ML20004C758
| ML20004C758 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 05/22/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20004C756 | List: |
| References | |
| NUDOCS 8106050177 | |
| Download: ML20004C758 (7) | |
Text
_
(
["%
UNITED STATES 3"
1 NUCLEAR REGULATORY COMMISSION j,
j WASHINGTON, D. C. 20555 e
s) u SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULAT. ION SUPPORTING AMENDMENT ~NO. 70 TO LICENSE NO. DPR-46 NEBRASKA PUBLIC POWER DISTRICT DOCKET N0. 50-298 COOPER NUCLEAR STATION' l.0 INTRODUCTION Nebraska Public Power District (the licensee) requested amendments ~
to the Technical Specifications for the Cooper Nuclear Station (CNS) by letter dated March 5,1981. The amendments are associated with core Reload Number 6, Cycle 7 operation.
2.0 CORE RELOAD NUMBER 6 2.1 Introduction By letter (l) dated March 5,1981, the Nebraska Public Power District (the licensee) requested amendment to the Technical Specifications appended to Operating License DPR-45 for Cooper iluclear Station (CNS).
The proposed changes relate to the refueling of CNS. This reload' involves the replacement of 32 exposed 7x7' fuel assemblies and 80 exposed 8x8 assemblies with an equivalent number of fresh, two water.
rod, P8x8R fuel assemblies designed and fabricated by the General Electric Company (GE).
In support of this reload application for CljS the licensee has submitted a supplemental reload licensing document (2}
prepared by GE and proposed plant Technical Specification changes.(3)
The descriptions of the nuclear and mechanical design of the fresh P8x8R fuel assemblies the exposed 8x8R fuel assemblies and-the exposed standard 8x8 fuel assemblies, which were used in connection with prjor CNS reloads, are contained in GE's generic licensing topical reportl4)-
for BWR reloads.
Reference 4 contains a complete set of references to other GE topical reports which describe GE's BWR reload methodologies for the nuclear, mechanical, thermal-hydraulic, transient and accident analysis calculations.
Infonnation addressing the applicability of these methods to reload cores containing a mixture of 7x7, 8x8,.8x8R and P8x8R fuel is also contained in Reference 4.
Portions of the. plant-specific data, such as operating conditions and design parameters used in transient and accident calculations, have also been included in the topical report.
8106 050l77
-2 Our safety evaluations.(6,7) of GE's' generic refoad licensing topical ~
report 'and supplement concluded that the nuclear and mechanical design of the 8x8R and P8x8R fuel and GE's analytical methods:for the.
nuclear,- thermal-hydraulic and transient and. accident. calculations, as
~
applied to mixed cores containing different fuel; types are acceptable.
As part of our evaluations (6) of Reference 4, we fo'und the cycle-independent input data for the reload transient _and accident. analyses for CNS to be acceptable. The supplementary cycle-dependent-infor-mation and input data are provided in ~ Reference 2, which follows-the fonnat and content of' Appendix A of Reference 4.
Finally,:the licensee l.
has changed the' initial core pressure used:in the transient analyses from 1045 psia to 1035 psia, to reflect: actual. plant-operating data.
As a ' result of the staff's generic evaluations (6,7) of a substantial number of safety considerations relating'to the use of P8x8R reload' fuel in mixed core loadings with 7x7, 8x8 and:8x8R fuel,'only a limited:
number of additional review items are included in this evaluation of Cycle 7 of CNS. These items include the plant and cycle-specific input, data and safety analysis res its presented in Refe'rencel3,.those-items reload reviews, and the p_roposed Technical Specification changes.g3 identified in our evaluation ) as requiring special considerattop d s'
2.2 Evaluation 2.2.1 Nuclear Characteristics Reload 6 consists of 112 new P8x8R fuel bundles with bundle average-i enrichments of 2.83 and 2.65 wt% U-235. The' remainder of the 548 fuel.
assembly reconstituted core will consist of. irradiated 7x7,' 8x8, 8x8R ~
and P8x8R fuel assemblies exposed during earlier cycles. The assumed -
cycle exposure has increased from 17,110 mwd /t.for Reload 5 to :17,441.
mwd /t for Reload 6.
The reference core loading-for Reload 6 is'shown in Figure 1 of Reference 2 will' result in quarter core symmetry, which
~
is consistent with previous cycles.
