
+
NUCLEAR REGUL ATORY COMMISSIC" y
jm, g E
WASHINGTON D. C. 20555
- .g "I [j t, M 8
gv
+....
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 39 TO FACILITY OPERATING LICENSE N0. DPR-53 BALTIM0RE GAS & ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-317 1.0 Introduction By application dated February 23, 1979 and supplemental infomation dated January 12, February 7, March 5 and 13, May 7, 29 and 31,1979, Baltimore Gas & Electric Company (BG&E or the licensee) requested an amendment to Facility Operating License No. DPR-53 for the Calvert Cliffs Nuclear Power Plant, Unit No.1 (CCNPP-1).
The amendmenc request consisted of:
Technical Specification (TS) changes resulting from the analyses of e
Cycle 4 relcad fuel, Approval to install a high burnup demonstration fuel assembly (SCOUT) and e
a prototype CEA; and Approval to operate another cycle with modified (sleeved and reduced flow) e Control Element Assembly (CEA) guide tubes.
The associated specified TS changes are described in Section 4.0 of this Safety Evaluation (SE).
2.0 Background
In the Cycle 4 reload application for CCNPP-1 (Ref. 6), BG&E proposed to replace 40 Batch A and 3' Batch C fuel assemolies with 72 fresh Batch F fuel assemblies.
The core rchtoj evaluations are presented in Sections 3.1 and 3.2 of this SE.
In Dccember 1977, a severe CEA guide tube wear problem was identi ied at the Millstone Nuclear Power Station, Unit No. 2.
Similar wear was su )sequently found at CCnPP-1 and other facilities designed by Combustion Engineering (CE).
The temporary repair for CCNPP-1 to allow Cycle 3 operation was to sleeve all fuel assemblies to be placed in CEA locations and the sleeving of other worn fu?1 assemolies in non-CEA locations to regain safety rargins.
Authori za tion fer CCNPP-1 to operate for Cycle 3 in this mode was granted by Reference 1.
P a result of the test progran to evaluate the acceptability of the sleeves for a second cycle of operation, BG&E and CE found that some of the sleeves have become loose in the guide tubes (Ref.14).
The evaluation of the propo;ed repair and the entire CEA guide tube wear problem is presented in Section 3.3 of this SE.
7907260 %,
483 302
_ In the process of this review, we have requested and received additional information necessary for our evaluation (Refs.10,11).
CCNPP-1 is currently licensed to operate at 2700 MWt. The rated power level and all operating conditions remair. the same for Cycle 4.
3.0 Evaluation In this evaluation of a cycle reload for CCNPP-1, considerable use is made of generic reviews of various topical reports (See Topical References).
Most of the topical reports have received formal NRC staff approval.
In all cases where a topical report has not received approval, the report has been examined, its methods judged to be reasonable, and an appraisal has been made that a complete review will not reveal the methodology to be significantly in error.
On this basis, all topicals referenced are judged to be acceptable for this reload evaluation.
3.1 Cycle 4 Fuel Design The 217 fuel c$sembly Cycle 4 core will consist of:
BATCH WEIGHT & (w/o)
NUMBER
- IDENTIFICATION ENRICHMENT FUEL ASSEMBlidS B
1 D
48 D/
24 E
48 E/
24 F
3.03 48 F/
2.73 24
- Irradiated fuel from Cycle 3 As a result of the CEA guide tube wear problem, all fuel assemblies presently in Cycle 3 that will be placed in CEA locations in Cycle 4, with the exception of the Batch B test assembly and one other assembly, will have stainless steel sleeves installed in the CEA guide tubes in order to prevent guide tube wear.
The Batch B test assembly was inspected during the current refueling outage and guide tube wear was found to be acceptable for another cycle of operation.
The center core position occupied by the Batch B assembly is typically a low wear location for fuel assemblies.
The other unsleeved fuel assembly in a CEA position is the result of a three way swap due to a problem sleeved fuel assembly as described in Reference 14 We find operation with two fuel assemblies unsleeved in CEA positions acceptable.
Of the new Cycle 4 fuel, eight Batch F assemblies and eight Batch F/ assemblies will be placed under dual CEAs and eight Batch F/ assemblies will be placed under single CEAs. These 24 new assenclies will have stainless steel sleeves installed in their CEA guide tubes.
BG&E has used the Cycle 3 reload analysis for CCNPP-1 as a " reference cycle" for the Cycle 4 reload analysis.
Our original evaluation of Cycle 3 operation is presented in Reference 1.
A reevaluation of Cycle 3 operation was necessary 483 303
. as a result of the reanalysis performed by BG&E in order to reach the licensed power level (Ref. 2). /.nalyses outside the envelope of the reference cycle have been reanalyzed.
3.1.1 Mechanical In addition to the sleeving of fuel assemblies as described above and evaluated in Section 3.' of this SE, the following other changes have been made to the mechanical, asign of the new fuel assemblies.
Upper End Fitting Assemcly - The holdcown plate in the upper end fitting has been thickened slightly.
Since this reduces the holddown spring working length, the free length of the springs has been reduced by the same amount.
Therefore, the holddown force has remained constant.
Lower End Fitting - The cross-bracing which connects the lower end fitting posts has been thickened and raised 1/8" from the lowermost surface of the fuel assembly.
