ML20237K899
| ML20237K899 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/31/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20237K898 | List: |
| References | |
| NUDOCS 8709080108 | |
| Download: ML20237K899 (9) | |
Text
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NUCLEAR REGULATORY COMMISSION p
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO LICENSE AMENDMENT FOR CYCLE 8 (RELOAD 7)
LICENSE NO. NPF-35 BOSTON EDISON COMPANY PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293 1
1.0 INTRODUCTION
4 By letter from R. G. Bird, Boston Edison Company (BECo) to U. S. Nuclear Regulatory Commission dated May 22,1987 (Ref.1), Technical Specification j
changes were proposed for the operation of Pilgrim Nuclear Power Station for Cycle 8 (Reload 7) (designated herein as PSC8) with a reload using General Electric (GE) manufactured fuel assemblies and GE analyses and methodologies.
The requested Technical Specification (TS) changes and reports (including Reference 2) discussing the reload and analyses done to support and justify the Reload 7 operation were included with the submittal. Additional information regarding thennal-hydraulic stability was provided by the licensee in Reference 6.
Some editorial changes to the TS were.also requested as pa.rt of the proposed amendment.
2.0 EVALUATION grR868aBl88$P P
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2.1 RELOAD DESCRIPTION 1
The Pilgrim reload will retain 388 issorted GE P8x8 retrofit fuel assemblies fr0m the previous cycles and add 192 new BP8DRB300 GE fuel assemblies. The reload is based on a previous cycle core-average exposure of 16658 Megawatt Days per Standard Ton (MWD /ST) and a Cycle 8 end of cycle exposure of 18682 MWD /ST. The loading will be a conventional scatter pattern with low reactivity fuel on the periphery.
t.2~ FUEL DESIGN The new fuel assembly to be used for PSC8, type BPRORB300, has been approved for inclusion in NEDE-24011, GESTAR II (Ref. 3). This fuel type has been 1
analyzed for this application with approved methods and meets the approved j
limits of GESTAR II (Ref. 3), therefore the fuel is acceptable for PSC8. The design analysis for the BP80RB300 fuel and TS changes related to MAPLHGR curves for the additional fuel assemblies have been addressed and approved in the NRC Safety Evaluation for Amendment No. 100 to the Pilgrim Facility Operating License (Ref. 4).
2.3 NUCLEAR DESIGN l
The nuclear design for PSC8 has been performed with methodology described in l
GESTAR II (Ref. 3). The results of those analyses are given in Reference 2.
The shutdown margin (SDM) is 3.8 % delta-k/k at the beginning of cycle and 1.0 % delta-k/k at the exposure of minimum shutdown margin. Therefore, it naets the required 0.29 % delta-k/k shutdown margin.
The standby liquid I
control system also meets shutdown requirements with a shutdown marain of 4.4%
delta-k/k.
Since these and other PSC8 nuclear design parameters have been obtained with previously approved methods and fall within expected ranges, the nuclear design is acceptable.
The description of low and low-low water level setpoints with respect to the top of the active fuel are revised to reflect the different dimensional length of the retrofit fuel. The trip setting descriptions which are affected are for trips which close isolation valves in certain process lines which penetrate containment. The water level trip settings are adequate to prevent core uncovery in the case of a break in the largest of these lines assuming a 60 second valve closing time. The required closing tire for the isolation valves in these lines is less than 60 seconds. Therefore the proposed change does not significantly alter the margin to core uncovery and is acceptable.
l 2.4 THERMAL HYDRAULIC DESIGN The thermal-hydraulic design for PSC8 has been performed with the methodology described in GESTAR II (Ref. 3) and the results are given in Reference 2.
The parameters used for the analyses are those approved in Reference 3 for the Pilgrim BWR/3 product line.
The Operating Limit Minimum Critical Power Ratio (0LMCPR) values are determined by analysis of the limiting transients, Rod Withdrawal Error (RWE),
Feedwater Controller Failure (FWCF) and Load Rejection Without Bypass (LRWBP).
The analysis of these events for PSC8, via the 00YN Option A and B approach, provide new Cycle 8 Technical Specification values of OLMCPR as a function of l
i average scram time and exposure ranges. Two exposure ranges from beginning of cycle (B.00) to end of cycle (EOC) were analyzed, 1) BOC to B0C + 7513 MWD /ST and 2)'BOC + 7513 MWD /ST to E0C.
For standard operating conditions, LRWBP is
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controlling at both option A snd B limits.
The resulting OLMCPR values are reflected in the prnposed TS changes. For exposure range 1), the OLMCPR is a constant value of 1.36 for all values of average scram speed; for exposure range 2), the OLMCPR varies from 1.39 to a maximum value of 1.44 as a function of average scram time. Approved methods (Ref. 3) were used to analyze the limiting transients and the results fall within expected ranges and are 1
acceptable.
The thermal-hydraulic stability of the Cycle 8 core has been analyzed using approved methods (Ref. 6). The result is a decay ratio of 0.64 at the intersection of the natural circulation line and the 100 percent rod line.