The reload application follows the procedure described in Reference 4.
i~
We have reviewed this application and the consequent Technical Specifi-cation changes. The transient analysis input parameters provided in-Section 6 of Reference 2 are typical for BWRs and are acceptable. Core l-wide-transient analysis results are given for the limiting transients 1 -
and the required operating limit values for MCPR are-given for each fuel' type. The revised MCPR limits are required by the reload and they are~
acceptable.
l'
=w-
-re,-ee.m-
-.,,.-w.%
,m,,_
4
3-2.2.2 Thermal Hydraulics Fuel Cladding Integrity Safety Limit MCPR As stated in Reference 4, for BWR cores which reload with GE's P8x8R.
fuel, the allowable minimum critical power ratio -(MCPR) resulting from either core-wide or localized abnormal operational transients is equal to 1.07.
When meeting this MCPR safety limit during a transient, at least 99.9% of the fuel rods-in the-core are expected.to avoid boiling transition.
The 1.07 safety limit minimum critical power ratio (SLMCPR) to be used for Cycle 7 is unchanged from the SLMCPR previously approved for Cycle 6.
The basis for this safety limit is addressed-in Reference 4, while our generic approval of the limit is given in References 6.and 7.
Operating Limit MCPR Various transient events can reduce the MCPR from its normal operating level. To assure that the fuel cladding integrity safety -limit MCPR will not be violated during any abnormal operational transient, the most limiting transients have been reanalyzed for this reload by the licensee, in order to determine which event results in.the largest-reduction in the minimum critical power ratio. These events have been-analyzed for the exposed 7x7, 8x8 and 8x8R fuels and for the fresh P8x8R fuel. ' Addition of the largest reductions in critical power ratio (a CPR) to the safety-limit MCPR establishes the operating limits.for each fuel
+
type.
The transients evaluated were the limiting pressure and power increase transient.(in this case, the load rejection transient without turbine bypass to the main condenser), the limiting coolant' temperature decrease _
transient (loss of feedwater heater), the feedwater controller failure transient, the control rod withdrawal error transient and the fuel loading error transient.
Initial conditions and transien' input parameters as specified in Sections 6 and 7 of Reference 2 were issumed.
The nonpressurization transients.were analyzed using the methods d gqribed 9
in Reference 4 As per a letter-to all BWR' Licensees from.the-NRCl 1 all operating BWRs reload submittals with General Electric analyses received after February 1,1981 are requested to have the limiting transients
(.overpressurization) recalculated with the ODYN code. The proposed technical specification changes for Cycle 7 will incorporate the require-ments of ODYN (.0ption B) and' new MCPR. limits associated with these new analyses.
The calculated system responses and aCPRs for the above -listed operational transients and conditions have been analyzed by the licensee. Results.
for 100% power /100% flow core conditions were as follows:
-%i y,e
~,
.w
,w'.-
, ~,, -
-y
+,-.--,.-+,m---,,
-,_,,,,,y,,,,,..-w--,-..-.,w,..-+,
..-~:pe--
. i TABLE 1 4
Q/A SL-Pv
. ACPR ACPR
-aCPR Transient Exposure
(% NBR)
(% NBR)
(psig)
-(psig) 7x7-8x8 P8x8R & 8x8R Load Rejection B0C-E0C 501.5
_122.3 1179
~1213
- 0.14 0.19 0.21 w/o Bypass Loss of'100 F 123' 121.7 1022 1069-
-0.12 10.14 0.14 Heater
- Feedwater 314.4 119.1 1135 ~
1172-0.09L0.13 0.14 Feedwater Controller Failure Rod Withdrawal 0.16' O.08 Error 910% RBM
-Set. Point The operating MCPR is dependent on the cycle average scram time (rave) and'the adjusted analysis scram time (TB).
if TB > rave,-the limiting event is a rotated fuel bundle (fuel loading error). The operating-MCPR(s) are obtained from the fuel loading error analysis. However, if rave >rB the limiting e' vent is load rejection without bypass and.the ODYN option B method is_ used to determine the limiting MCPR(s).
Based on the most limiting transient the licensee 'has proposed the-following operating limit MCPRs (OLMCPR's).
TABLE 2 7x7 8x8 8x8R P8x8R Fuel Loading Error 1.23 1.24 1.24 1.24 ODYN OLMCPR's*
T/S Fig T/S Fig T/S Fig T/S Fig 3.ll-2a 3.ll-2h 3.11-2c 3.11-2d
- Determined from ODYN Option E Since the higher OLMCPR obtained from the analyses will preclude violation of the safety limit MCPR of l.07'in the event of any anticipated operational occurrence, we find these limits to be acceptable.
+,.