Guide Tube Flow Holes - 16 Batch F assemblies have guide tube iow holes identical in size to the Batch E fuel. Another 16 assemblies have the reduced flow holes described in Reference 6.
This modification is identical to that made to 16 fuel assemblies installed in the present cycle at CCNPP-2 and evaluated in Reference 3.
The remaining forty fuel assemblies were modified to have slightly less flow than tbe normal Batch E fuel assemblies.
The effect of the modified cooling flow through the CEA guide tubes on the thermal hydraulics of the core will be evaluated in Section 3.1.3 of this safety evaluation.
An analytical prediction of the time of cladding creep collapse for all Cycle 2 fuel has been performed by CE using the CEPAN code which has teen reviewed and approved oy NRC. From this analysis, it has been concluded by CE that the collapse resistance of all the fuel rods is sufficient to preclude cladding collapse during its design lifetjme.
The design lifetime of this fuel will not be exceeded during Cycle 4 operation. The Batch B fuel which is the most limiting witn regard to clad collapse will have accumulated 35,400 Effective Full Power Hours (EFPH) Dy the end of cycle (EOC).
This is below the predicted time to clad collapse whicn has been calculated to be greater than 38,500 EFPH for any standard fuel rod in this assembly. We have reviewed this analysis and found it to be acceptable.
kb '
This cycle will also contain an additional change. This is the installation of a new fuel assembly called Scout which is a high burnup demonstration assembly that will provide information that will be useful in formulating a technical basis for the design, licensing and operation of fuel at high burnups for use in an ex-tended fuel cycle.
The Scout high burnup demonstration assembly consists of 161 standard fuel rods and 15 demonstration rods. The mechanical design of the assembly components other than the 15 demonstration rods in this assembly is identical to the design of the other new fuel assemblies being loaded into the core. The 15 demonstration fuel cins are of two dif ferent mechanical designs.
In one design, which is representa-tive of six fuel pins, the spacer grid contacts the fuel pins at non-fueled regions. This could result in reduced grid / pin contact forces.
To offset this possibility, the initial fill pressure in these rods was increased to decrease the magnitude of clad creepdown.
A larger void volume exists in the rods with the greater initial pressurization which will result in no appreciable increase in the end of life internal pressure.
CE has performed analytical pre-dictions of the cladding creep collapse time for the demonstration fuel rods and has concluded that the collapse resistance of the demonstration fuel rods is sufficient to preclude collapse during their design lifetime. This lifetime will not be exceeded by the Cycle 4 duration.
3.1. 2 Nuclear Analyses Methodology The Nuclear Design Model used in previous cycles has been PDQ, a twc-dimensional diffusion code using four energy groups.
PDQ has been accepted industry wide.
For Cycle 4, CE performed the calculations of certain para-meters using the ROCS code instead of PDQ. Using a higher order differencing methndology than PDQ and only one and a half energy groups, ROCS is able to compute many parameters nearly as accurately as PD0 in three dimensions with more reasonable computer run time.
For Cycle 4, the following safety parameters were computed using the ROC 3 code:
- Fuel Temperature Coefficients
- Moderator Temocrature Coet ficients
- Inverse Boron Worths
- Critical Boron Concentrations
- CEA drop distortion factors and reactivity worths
- Reactivity Scram Worths and Allom nces
- Reactivity worth of regulating CEA banks
- Changes in 3-D core power distributions that result from inlet temperatures maldistributions (asymmetric steam generator transient)
None of these parameters require the detailed knowledge of pin powers normally computed by PDQ.
BG&E states that in most cases, their parameters are cal-culated more accurately by RCCS because of its ability to account for three dimensional effects. BG&E has also stated that they observe guidelir.es to evaluate the adequacy of ROCS for computing these parameters on a case by case basis.
If ROCS is judged to be not adequate for certain computation, then the computation is repeated using PDQ.
483 305
. Based on our review, we find the use of ROCS to be acceptable for this reload.
3.1.3 Nuclear Parameters In the Reference 1 SE, we found that introducing of stainless steel sleeves into the CEA guide tube had minimal effect on reactor physics.
The operation of the CCNPP-1 for one cycle with all CEA guide tube s sleeved has borne out this conclusion.
In the SE supporting the Cycle 2 reload for CCNPP-2 (Ref. 3), we approved a demonstration test consisting of 16 fuel assemblies with reduced CEA guide tube flow.
BG&E has also proposed a 16 fuel assembly demonstration test for Unit 1 Cycle 4.
They anticipate no substantial change in axial and radial power distribution as a result of the decreased flow in the modified CEA This demonstration test will be discussed in Section 3.3 of this guide tubes.
SE.
The licensee has stated that 40 Batch F assemblies have a flow hole configuration that presents a greater flow area and a consequent increase in guide tube flow over the standard Batch E assemblies.
Since the flow area is greater than the standard assemblies by only 4%, the licensee has judged this to have an ins,ignificant effect on axial and radial power distributions.