Existing technical specifications do not allow continued operation in natural circulation. Operation at the combination of low flow and high power sufficient to produce a high decay ratio is thus limited.
Based on the similarity of the Cycle 8 decay ratio to the previously evaluated Cycle 7 l
(reported as 0.65), we conclude that appropriate consideration has been given to compliance with 10 CFR 50 Appendix A General Design Criterion 12 i
(Suppression of reactor power oscillations) and the licensee's submittal is responsive to NRC Generic Letter 86-02 (Ref. 7).
2.5 TRANSIENT AND ACCIDENT ANALYSES The transient and accident analysis methodologies used for PSC8 are described and NRC approval indicated in GESTAR II (Ref. 3).
Generally, the ODYN Option l
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i A and B approach was used for transient analyses.
The Loss of Feedwater Heating event was analyzed with the GE BWP, Simulator code, approved in Reference 3.
The limiting MCPR events have been previously indicated in l
. Section ?.4 of this SE. The core wide transient analysis methodologies are acceptable and the results fall within expected ranges.
The RWE was analyzed with a generic bounding analysis and a rod block setpoint of 1.07 was selected to provide an OLMCPR of 1.29 for the retrofit fuel types.
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l The fuel assembly disorientation event was analyzed with standard methods for l
the PSC8 D lattice fuel, giving a non-limiting MCPR of 1.P6. As approved in 1
Reference 3, the mislocated' assembly is not analyzed for reload cores on the basis of studies indicatina the small probability of an event exceeding MCPR limits.
The limiting pressurization event, the Main Steam Isolation Valve Closure with Flux Scram, was analyzed with standard GESTAR II methods. Results for peak steam dome and vessel pressures were well under required limits. These are acceptable methodologies and results.
i LOCA analyses, using approved methodologies and parameters (Ref. 3), were perfomed to provide MAPLHGR values for the new reload fuel assemblies (BP8DRB300). These analyses.and results are acceptable. The resulting LOCA results and MAPLHGR values are documented in Attachment 4 (Addendum to Ref.
- 5) of the base Reload 7 submittal (Pef.1).
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2.6 TECHNICAL SPECIFICATIONS The Technical Specification changes are for the most part minor and provide for MCPR changes due to Cycle 8 parameter changes, MAPLHGR values for the new fuel assemblies, and editorial changes.
Details of the specification changes follow:
(1) Specification 2.1, LC0 pages 203, 205A-1, 2058-2, 205C-2, and Major Design Features page 206m References to non-retrofit fuel are removed since this fuel type is not to be used in Cycle 8.
These changes are acceptable.
(2) Pages 46, 53, 59a and 68 I
Descriptions of water level setpoints are changed to reflect dimensional changes for retrofit fuel assemblies. The changes are acceptable as discussed in Section 2.3.
(3) Table 3.11-1, page 2CSB-2 and Bases page 205C-3 Operating limit MCPR values are revised for Cycle 8 operation. The revised information is acceptable as discussed in Sections 2.4 and 2.5.
(4) Figures 3.11-1 through 3.11-7:
MAPLHGR versus Planar Average Exposure Curves
9 A new MAPLHGR curve is provided for the new fuel. A standard unit of fuel exoosure, megawatt days per standard ton (MWD /ST), was established as' uniform nomenclature. -These changes are acceptable as discussed in Section 2.5.
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- ENVIRONMENTAL CONSIDERATION 1
l This amendment involves a change in the installation 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
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may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation. exposure.
The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. 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 impact statement or environmental assessment need i
be prepared in connection with the issuance of this amendment.
3.0 CONCLUSION
As a result of our review, which is described in Section P.0 of this evaluation, we conclude that the proposed reload and technical specification changes are acceptable.
The staff has concluded, based on the considerations discussed above, tha.t (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the j
issuance of this amendment will not be inimical to the health and safety of the public.
r tate:
August 31, 1987 l
Principal Contributor: Michael McCoy l
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4.0 REFERENCES
i 1.
Letter from R. G. Bird, BECo, to US NRC, dated May 22, 1987 " Reload 7 Licensing Submittal and Proposed Change to Technical Specifications (with Attachments) 2.
GE Report 23A4800, dated December 1986,
" Supplemental Reload Licensing j
Submittal for Pilgrim Nuclear Power Station, Reload 7" 3.
NEDE-24011-P-A-8, May 1986, " General Electric Standard Application for i
Reactor Fuel,"
(GESTAR II) 4.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.100 to Facility Operating License No. DPR-35, 'ilgrim Nuclear Power Station, April 9,1987 5.
NEDO-30767, September 1984, " Pilgrim Nuclear Power Station Loss of Coolant Accident (LOCA) Analysis Update" 6.
Letter, R. G. ' Bird BEco, to U. S. NRC, dated July 22, 1987 "Results of Thermal-Hydraulic Stability Analysis for Cycle 8 Operation".
7.
Technical Resolution of Generic Issue B Thermal Hydraulic Stability (Generic Letter NO. 86-02), dated January 23, 1986.
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