ML20071J004

Safety Evaluation Supporting Amend 54 to License DPR-35

ML20071J004
Person / Time
Site:	Pilgrim
Issue date:	03/20/1982
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8204260050
Download: ML20071J004 (13)

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o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 54 TO FACILITY LICENSE NO. DPR-35

,B0STON EDISON COMPANY PILGRIM NUCLEAR POWER STATION DOCKET N0. 50-293 Principal Authors:

B. Hardin, S. Sun, J. Voglewede and K. Eccleston A.

Pilgrim Nuclear Power Station, Reload 5 Application I.

Introduction By letter dated September 22,1981 (Reference 1), Boston Edison Company (licensee) provided their Reload 5 submittal including requested technical specification changes for the Pilgrim Nuclear Power Station.

The proposed technical specification changes are designed to allow operation during Cycle 6 and address (1) a reduction in Maximum Average Planar Linear Heat Generation Rate (MAPLHGR). values to account for taking no credit for core spray heat transfer during a LOCA (due to assuming degraded core spray perfonnance from spray sparger cracks), and (2) re. vised operating limit Minimum Critical Power Ratios (MCPRs) to allow variation with measured scram times. The revised MCPR values were determined-from

_ _ _.0DYN transient analysis code calculations for the Cycle 6. fuel loading. _

II. Evaluation 1.

Effect of Core Sprav Sparger Cracks on ECCS Performance As a result of the discovery in 1980 of crack-like indications on the core spray spargers inside the reactor vessel, the operating limit MAPLHGR values were modified by the application of a set of reduction factors to ensure that the LOCA limits defined in 32 s. eg i, n.

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2 10 CFR 50.46 would be met. The indications were discovered during the ten year inservice inspection performed in February 1980.

Safety Evaluation for Amendment 42 for Pilgrim 1 addressed the potential effects of sparger cracks and documented the NRC finding of the acceptability of operation during Cycle 5.

At a meeting held in Bethesda on October 14, 1981, the licensee presented results from inspections of the core spray piping and spargers made during the recent refueling outage.

The most recent inspections were made using improved remote visual inspection techniques and showed that most of the visual indications noted in 1980 were not actually cracks.

Indications of cracks in one remaining area indicate no change in size since the 1980 inspection.

In spite of this most recent evidence supporting the continued integrity of the core spray system, the lead time required for scheduling LOCA (Appendix K) calculations dictated that assumptions regarding the effectiveness of the core spray system be nade prior to making these most recent inspections of the core spray spargers. To be conservative, the licensee chose to have the LOCA calculations for Cycle 6 performed assuming no credit for core spray heat transfer. The results of these calculations were submitted in References 2 and 8 and show that the acceptance criteria of 10 CFR 50.46 are satisfied if MAPLHGR reduction factors are applied for each fuel type for Cycle 6.

The results of the worst case (DBA)

LOCA calculations are summarized-in Table 1 of this report.

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3 is our understanding that the licensee may submit revised LOCA calculations for Cycle 6 assuming credit for core spray heat transfer at some future date based on the results of the more recent core spray sparger inspection.) The MAPLHGR reduction factors are independent of fuel exposure and are listed in Table 2.

The reduction factor (multiplier) for each fuel type is applied to the exposure dependent MAPLHGR values determined from calculations in which credit was taken for core spray heat transfer. The revised LOCA calculations were performed using the standard, approved 10 CFR 50 Appendix K analysis methods with the following exceptions:

(1) Core spray coefficients for the hot assembly were set equal to zero, and (2) Credit was taken for a heat transfer coefficient of 25 BTU /hr-ft _oF applied to the outside of the channel starting at 2

the time when the water level in the bypass (space between the channels) was calculated to reach the elevation of the hot node.

Credit is not normally taken for outside channel cooling but was needed in the current calculations in which core spray cooling is not assumed. While we agree that the value chosen is conservative, the time at which it is applied may not be conservative due to the existence of a time delay in reflooding of the hot node caused by the lateral flow resistance of the fuel bundles.

In addition to taking no credit for core spray heat transfer, the licensee has stated that there are other conservatisms in the calculations for heat transfer in the hot essembly.

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Breakdown of upper tie plate counter current flow will cause reflood earlier than now predicted.