The Batch F reload fuel is comprised of two sets of assemblies with two enrichments as previously described in Section 3.1 of this safety evaluation. Cycle 4 burnup is expected to be between 10,000 Megawatt The licensee Days per Metric Ton Uranium (MWD /MTU) and 10,555 MWD /MTU.
has examined the Cycle 4 performance characteristics for a Cycle 3 termination point of between 8950 and 10,000 MWD /MTU. The actual Cycle 3 burnup, as stated by the licensee, was 9465 MWD /MTU.
The Cycle 4 moderator temperature coefficieni is calculated to be
-0.4x10-4 AP/"F at the EOC. The values for MTC are bounded by the values used in the reference cycle which are -0.4x10-4aP/ F at beginning of cycle
(?OC) and -2.1x10-4 AP/ F at EOC. We find these v11ues of MTC to be acceptable.
Doppler coefficients calculated for Cycle 4 a; J. 50x10-5 d/*F at BOC hot zero power (HZP), -1.20x10-5dF/ F at OC hot full power (HFP) and -1.37x10-5N'/*F at ECC HFP. These values are slightly more negative at HFP for both BOC and EOC conditions.
Changes of this magnitude, 5, more negative at HFP BOC and 10% more negative at HFP EOC have a minimal impact on the analysis of postulated Anticipated Operational Occurrences ( A00s) and accidents that result in a reactor cooldown. The slightly more negative values of the Doppler coefficient act to add additional conservatism to A00s and accidents during which fuel temperature is tending to increase.
'ac find the values of the Doppler coefficient calculated for Cyc!e 4 to be acceptable.
/} d [
3Ob
6-The total delayed neutron fractio, for Cycle 4 has decreased slightly at E0C and increased s! lightly at COC from that in the reference cycl e.
This would have a mincr impact on the CEA eiection accident.
The CEA ejection accident has been reanalyzed and i., discussed in Section 2.5 of this safety evaluation.
At E0C 4, the reactivity worth of all CEAs inserted, less tN highest worth CEA stuck allowance, is 7.7%dd The reactivity worth n eQuired to shut down the plant including power defect HrP to HZP, shttdown margin and safeguards al awance required tu control the stean line break incident at EOC 4 is b.2%dP. The margin available it negative reactivi ty it 1.5%dp which is more than adequate to account for any uncertainty.. nuclear calculations. We find these shutdown margins to be act.ptable.
3.1 4 Thermal Hydraulics The licensee states that the steady str'e departure from Nucleate Boiling Ratio (DNBR) analyses of Cyc!t. 4 at the rated power of 2700 MWT /MWt has been performed using the TORC code whicn employs the CE-1 DNBR cor-relation. The TORC code has been approved by Reference h for use in licensing and the CE-1 correlation has been approved with a 1.19 ONBR limi t.
TORC /CE-1 was also used in tne generation of limiting conditions for operation (LCOs) on DNBR margin in the TS and all A00s and nostulated accidents which were reanalyzed for Cycle 4 The fuel rod bowing effects on DNB margin for CCMPP-1 have been evaluated within the guidelines set forth in Reference 9, as approved in the reference cycle SE (Ref.1).
A total of 81 fuel assemblies will exceed the NRC-specified DNB penalty threshold burnup of 24,000 MWD /MTU, as established in Refer-erce o, during Cycle 4.
At the end of Cycle 4, the maximum ournup attained by any of tnese assemolies will be 42,800 MWD /MTU.
Frcia Reference g, the corresponding DNBR penalty for 42,800 MWD /?4TU is 6.30 percent.
An examination of power distributions for Cycle 4 shows that the maximum radial peak at hot full power in any of the assemolies that eventually exceed 24,000 MWD / MTU is at least 10.30% less than the maximun radial peak in the entire core. Since the percent in-crease in DNBR nas been confirmed to be never less than the percent decrease in radial peak, there exists at least 10.301, DNBR n.argin for assemblies exceeding 24,000 MWD / MTU relative to the DNBR limits estab-lished by other assemblies in the core. This margin is considerably greater than the Reference f reduction penalty of 6.30% imposed upon fuel assemblies exceeding 24,000 MWD / MTU in Cycle 4.
Therefore, no power penalty for fuel rod Dowing is required in Cycle 4.
483 307
. The modifications to the fuel assemblies to alleviate the CEA guide tube wear problem have a small effect on their thennal hydraulic performanca.
As identified previously in this SE, Cycle 4 will have essentially two different modifications: 1) guide tube sleeving and 2) reduction in quioe tube flow.
The flow characteristics of the assemblies with four 0.25" diameter hole and one 0.125" diameter hole and the assemblies with four 0.25" diameter holes and three 0.093" diameter noles are essentially equiva-lent.
The guide tube sleeving affects thermal hydraulic perfonnance in three core bypass flow, boiling in the guide tuDe sleeve annulus, areas:
and CEA cooling. As stated by the licensee, sleeving reduces the guide tube flow from 1400 lbm/hr to 700 lbm/hr. This change, however, compared to total core bypass flow is a minor effect which is in the conservative direction; i.e., it tends to inc.rease the flow slightly through the core. Bypass flow must be maintained below 3.7% to preserve the design thermal margi n.
Sleaing improves this margic The secon<! area of considerat on is the potential for bol ting in the i
guide tube-sleeve annulus. The licensee states that no boiling will occur in the region in which the sleeve is expanded into contact with the guide tube since the CEA linear heat rate of 3.68 KW/f t is below the boiling limit of 6.5 KW/f t.