2.2.3 Overpressure Analyses-For Cycle.7 < the li.censtee has reanalyzed the limiting pressurization event to demonstrate the ASME hoiler and pressure vessel code require-ments are met. The overpressure analysis for the MSIV closure with high-flux scram has been performed in_ accordance with the requirements:of' Reference 7.
The sensitivity of peak vessel pressurp to. failure of-one safety _ valve has also been evaluated. The acceptance criteria-for this event is that the calculated peak. transient' pressure not exceed L110% of design pressure, i.e., 1375 psig. The reanalysis:shows that the: peak pressure at the bottom of the reactor vessel-does not exceed.1290.psig for worst-case end-of-cycle conditions, even when' assuming-the effects of one ! failed -
safety valve. Therefore, the limiting overpressure as analyzed by the licensee is acceptable.
2.2.4 Thermal Hydraulic Stability The results of the thennal' hydraulic stability analysis (Reference'2):
shows that the channel hydrodynamic-and reactor core decay ratios at
+
the natural circulation - 105% rod line intersection (which is the least stable physical attainable point of' operation) are within the stability.
limit. Decay ratio for-Reload 6 was 0.80-as compared to 0.78 for Reload 5.
Since ' operation in the natural circulation mode will be. prohibited -
by the Technical Specifications, there will'be added margin to the stability?
limit and this is acceptable to the staff. This predicted _ decay ratio is below the 1.0 ultimate performance limit decay ratio proposed by General Electric.
~
The staff has expressed generic concerns regarding reactor core thermal-hydraulic stability at the least stable reactor condition.~ This condition
(
could be reached during an operational transient from high power if the.
plant were to sustain a trip of both recirculation pumps without a reactor trip. The concerns are motivated by increasing decay ratios as. equilibrium fuel cycles are approached and as reload fuel designs change. The staff ccacerns relate to both the consequences of operating with a decay ratio-of 1.0 and the capability of the analytical methods to accurately predict decay ratios. The General Electric Company is addressing these staff.
concerns through meetings, topical reports and a stability test program.
It is expected that the test results and data analysis, as presented in a final test report, will aid considerably in resolving the staff concerns.
2.3 Physics Startup Testing L
Several of the key reload safety analysis inputs and results can be =
assured via preoperational testing.
Ir. order to provide this ' assurance, the licensee will perform a s*.,r physics startup tests, which was described in Reference 9.
This test program was submitted previously in connection with the Cycle 5 reload. Our Cycle 5 review found this projram to be acceptable. A written report, describing the results of the physics startup tests, will also be provided by the licensee for staff review following completion of the Cycle 7 tests.
-..__,,_____.__a.__...._,_._
,_.~,,. ___-.
i 6-2.4 Technical Specifications The licensee has submitted proposed changes to the Cooper Technical Specification (31. The effects of these changes are to change the MCPR.
limits to make them consistent with the values presented in Table 2 of this evaluation. Based on the analysis results, these changes to the MCPR limits in the Technical Specifications are found to be acceptable.
~
3.0 ENVIRONMENTAL CONSIDERATION
S We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which.is insignificant from the standpoint of environmental impact, and pursuant to 10 CFR Section 51.5(d)(4) that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of the amendment.
4.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in 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 will be conducted in compliance with the Commission'r regulations and -
the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: May 22, 1981
,,---,-n v.
e-
~,
,,,,,,---w, n
w
REFERENCES
'l.
Nebraska Public Power District letter (J. Pilant) to USNRC (T. Ippolito)
~
' dated March 5,1981.
2.
" Supplemental Reload Licensing Submittal-for Cooper Nuclear' Station Unit 1, Rcload 6," Y1003J01 A17, January 1981.
3.
" Proposed Changes to the Cooper Nuclear Station Technical Specification" appearing as an Enclosure to the NPPD letter to USNRC dated March 5,1981.
4.
" General Electric Boiling Water Reactor Generic Reload Fuel Application,"
NEDE-240ll-P-A July 1979.
5.
General Electric BWR Thermal Analysis Basis (GETAB):
" Data Correlation and Design Application," General Electric Company, BWR Systems Department,.
November 1973 (NEDO-10958).
6.
USNRC letter (D. Eisenhut) to General Electric (P. Gridley) dated May 12, 1978.
7.
USNRC letter (T. Ippolito) to General Electric (R. Gridley) dated April 16 1979.
8.
NRC letter (O. Eis.enhutt to all BWR licensees, dated November 4,1980.
9.
Nebraska Public Power District letter (J. Pilant)' to USNRC (T. Ippolito)-
dated April 16, 1979.
y---
.>,-~--w-
--y, n,-
, -,,,.m--
-p,-g
--e-
,