Side entry orif'ces and bypass leakage will improve heat removal from the hot assembly, and the correlatio.1s used for decay heat and film boiling are conservative.

We have reviewed this information and conclude that there is sufficient conservatism in the present calculations when considered with the recent evidence of core spray sparger integrity to justify operation using the derived restricted MAPLHGR values until the end of Cycle 6.

If af ter Cycle 6, credit is still not taken for core spray heat transfer for licensing analyses of a LOCA, we will require that the licensee provide further justification for the time at which outside channel cooling is assumed to commence.

Based on our review, we conclude that the LOCA calculations submitted for Cycle 6 operation resulting in reduced MAPLHGR values satisfy the criteria of 10 CFR 50.46 and are therefore acceptable.

Figure 1 shows the MAPLHGR values plotted vs. planar average exposure (Reference 3).

Also, we have reviewed the licensee's proposed changes to Technical Specification 3.llA (Pages 205C; 205C-2; 205D; and 205E-6, Fig. 3.11-6) involving the reduced MAPLHGR values and conclude that those changes are in accordance with these calculations. Consequently, we find them acceptable.

2.

Cycle 6 Transient Analyses Generic information relative to the reload analyses of BWR fuel is presented in the NRC reviewed and approved General Electric Licensing i

Topical Report NEDE-240ll-P-A, " Generic Reload Fuel Application," July 1979 (Reference 4). This report is supplemented by plant-specific e

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information contained in Reference 5.

Together, these two documents provide the bases for the licensee's safety analysis and safety evaluation for Reload 5, and the proposed Technical Specification changes associated with the reload for Cycle 6.

The safety limit MCPR has been imposed to assure that at least 99.9 percent of the fuel rods in the core are not expected to experience boiling transition during the most moderate frequency transient events. As stated in Reference 4, the safety limit MCPR is 1.07.

There has been no change in the safety limit MCPR for Pilgrim from Cycle 5 to Cycle 6.

The licensee stated (Reference 1) that all transients that are the basis of the Pilgrim license were reviewed for Cycle 6 and that those transients that were critical with respect to safety marQins and sensitive to the core reload parameter changes were reanalyzed. The ODYN code was used in the determination of Critical Power Ratios (CPRs) for the raoid pressurization transients. The REDY code was used for the slower, non-pressurization events. The licensee provided both tabular and graphical results of their transient calculations for Cycle 6 in Reference 5.

A summary of the calculated Cycle 6 MCPRs is given in Table 3 of this report. The most restrictive condition was calculated to occur as a result of a postulated Generator Load Rejection without bypass. For this event a MCPR of 1.12 was predicted which is still above the MCPR safety limit of 1.07.

This is acceptable.

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6 Reactor vessel overpressure protection was verified by an ODYN analysis of the closure of all main steam line isolation valves (MSIV) with an indirect (flux) scram. MSIV closure is the limiting -

event with regard to vessel overpressurization. At the end of Cycle 6 with all safety relief valves operating and an indirect scram, the peak vessel pressure was predicted by this ODYt! analysis to be 45 psi below the peak allowable ASME overpressure of 1375 psig at the vessel bottom. This is acceptable.

3.

Use of ODYN Option 8 MCPR Scram Insertion Time Conformance Procedure The operating MCPR Technical Specification has been revised to the ODYN " Option 3" format where the Operating Limit Minimum Critical Power Ratio (0LMCPR) varies with measured scram times. The specification is based on measurements to the 30% inserted position which was chosen to coincide with present surveillance procedures at Pilgrim. The numerical values of the MCPRs are based on the CPRs from the Load-Rejection-without Bypass transient given in Reference 5.

The analysis of this transient was done using Technical Specification scram times but the uncertainty penalty applied to the nominal results was based on sensitivity studies done by GE using a generic population of scram speed data. The proposed Technical Specification changes include a verification that Pilgrim is not outside this population, or if it is, the OLMCPR linearly approaches a conservative value of 1.40 for P8 x 8R fuel or 1.37 for 8x8 fuel. Operation within the proposed limit will avoid violation of the Safety Limit MCPR at any time during Cycle 6.