In the non-expanded region, axial peaks can be maintained such that CEA linear heat rates are below the 1.2 KW/f t boiling limit. Therefore, boiling is unlikely in this region.
If boiling does occur, slots and holes in the sleeve assure that any expansion due to boiling is relieved and nc mechanical damage will be caused.
It is our opinion that limited boiling in this region is acceptable.
The criteria for adequate CEA cooling is that there is no bulk boiling ia the guide tube during operati7n. The licensee states that cooling flow of 388 lbm/hr is required to meet this criteria. The cooling flow of 700 lbm/hr exceeds the minimum by a substantial margin.
We find this to be acceptable.
The 16 fuel assemblies will have reduced guide tube cooling flow due to the reduction in numoer and size of the flow holes.
The CCA cooling flow for this design has been stated by the licensee to De 565 lom/hr.
This exceeds the bulk boiling criterla of 388 lbm/hr and has a minimal impact in the conservative direction on total core bypass flow. Howeve r, for Cycle 4 none of these 16 assemolies will ce in CEA locatior.s.
The licensee has stated tnat the maximum peaking f actor in any fuel rod in the Scout hign burnup demonstration bundle is predicted to be more than 12% belcw the limi ting pin peak in the core and the maximum pin peaking f actor in any demonstration rod is predicted to be more than Ih b!
300
. 15% below the limiting pin peak in the core. Considering that the bundle geometry of the Scout assembly is identical to the other Batch F assemblies and the Scout assembly power is weii below the limitg core bundle the thermal hydraulic design of this assembly is acceptable.
32 Uncertainty in Nuclear Power Peaking Factors In-core detector measurements are used to compute the core peaki.1g factors using the INCA Code (Ref. c).
The coefficients required to perform this data reduction are performed using the methodology described in the toDical report.
For Cycle 4 operation, the licensee has proposed measurement uncer-tainties of 6% for the total integrated racial peak.ing f actor (Fr) and 7% for the total power peaking f actor (F ) for base load operation and q
R.0", and 10.0% for load follow operation.
The initial CE evaluation of peaking factor uncertainty was presented in References c and d.
In a meeting with CE on March 6. 1979, data was pre-sented showing measurement uncertainty of 6% in r and 7 ', in Fq to be con-e servative (Ref. 8).
On this basis. we find these measurement uncertainties of 6% and 7% for Fr and Fq, respectively, to be acceptable without the load follow operation restrictions.
- 3. 3 CEA Guide Tube Integrity BG&E instituted an Eddi Current Testing (ECT) inspection program at CCNPP-1 to ascertain the condition of sleeves in arsemblies located under CEV s during Cycle 3 (Ref. 4).
No indications of sleeve wear were found in these assemblies, however several guide tube sleeves, when subjected to pull tests. did not exhibit the expected resistance to axial motion (Pef.14).
Because the CCNPP-1 wear inspection program showed ECT signals with widely varying magnitudes at the crimped regions of the sleeves, the inspectico p ogram was extended to assess the crimp size in a number of different type fuel assemblies.
This inspection for crimp integrity was performed using the same probe ?nd test procedure used in the wear inspection program.
The results of these inspections revealed a large nurber of sleeved fuel assemblies outside the ECT and oull test acceptar.ce crit <<'a used at other CE designed facilities. The explanations cf CcNFP-1 results in comparison with the results from the other CE facilities were that the sleeving sequence used at other facilities in 1978 differed from that used at CCNPP-1 (the first facility where sleeving was performed). At the other facilities, pull tests were performed on the sleeves after the crimpirg step to verify the adequacy of the crimp. Following the " crimp verification" pull test, expanding steps were then performed on the sleeves.
However, at CCNPP-i, the pull tests were not performed until af ter bcth the crimping and the evanding steps were completed. The licen_2es and CE have concluded that this sequence change added fricticial resistance between the expanded 483 309
_9 sleeve and the guide tube wall to mask the presence of inadequate crimps that would have been identified by an intermediate " crimp verification" pull test.
In addition, the low ECT results at CCNPP-1, which indicate inadequate crimps, were unique to a particular fuel category.
This fuel category consists of those assemblies that had been irradiated prior to sleeving in 1978.
In this fuel category at CCNPP-1, the EC signals were low for approximately 50% of the 235 sleeves tested.
The low signals for irradiated Del were not evident at the other facilities.
Thus, it appears that the increased yield strength of irradiated guide tubes reduced the displacement of the crimp.
To remedy the observed inadequacy of the crimps at CCNPP-1, a total of 28 assemblies were designated for recrimping, using the new style crimp over the previously made old style crimp.
ECT was performed on each sleeve after recrimping to measure actual crimp size.
The basis of selecting the 28 fuel assemblies was that these assemblies were in the category of those assemblies sleeved in 1978 in the irradiated condi' ton and are to be under CEAs for Cycle 4 operation.
Because the recrirp i. psitioned at some distance from the bottom of the sleeve, a second operation, in which the bottom is re-expanded against the guide tubc wall, was also performed.
This operation, together with a free path geuge check was useo to t w e that the end of the sleeve would not interfere uith CEA insertion.
