ML20211P141
| ML20211P141 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 02/19/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211P120 | List: |
| References | |
| NUDOCS 8703020302 | |
| Download: ML20211P141 (9) | |
Text
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o UNITED STATES
[1 3 m [,h NUCLEAR REGULATORY COMMISSION n
7-l WASHINoTON, D. C.'20555
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- SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i
SUPP0kTING AMENDMENT NO. 89 TO FACILITY OPERATING LICENSE NO. DPR-6 CONSUMERS POWER COMPANY BIG ROCK POINT PLANT DOCKET NO. 50-155
- 1. 0 INTRODUCTION By letter dated December 5, 1986 (Ref. 1) from K. W. Berry, Consumers Power Company (CPC), to the Director of NRR, NRC, CPC proposed changes to the Big Rock Point Plant Technical Specifications (TS) to incorporate a new design of hafnium hybrid control blades and revise the core operating limits Tables 1 and 2 of TS Section 5.2.1 to include Reload I-2 fuel for Cycle 22 operations.
In a separate submittal dated December 29, 1986 (Ref. 2), from Ralph R. Frisch, CPC, to the Director of NRR, NRC, the licensee provided documentation regarding fuel performance in recent operating cycles and fuel inspection plans for the Cycle 22 reload fuel.
This information was provided in response to concerns identified by the NRC Region III inspectors and the staff in several conference calls regarding CPC policy for identification and removal of known leaking fuel pins from a new reload core.
During the staff's preliminary review of Reference 1, the staff informed the licensee of a need for more detailed information regarding the hafnium hybrid control rod design and for additional information regarding the Cycle 22 reload and safety analysis on core operating limits.
The licensee responded with submittals dated January 20, 1987 (Ref. 3) and January 30, 1987 from Ralph R.
Frisch, CPC to NRC (Ref. 4).
Subsequently, during a February 2, 1987 telephone conversation with CPC personnel, the staff informed the licensee that a denial safety evaluation (SE) (Ref. 7) rejecting the proposed TS changes to core operating limits tables to accommodate the Cycle 22 reload was being issued.
In response to the NRC denial and a request for additional information (Ref. 5),
the licensee submitted modified proposed TS changes for the Cycle 22 operations i
(Ref. 6).
This report describes the staff review of References 2 and 6 and provides the staff's SE approving the modified proposed TS changes for the Cycle 22 reload and operations at the Big Rock Point Plant.
The staff approved the proposed changes to the TS addressing the use of the hafnium hybrid control blades in j
a separate SE (Ref. 8).
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2.0 BACKGROUND
i 2.1 CoreDkscription Big' Rock Point is a BWR-1 rated at 240 MWt or about 72 MWe, but has been 4
restricted to a lower power _ level because of restrictive core thermal limits-during much of its operating history.
Detailed descriptions of the core and its unique features and the Cycle 22 reload are given in References 2 and 3.
2.2 Fuel Performance Big Rock Point has experienced evidence of failed fuel in all operating cycles since Cycle 18.
During Cycle 19, the off gas release rate increased from an initial 300 pCi/sec'of noble gas to 30,000 pCi/sec within a period of ten months (Ref. 2).
The reactor. coolant radioiodine_ activity increased by a factor of 10 to 100, but was still well below the TS action level. The entire core was sipped during shutdown but only three leaking bundles were identified even though the off gas level was indicative of about 30 leaking fuel pins.
During this cycle plant radiation levels increased only around the piping and components associated with the off gas system. The failures were attributed primarily to'a bad ingot of Zircaloy fuel. cladding material which was susceptible to corrosion, i
Cycle 20 operated with an off gas release rate ranging from 300 pCi/sec initially to a final value of about 4000 pCi/sec.
Five leaking fuel pins were identified by sipping the entire core after Cycle 20 operation.
None of the identified leakers were returned to the core.
During Cycle 21, the off gas release rate ranged from an initial value of 600 pCi/sec to a final value of about 3000 pCi/sec.
The licensee concluded, based on the information available, that there is reasonable assurance that the two or three leaking fuel pins were among the fuel scheduled for discharge after Cycle 21.
A decision was made to not schedule any fuel sipping for the pre-Cycle 22 refueling outage.
This decision was justified (Ref. 9) based on the following:
(1) Experience from past operating cycles indicates that the failed fuel pins are in the highest exposure fuel.
This failed fuel is scheduled to be either discharged or located in lower power perimeter regions of the core.
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(2) The licensee concludes that the dose to on-site personnel from sipping operations would be worse than the dose to the total population during Cycle 22 operation if the leakers are returned to the core.