The lic.ensee stated that bench tests wt e completed on sample guide tube and sleeves to determine effects on sleeve and guide tube geometry by installing a second crimp over a previously incall M crimp.
Results of these test samples showed that the new style crimp can be installed over the old style crimp without " rolling in" the end of the sleeve, or causing any other anomalies io geometry. The tests e ko indicated no need for an additional lower end expansion, however, this srocedurt. was retained in field crimping operations to preclude any chance cf sleeve edge protrusion.
For the actual recrimps placed in tne fuel assemblies in question, all sleeves have been ECT and shown to have crimp sizes sufficient to prevent axial motion (Ref.14).
All other crimping and sleeving operations for this outage have used the new style crimping tools.
The higher crimp pressure inherent with the new style crimp provides a greater force to locally deform (crimp) the higher strength irradiated guide tubes and likewise provides a more defined crimp geometry to resist axial motion of the sleeves.
We have reviewed the proposed crimping, and recrimpino of the CEA guide tubes, and the results of the surveillance tests at CCNPP-1.
Based cn the infor-mation provided in Reference 14, we agree that the guide tube sleeving operations at CCNPP-l provide acceptable repairs to the guide tubes for Cycle 4 operation.
In Reference 14, BG&E stated that ? recommended operational guidelines to reduce relaxation effects in the gui6e tube sleeves during Cycle 4 operation.
This reconnended guideline is to restrict movement of the CEAs at systems temperatures below 400 F except for normal movement assnciatM with refueling opera tions. We find the reconmended operational guideline reasonable. BG&E has agreed to implement this restriction on CEA movement.
483 ai0
4 Sixteen Batch F fuel assemblies have been modified by decreasing the number and size of the flow holes and the size of the bleed holes.
Tests have indicated that the resulting decrease in guide tube flow was accom-panied by less CEA flow-induced vibration and, therefore, less guide tube The SE for CCNPP-2, Reference 3, found the demonstration test wear.
similar to that croposed for CCNPP-1 with 16 fuel asserblies to be acceptable.
The increase in the CEA insertion time to 3.1 seconds was also found acceptable.
We, therefore, conclude that the demonstration test of 16 modified fuel assemblies with reduced guide tube flow is acceptr';1e for Cycle 4 operation of CCNPP-1.
BG&E has agreed to provide a Cycle 5 guide tube evaluation program, identifying changes from the Cycle 4 program at least 90 days prior to the CCNPP-1 shut-down for the Cycle 5 reload outage.
3.4 Analyses of Anticipated Operational Occurrences ( A00s)
Reference 5 discusses the safety analyses of postulated A005 for CCNPP-1 Cycle 4.
The licensee classifies the list of postulated A00s into two cate-gories.
The first category includes those A00s for which the Reactor Pro-tection System (RPS) Limiting Safety Systam Settings (LSSS) as specified in the plant 15 assure that the Specified A ceptable Fuel Design Limits (SAFDLs) are not exceeded. The second category includes these A00s for which initial steady state overpower margins are maintained by adherence to the Limitir.g Conditions far Operation (LCOs) specified by the TS for the plant.
Adherence to the LCOs assure that SAFDL limits are not exceeded.
The loss of flow transient causes the most rapid change in DNBR and both a reactor trip and steady-state overpower margin is required to maintain the SAFDLs.
The LCO: and LSSSs for Cycle 4 TS were calculated using the methods described in Reference f.
The required A00 reanalyses were done using the computer code CESEC (Ref. i).
The licensee stated in Reference 5 that the need for reanalysis of a particular A00 is determined by comparison of the key parameters for that A00 to those of the last cycle for which a complete analysis was performed.
If the key parameters are within the envelope of the reference cycle data, no reanalysis is required. A re3nalysis might also be performed in case it could lead to a significant relaxation of TS.
The results of that comparison show that the key parameters to all the A00s and postulated accidents for Cycle 4 operation are tne same as the specified reference cycle input parameters, except for the following:
1.
CEA drop t.ime to 90% inserted 2.
Integrated radial peaking factor (Fr) 3.
Seized rotor pin census 4.
Core bypass flow fraction 5.
RTD response time 48?
3il
. For all A00s and postulated accidents other than those reanalyzed, the licencee has stated that the CCNPP-1 safety analysis submitted ei ther in the FSAR or in previous reload cycle license sub-mittals bound the results that would be obtained for Cycle 4 and demonstrate continued safe operation of CCNPP-1 at 2700 MWt.
Since the CEA drop time to 901 insertion has increased for Cycle 4, the loss of Flow Event, CEA Ejection Event, RCS Depressurization Event, Seized Rotor E vent and the CEA Withdrawal Event were reanalyzed. These events are adversely impacted by the CEA drop time, since a reactor trip is necessary to terminate the event.
The sleeving of the CEA guide tubes has a negligible effect on CEA rod drop times but the reduction of the CEA guide tube flow holes does impact on the rod drop times. As previously stated, the Cycle 4 reload will have 16 fuel assemblies with reduced flow holes. The eff(Ct of these flow holes on rod drop times is to increase the time to 90%
insertion from 2.5 to 3.1 seconds. oG6E has identified this as a proposed change to the TS 3.1.3.4 at this tim 2, even thougn none of these assemblies are under CEAs during this cycle.
To assess the impact of th's change in rod drop time, the licensee has examined all thE design basis events which could require a trip to prevent exceeding SAFDL limits.
An evaluation of these design basis events showed that only five events may be adversely affected by in:reased scran time.
For these evaluations, it was conservatively assumed that all the CEAs are inserted at the same insertion versus time characteristic cur ve as in the 16 fuel assemblies with the reduced guide tube flow. Those transients which were reanalyzed are dis-cussed below.
BG&E has proposed a change to the TS Table 2.2-1 raising the higt power level trip from 105.E% to 107.0% power. The safety analysis assumes a trip at il2%
of rated power. A 5% power msurement uncertainty has always been applied in the process of generation LSSS limits.
In the past, this uncertainty was applied in a multiplicati re f ashion (which yields the equivalent of a 5.5%
of rated power uncertainty), but evaluations showed tn t application of the uncertainty in this fashion is conservative.
In accordance with current methods (as described in Reference f), the power measure-ment uncertainty is new deducted algebraically.
It is this difference in tne manner in which the uncertainty is applied that leads to tha 107% versus 106.5t LSSS limi t.
We have reviewed this change and find it to be acceptaole.
3.4.1 CEA Withdrawal Event The CEA Witndrawal event was reanalyzed for Cycle 4 due to tne increase in the Resistance Temperature Detector (RTD) response time to envelope future cycles and the increase in the CEA drop time to 90% insertion from 2.5 seconds to 3.1 seconds. The CEA Withdrawal event was re-analyzed for reactor initial conditions of zero power and full power and the licensee has stated that the Departure from Nucleate Boiling (DNB) and fuel centerline melt Specified Acceptaole Fuel Design Limits (SAFDLs) will not be exceeded during CEA Witndrawal transient.
/{ ?])
bk2
. The CEA Withdrawal transient initiated at rated themal power results This Dias factor in the maximum pressure bias f actor of 62.0 psia.
accounts for measurement system processing delays during the CEA The pressure bias factor for this cycle has in-Withdrawal event.
creased from the reference cycle due to the increase in the RTD time constant and the increase in the CEA drop time to 90% insertion. This pressure bias factor is used in generating tmh ? trip setpoints to prevent the SAFDLs from being exceeded during a CF.A Withdrawal Event.
The TS have been changed to reflect the 62.0 psia pressure bias factor.
We find this analysis and the change to the plant TS to be acceptable.
3.4.2 RCS Depressurization Event the RCS Depressurization event was reanalyzed for Cycle 4 to assess the impact of increasing the CEA drop time to 90% insertion from 2.5 seconds for Cycle 3 to 3.1 seconds for Cycle 4.
As stated in Reference f, this is one of the events analyzed to detemine a bias term input to the TM/LP trip. Hence, this event was analyzed for Cycle 4 to obtain a pressure bias factor. This bias factor accounts for neasurement system processing delays during this event.
The trip setpoints incorporating a bias factor at least this large will provide adequate protection to prevent the DNBR SAFDL from being exceeded during this cvent.
The analysis of this event shows that the pressure bias factor is 35 psia which is less than that required by the CEA Withdrawa'i Event.
Hence, the use of the pressure bias factor determined by the CEA Withdrawal event will prevent exceeding the SAFDLs during an RCS Depressurization event.
3.4.3 Loss of Coolant Flow Event The Loss of Coolant Flow event was reanalyzed for Cycle 4 to determine the impact on margin requirements that must De built into the LCOs due to the increase in the CEA drop time to 90% insertion.
The low flow trip setpoint is reached at 1.0 seconds ano the CEAs start dropping into the core one second later. A minimum DNBR of 1.25 is reached at 2.3 seconds.
The low flow trip, in conjunction with the initial overpower margin maintained by the LCOs in the TS assure that the minimum CNBR will be greater than or er ' to 1.19 for the Loss of Coolant Flow Event.
4B:
313
a 3.4.4 Conclusion We have reviewed the licensee's analyses of A00s for Cycle 4 operation of CCNPP-1 and conclude that they are acceptable.
3.5 Postulated Accidents Other Than LOCA The licensee has reviewed the postulated accidents other thar. LOCA.
Reference 5 discusses the safety analysis performed for this category of accident for CCNPP-1 Cycle 4.
Postulated accidents as other plant events, need to be reanalyzed only if the key parameters influencing the_ event are not enveloped by the reference cycle data. Those accidents that were reanalyzed are discussed below.
3.5.1 CEA Ejection Event The CEA Ejection Event was reanalyzed for Cycle 4 to assess the impact of increasing the CEA drop time to 9fA insertion and the increase in the augmentation factor in comparison to the reference cycle.
In addition, the zero power case was analyzed due to the decrease in axial peak in comparison to the reference cycle. The reference cycle for this event is the analysis upon which the licensing of CCNPP-2 Cycle 2_was based.
Our evaluation of this reload is found in Reference 3.
Hence, this event was reanalyzed to demonstrate that the criterion for clad damage is not exceeded during C.ycle 4 operation.
lhu licensee's analysis shows thct for both the zero power and full power cases the clad damage pellet enthalpy threshold of 200 cal /gm is not violated. Therefore, no fuel rods are predicted to suf fer clad damage.