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(3) Economic considerations.
i (4) Ablimitedultrasonictestinspectionscheduledforhigherexposure fuel being returned to the core may provide indication of leakers.
The licensee stated that it was not its policy to install known leaking fuel pins into a reload core.
The ultrasonic test inspection program has been performed by EXXON Nuclear Company for a total of 21 assemblies being returned to the reactor core (Ref.
3).
The numbers and types of fuel tested are:
2 H-1 types 2 H-2 types 16 H-3 types 1 H-4 type No leaking assemblies were detected.
- 3. 0 PROPOSED CHANGES TO THE BRP TECHNICAL SPECIFICATIONS In its submittal of February 4, 1987 (Ref. 6), CPC proposed changes to the TS contained in the Facility Operating License DPR-6, Docket No. 50-155, issued to CPC on May 1, 1964, for the Big Rock Point Plant as follows:
A.
Add to Section 5.2.1(b), Table 1:
" Reload"
" Reload" "12" to the last column head Il so that it will read 11/12.
B.
Add to Section 5.2.1(b), Table 1:
Footnote "(1)" to Minimum Critical Power Ratio at Normal Operation Conditions, Maximum Heat Flux at Overpower, Maximum Steady-State Heat Flux, Maximum Average Planar Linear Heat Generation Rate, Steady-State and Maximum Bundle Power.
Footnote "(2)" to Maximum Steady-State Power Level.
C.
Add to Section 5.2.1(b), Table 1, the following footnotes:
" Note 1:
Actual Calculated Critical Power Ratio, Heat Flux, Average Planar Linear Heat Generation Rate, and Bundle Power values used to monitor conformance to the associated thermal-hydraulic limits shall include the uncertainties listed in Table 3 and be bounded by the Technical Specification limits found in Table 1.
4-Note 2:
Maximum Steady-State Power level shall not exceed a i
b' power such that any thermal-hydraulic limit listed in Table 1, Table 2, Figure 1, or Figure 2 is exceeded."
As agreed upon with the licensee, the wording of Note 1 has been changed slightly from the wording in the licensee's amendment request to provide clarification.
D.
Add to Section 5.2.1(b), Table 2, the following footnote and add an asterisk after LIMITS in the title:
"*MAPLHGR ratios shall include the uncertainty listed on Table 3."
E.
Add to Section 5.2.1(b), Table 2:
" Reload"
" Reload" "I2" to the last column heading Il so that it will read 11/I2.
F.
Add to Section 5.2.1(b) the following table after Table 2:
" TABLE 3 UNCERTAINTY FACTORS APPLIED TO THERMAL-HYDRAULIC PARAMETERS Parameter Uncertainty Factor MCPR
.1531 Heat Flux 9.4017%
MAPLHGR Ratio
.0784 Actual Cundle Power / Max Allowed Bundle Power 3.7661%"
G.
Add to Section 5.2.1(b), Figure 1, the following footnote and add an asterisk after RELATION in the title:
"*The maximum bundle peak linear heat generation rates shall include uncertainties listed for MAPLHGR in Table 3."
H.
Add to Section 5.2.1(b), Figure 2, the following footnote and add an asterisk after MWt in the title:
"* Maximum bundle power ratios shall include the uncertainty listed in Table 3."
4.0 EVALUATION The staff review and evaluation of the fuel design, nuclear design, thermal-hydraulic design and transient and accident analyses indicate that the TS limits listed in Section 5.2.1(*), along with the associated parameter uncertainties for the limits of Minimum Critical Power Ratio (MCPR), Heat
_5-o Flux, Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Ratio, I
and Actual pundle Power / Max Allowable Bundle Power, were documented in the staff-approked Big Rock Point " Physics Methodology Report" Revision 3 (Ref. 9).
Furthermore, the core operating limits specified for the Cycle 21 reload I-1 fuel assemblies are also applicable to the EXXON I-2 fuel assemblies for the Cycle 22 reload since they are identified except for the burnup during Cycle
- 21. The proposed TS change is an administrative change which the staff finds to be acceptable and appropriate.
To provide a constant assurance of compliance to the TS thermal-hydraulic limits, the plant reactor engineering group analyzes each notch to be withdrawn before it is released for actual withdrawal.
This analysis is based upon updated and current fuel exposure, void fraction history, power history, control rod depletion, and control rod patterns.
Empirical verification for the accuracy of the code used for these calculations is assured through periodic comparison to flux shape measurements. The thermal-hydraulic properties, with associated uncertainties, are compared to and must meet the TS requirements before the analyzed notch is released to operators.