3.5.2 Seized Rotor Event The Seized Rotor event was reanalyzed for Cycle 4 due to the changes in the followino key parameters.
e The increase in the CEA drop time to 901, insertion The decrease in core bypass flow, which increases the net core flow a
5 The decrease in tne Radial Peaking Factor e A more adverse (flatter) pin census.
48!
3i4 The increase in the CEA drop time and the flatter pin census adversely impact the consequences of this event.
Increasing the net core flow and decreasing the Radial Peaking Factor will decrease the consequences of this event. Hence, a reanalysis was performed for Cycle 4 to ensure that only a small fraction of fuel pins are predicted to f ail during a Seized Rotor event.
A conservatively " flat" pin census distribution (a histogram of tne numoer of pins with radial peaks in intervals of 0.1 in radial peak normalized to the maximum peak) was used to determine the number of pins that experience DNB.
The results indicate that increasing the core flow and decreasing ti e radial peaking f actor offset the i.icrease in the CEA drop time to 90t insertion.
It was calculated that for Cycle 4, less than 0.5% of fuei pins will experience DNB for even a short period of time.
For the case of the loss of coolant flow arising from a seized rotor shaf t, it is assumed that there is an instantaneous reduction to three pump flow. The low flow trip assures that less than 0.5% of fuel pins experience DNB. This is the same as that calculated for the reference cycle.
Hence, the conclusions reached for reference cycle remain valid for Cycle 4.
3.5.3 Conclusions We have reviewed the accident analyses for events other than LCCA for CCNPP-1 Cycle 4 and concludJ that they are acceptable.
3.6 Cycle 4 LOCA Analysis Reference 5 provides a comparison of the fuel specific parameters for the limiting fuels e2 ring Cycles 3 and 4.
The Cycle 4 core contains 216 high density fuel assemblies and one low density Batch B assembly. Tne highest power pin in the low density Batch B assemDly will not achieve a power level greater than 751 of the highest power pin in the core.
Therefore, a Batch B fuel pin will not ce limiting in C.vcle 4.
483 M5
. The remaining 216 high density fuel assemblies contain 72 partially depleted Batch D assemblies, 72 <1rtially depleted Batch E assemblies and 72 fresh Batch F assemblies. Burnup dependent calct.lations were performed for the high density fuel 3ssemblies with the FATES (Ref. b) and STRIKIN-II(Ref.a ) codes. The results dem. strate that the most limiting fuel pin curing Cycle 4 is located in one of the partially depleted Batch E assemblies.
The limiting high density fuel in Cycle 4 has a stored energy 266*F lower than the limiting fuel in Cycle 3.
Consequently, the ECCS per-formance results reported for Cycle 3 conservatively bound the perform-Therefore, tt e peak linear heat generation rate of arce for Cycle 4.
14.2 KW/f t which was demonstrated to be acceptable for Cycle 3 is also an acceptable limit for Cycle 4 operation.
In order to comply with 10 CFR 50, Appendix K, the LOCA analysis must demonstrate that the peak clad temperat ure (PCT) remains below 2,200 F and the maximum local -ladding oxidation, which is a function of the time dependence of the PCT, remains below 17 percent.
During a LOCA, the cladding swells due to the decreased coolant pressure The clad swelling is and tre increased fuel temperature and gas pressure.
terminated if the cladding ruptures.
The Rupture-Strain curve is a plot of clad strain (clad swelling) vs clad temperature at the point of clad rupture in a LOCA Event. The Rupture-Strain curve is :n integral part of the CE ECCS flow blockage model.
Recently the NRC staff has determined that, for clad rupture which occurs during the reflood phase of the LOCA, the Rupture-Strain curve used by CE is possibly nonconservative.
- However, this is not a problem for CCNPP-1,because clad rupture is predicted to occur during the blowdo'en phase and not the reflood phase.
The staff reviaw has found the CE analyses for the case of rupture during the blowdown phase ts be acceptable.
We conclude, as a result of our review, that the CCNPP-1 Cycle 4 ECCS performance is in coi.formance with the criteria specified in 10 CFR 50.46(D) and is, therefore, acceptable.
4.0 Technical Specifications The TS changes proposed for this amendment are sunnarized in the following statenents.
Pace 1-3 The ce#inition of Shutdown Margin (Section 1.13) would be revised to eliminate the reference to part length CEAs.
Me 2-7 The vower Level-High RPS trip would be increased 0.5% to 107.0% as a result of the Cycle 4 analyses.
483 316
,. Pages 2-12 & 2-13 Figures 2.2-2 and 2.2-3, relating to the TM/LO trip setpoint, would be modified as a result of the Cycle 4 analyses.
Page 3/4 1-23 The CEA drop time, TS 3.1.3.4, would be increased from 2.5 seconds to 3.1 seconds as a result of the changed hy?,3ulic characteristics of the 16 demonstration fue' assemblies.
Pages 3/4 2-4 & 3/4 2-5 New axial flux offset (Figure 3.2-?) and augmentation factors (Figure 4.2-1) would be added based on revised physics calculations.
Paaes 3/4 2-8 & 3/4 2-9 These pcwer distribution limit changes would be made basrd er ;evised pnysics calculations and application of the standard CE setpoint tethodology.
Page 3/4 2-11 Figure 3.2-4 would include the increase in allowable azimuthal tilt.
Page 3/4 2-13 The old TS 3.2.5 would be eliminated since the core can not achieve a core exposure that would result in clad collapse.
Pace 3/4 2-15 Table 3.2-1 would be revised to increase the cold leg temperature used in DNB calculations by 1 F to 548 F.
Parameter values for less than four RCP operation would be eliminated pending NRC review of ECCS analyses for operation in that mode.
Pace 3/4 3-6 Table 3.3-2 would be revised to increase the RTD response time from 5 to 8 seconds in accordance with the Cycle 4 analysis.
5.0 Physics Startup Testing The physics startup test program as described in Reference 6 has been reviewed.
The low power tests include CEA symmetry check, critical boron concentration measuremen.s, isothermal temperature coefficient measurements and CEA group worth measu,ements.
The power ascension tests include power coefficient and power distribution tests.
The staff discussed the CEA symmetry test and the review criteria for this test with the licensee.
The licensee agreed to perform the CEA symmetry test on 2 shutdown banks and review criteria as stated in Reference 13.
The review criteria for power distribution measurements are also given in Re fe renc e. ~l.
kb
The staff finds the entire program including the acceptance and review criteria and the remedial actions acceptable.
6.0 Environmental Consideration We have detemined that the amendment does not authorize a change in effluent types o' total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further cont.luded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 651.5(d)(4), that an environmental impact statement, or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of tais amendment.
7.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase ir.
the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities wi.11 be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the commcn defense and security or to the health and safety of the public.
Dated: June 14,1979
/} 8 [.
3b
. TOPICAL REFERENCES CENPD-135-P, "STRIKIN II, A Cylindrical Geometry Fuel Rod Heat Transfer a.
Program," August 1974, February 1975 (Supplement 2-P), August 1976 (Supplement 4-P) and April 1977 (Supplement 5-P).
b.
CENPD-139, "CE Fuel Evaluation Model", July 1974.
c.
CENPD-145, " INCA: Method of Analyzing In-Core Detector Data in Power R6 actors", April 1975.
d.
CENPD-153, " Evaluation of Uncertainty in tne Nuclear Form Factor Measured by Self-Powered Fixed In-Core Detector Systems", August 1974.
CENPD-161-P, " TORC Code - A Computer Code for Determini g the Thermal e.
Margin of a Reactor Core," July 1975.
f.
CENPD-199-P, "CE Setpoint Methodology", April 1976.
g.
CENPD-225, " Fuel and Poison Rod Bowing", October 1976.
h.
Evaluation of Topical Report CENPD-161-P, K. Kniel (NRC) to A. E. Scherer (CE), September 14, 1976.
i.
CENPD-107, "CESEC-Digital Simulation of a CE Nuclear Steam Suroly Fystem,"
April 1974.
l} [j !
') \\ l
. LETTER REFERENCES 1.
NRC Amendment No. 32 for CCNPP-1, Cycle 3 Reload, R. W. Reid, to A. E. Lundvall, March 31,1978.
2.
NRC Amendment No. 33 for CCNPP-1, Cycle 3 Reanalysis, R. W. Reid to A. E Lundvall, June 30,1978.
3.
NRC Amendment No.18 for CCNPP-2, Cycle 2 Raload, R. W. Reid to A. E. Lund/all, October 21, 1978.
4.
BG&E Sleeved CEA Guide Tube Inspection Program, A. E. Lundvall to R. W. Reid, January 12, 1979.
BG&E High Burnup Demonstration Program, A. E. Lundvall to R. W. Reid, February 7,1979.
5.
6.
BG&E Aoplication for Cycle 4 Reload, A. E. Lundvall to R. W. Reid, Feb; uary 23, 1979.
7.
BGSE Supplement 1 to Application for Cycle 4 Reload - B4C Type CEA Design, A. E. Lundvall to R. W. Reid, March 5, 1979.
8.
CE Data Justifying Measurer..cnt Uncertainties of 6 percent in Fr and 7 percent in Fq, A. Sherer to P. Check, March 7, 1979.
9.
BG&E Supplement 2, Application for Cycle 4 Reload - Small Break LOCA Analysis, J. W. Gore to R. W. Reid, March 13, 1979.
10.
NRC Request for Additional Information, R. W. Reid to A. E. Lundvall, April 13, 1979.
11.
BG&E Reponse to Cycle 4 Reload Questions, A. E. Lundvall to R. W. Reid, May 7, 1979.
12.
NPC Safety Evaluation - Small Break LOCA Analysis with NL Credit for Charging Pu o Flow, R. W. Reid to A. E. Lundvall, May 18, 1979.
13.
BG&E Supplement 4 to Application for Cycle 4 Reload - Physics Startup Testing, A. E. Lundvall to R. W. Reid, May 29, 1979.
14.
BG&E Supplement 5 to Application for Cycle 4 Reload - CEA Guide Tube Test Results and Repairs, A. E. Lundvall to R. W. Reid, May 31, 1979.
483 320