This notch by notch analysis is initiated at the beginning of each cycle during power escalation before thermal-hydraulic parameter to TS margins are less than 20% and this analysis continues throughout the cycle.
The parameter uncertainties are applied to associated operating limits as described below:
Minimum Critical Power Ratio The Critical Power Ratio is expressed as the ratio of Critical Power Level and Actual Power Level.
This ratio for operating conditions must satisfy the following relationship:
Calculated Critical Power Ratio - Uncertainty > MCPR Limit OR Calculated Critical Power Ratio
.1531 > 1.59 (1.61 for I feel) OR Calculated Critical Power Ratio must be greater than 1.743 (1.763 for I fuel)
Heat Flux Heat Flux must satisfy the following relationship:
Calculated Heat Flux + Uncertainty < Maximum Heat Flux Allowed per Technical Specification OR 2
Calculated Heat Flux + UncertaTnty < 322,100 BTU /HR - FT OR CalculatedHeatFlux(1+9.4017%)<322,1g0 BTU /HR-FT OR Calculated Heat Flux < 294,419 BTU /HR - FT
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MAPLHGR i
TherakioofactualMaximumAveragePlanarLinearHeatGeneration Rate to allowed Maximum Average Planar Linear Heat Generation Rate must satisfy the following relationship:
+ Uncertainty < 1; Uncertainty =.0784 OR g
MAPLHGR Ratio < 92.16%
Bundle Power The ratio of actual calculated bundle power to the allowed bundle power of 5.2.1(b) Figure 2 must satisfy the following relationship:
Actual Bundle Power Allowed Bundle Power + Uncertainty < 1 OR Bundle Power Ratio < 96.23%
The proposed TS changes incorporate the uncertainty factors approved for the safety analyses methodology into the TS definition of core operating limits.
Since the safety analyses, the TS and the operational control to assure conformance with the TS operating limits are now consistent and enforceable, the staff finds these changes acceptable.
5.0 EMERGENCY CIRCUMSTANCES On December 5, 1986, the licensee submitted a TS change request to reflect the use of hafnium hybrid control rods and new Reload I-2 fuel in the upcoming cycle (Cycle 22).
On January 2,1987, the Big Rock Point Plant entered into a refueling outage to install the hybrid control rods and Reload I-2 fuel.
On January 20, 1987, in response to the staff's request, the licensee submitted supplemental information concerning their reload evaluation for Reload I-2 fuel and a more detailed description of their hybrid control rods.
After reviewing this supplemental information, the staff determined that the licensee's proposed changes adopting existing core operating parameters to the Reload I-2 fuol were unacceptable.
During a February 2, 1987 telephone conversation, the staff notified the licensee that it was denying the portion of their December 5, 1986 application for amendment which adopted existing core operating parameters for Reload I-2 fuel.
The licensee responded to this denial on February 4,1987, submitting a revised TS change request for core operating parameters for Reload I-2 fuel.
The staff has found this new change amendment acceptable.
The Big Rock Point Plant is scheduled to restart from the current refueling outage on February 23, 1987.
However, the plant cannot restart without approval of the February 4, 1987, TS change request.
The staff has determined that this constitutes an emergency situation since resumption of power would be precluded if this change request is not approved.
_7-5.1 No Significant Hazards Consideration Determination i
TheCommisskonhasprovidedstandardsfordeterminingwhetherasignificant hazards consideration exists, as stated in 10 CFR 50.92(c).
This regulation states that a license amendment involves no significant hazards consideration if operation of the facility in accordance with the amendment would not:
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any I
accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
The licensee has performed an evaluation using these criteria and has applied them to the proposed TS changes.
A summary of the license's evaluation is presented below.
The licensee has submitted this TS change request in response to the staff's denial of their previous TS change request (December 5, 1986) which-proposed adopting existing core operating parameters to the Big Rock Point Reload I-2 fuel.
This proposed change request adds a table of uncertainty factors associated with the analysis of the core thermal-hydraulic parameters heat flux, MCPR, MAPLHGR, and maximum bundle power.
Footnotes are also provided delineating the manner in which these uncertainty values are to be used in establishing conformance with associated thermal-hydraulic limits.
These uncertainty factors were previously derived as a part of the Big Rock Point Physics Methodology Topical Report, Revision 3 approved by NRC letters of September 16, 1982 and February 9, 1983.
The mechanical, thermal-hydraulic, and neutronic analysis for Big Rock Point Reload I-2 is the same as that for Reload I-1.
The design report previously issued for Big Rock Point Reload I-1 (Exxon Nuclear Company (ENC) report XN-NF-85-38(P), Rev 0) entitled, " Design Report for Big Rock Point I-1" is applicable for Reload I-2.
This reload does not contain any fuel assemblies significantly different from those previously found acceptable by the NRC.
This change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the thermal-hydraulic i
limits are derived in a manner identical to that described in Exxon Nuclear Corporation (ENC) report XN-NF-79-32, Revision 1, entitled, " Big Rock Point LOCA Analysis Using the ENC WREM NJP-BWR ECCS Evaluation Model - MAPLHGR Analysis".
This change does not create the possibility of a new or different kind of accident from any accident previously evaluated because the change does not represent a departure from Big Rock Point's current methodology of calculating thermal-hydraulic limits.
It merely applies these previously approved methods to core characteristics for the next operating cycle. This change incorporates i
> j within the TS uncertainty values (associated with the thermal-hydraulic limits) i which are dgcumented in t5e NRC approved Big Rock Point " Physics Methodology Report", Refision 3.
The limits observed in operating the core have been inclusive of these uncertainties.
4 This change does not represent a change in the margin of safety since the required spectrum of break locations, sizes, and configurations for the Big Rock Point Plant (contained in the ENC report XN-NF-79-21) do not change.
The addition of uncertainty factors and footnotes to the TS only clarifies Big Rock Point Methodology for controlling operating parameters.
The staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis.
Therefore, the staff has determined that the application for amendment involves no significant hazards consideration.
5.2 State Consultation Consultation was held with the State of Michigan by telephone.
The State expressed no concern either from the standpoint of safety or of no significant hazards consideration determination.
6.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the installa-tion or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has made a final no significant hazards consideration finding with respect to this amendment.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental 'mpact statement nor environmental assessment need be prepared in connection with the issuance of this amendment.
- 7. 0 CONCLUSIONS The staff has reviewed the TS change request and other information, including fuel integrity considerations, submitted by CPC for the Cycle 22 reload.
The staff concludes that approved methods were used to perform the design and analysis of the Big Rock Point reactor for Cycle 22 operation and that the
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proposed TS changes are consistent with the safety analyses and its implementa-tion during plant operation, and are therefore acceptable.
Likewise, the staff has concluded that the licensee has taken reasonable measures to assure the integrity of fuel to be included in the reload core for operating Cycle 22, and the staff finds this acceptable.
_g.,
The staff has concluded, based on the considerations discussed above that:
i (1) the amepdment does not (a) significantly increase the probability or consequencet~of an accident previously evaluated, (b) increase the possibility of a new or different kind of accident from any previously evaluated or (c) significantly reduce a safety margin and, therefore, the amendment does not involve 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 i
in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and the security or to the health and safety of the public.
8.0 REFERENCES
1.
Letter from K. W. Berry, Consumers Power Company (CPC), to Director, Nuclear Reactor Regulation, NRC, dated December 5, 1986.
2.
Letter from R. R. Frisch, CPC, to Director, Nuclear Reactor Regulation, NRC, dated December 29, 1986.
3.
Letter from R. R. Frisch, CPC, to Nuclear Regulatory Commission Document Control Desk, dated January 20, 1987.
4.
Letter from Hoffman/Blanchard, CPC to NRC, " BIG ROCK TELECOPY 2," dated January 30, 1987.
5.
Letter from J. A. Zwolinski, NRC, to K. W. Berry, CPC, " Request for Additional Information - Technical Specification Changes Pertaining to Cycle 22 Reload and Hybrid Control Blades Installation," dated January 30, 1987.
6.
Letter from K. W. Berry, CPC, to Nuclear Regulatory Commission Document Control Desk, dated February 4, 1987.
7.
Letter from J. Zwolinski, NRC, to K. W. Berry, CPC, " Proposed License Amendment - Denial of Reload I-2 Fuel Portion of Technical Specification Change Request for Operating Cycle 22 (TAC 64593)," dated February 12, 1987.
8.
Letter from J. Zwolinski, NRC, to K. W. Berry, CPC, " Proposed License Amendment - Approval of Control Rod Portion of Technical Specification Change Request for Operating Cycle 22 (TAC 64087)," dated February 17, 1987.
9.
Letter from D. M. Crutchfield, NRC, to D. J. VandeWalle, CPC, Revision 3 -
PHYSICS METHODOLOGY REPORT, Big Rock Point, dated February 9,1983.
PRINCIPAL CONTRIBUTOR:
U. Cheh and C. Hinson Dated February 19,1987