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7 The staff has reviewed the General Electric (GE) generic option C scram time specification procedure using ODYN and has found it to be acceptable (Reference 6 and 7). The licensee has duplicated these procedures and, therefore on this basis, we conclude that this is acceptable. We have also reviewed the licensee's proposed changes to Technical Specification 3.11C related to MCpR (pages 205B, 2058-1, 2058-2, 205C-2, 205C-3, and 205D) and conclude that those changes are in agreement with the option B scram time specification procedure using ODYN.

Consequently, we find them acceptable.

4.

Thermal-Hydraulic Stability The results of the thennal-hydraulic analysis (Reference 5) show that the maximum thermal hydraulic stability decay ratio is 0.59 for Cycle 6 as compared to 0.61 for Cycle 5.

Because operation in the natural circulation mode will be prohibited by Technical Specification 3.6.A.6, there will be added nargin to the core stability and this is acceptable to the staff.

5.

Fuel Syste:n Design In References 8 and 9 GE requested that credit for calculated peak cladding temperature margin as well as credit for recently approved, but unapplied, ECCS evaluation model changes be used to offset any operating penalties due to high burnup fission gas release.

We reviewed and found the proposal acceptable (Reference 10) provided the generic analysis was found to be applicable to each plant relying on the GE position.

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8 In a letter dated December 2,1981 (Reference 11), the licensee stated that the generic analysis is applicable to Pilgrim.

On this basis, we find the revisions to the MAPLHGR limits given in Attachment C of Reference 5 acceptabl'e.

III. Summary He have reviewed the licensee's proposed Technical Specification changes and supporting information for Cycle 6 operatioi. involving Specifications 3.11. A and 3.ll.C and conclude that they are acceptable.

IV.

Environmental Considerations We have determined that the amendment does not involve 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 Sl.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.

V.

Conclusions 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

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a does not involve a significant hazards consideration, -(2) there is 1-reasonable assurance that, the health and safety of the public will not

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'will_be conducted in compliance with the Commission's regulations and the issuarice of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Dated: March 20,1982.

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Table 1 LOCA Analysis Results for Design Basis Accidenk Calculated Allowabib}

Peak Cladding Temperature (*F) 2200 F 2200 F Local Cladding 0xidation 2.5%

17.0%

Core Wide hydronen Generation 0.17%

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.(1) Break size for Design Basis Accident = 4.343 Ft (2) Acceptance Criteria from 10 CFR 50.46

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' b Table 2 MAPLHGR Multipliers Assuming No Core Spray Heat Transfer Credit Fuel Type Core Flow > 90% Rated -

Core Flow < 90% Rated

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0.93 0.85 8DB219H 0.93 0.85 8DB262 0.94 0.86 P80RB265L 0.91 0.84 P8DRB282 0.92 S.85 P80RB265H 0.90 0.82 1

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Table 3 i

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Pressurization Events 8x8 Fuel P8x8R Fuel Generator Load Rejection 1.12 1.12 No Bypass Feedwater Controller 1.20 1.22 Failure Non-Pressurization Events Loss of 100*F Feedwater Heating 1.23 1.25 Fuel Loading Error Not Applicable 1.23 i

Rod Withdrawal Error 1.15 1.18 t

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References 1.

J. E. Howard (30ston Edison) to T. A. Ippolito (USNRC), " Reload 5 submittal and Request for Technical Specification changes," BE Co.

Ltr.81-222, September 22, 1981.

2.

Revision I to " Loss of Coolant Accident Analysis Report for Pilgrim Nuclear Power Station," NED0-21696, August 1981.

3.

L. M. Fulton (Coston Edison) to T. A. Ippolito, " Corrected Page for Reload 5 Submittal," BE Co. Ltr.81-237, October 5,1981.

4.

" Generic Reload Fuel Application," NEDE-240ll-P-A,. July 1979 as amended.

5.

"Suplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station Unit 1 Reload 5,"

Y1003J1 A28, August 1981.

l 6.

" Safety Evaluation for the General Electric Topical Report Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors NED0-24154 and NEDE-24154-P Volumes I, II, and III," June 1980.

7.

" Supplemental Safety Evaluation for the General Electric Topical Report Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors NE00-24154 and NEDE-24154P Volumes I, II, and III," January 1981.

8.

R. E. Engel (GE) letter to T. A. Ippolito (NRC) dated May 6,1981.

9.

R. E. Engel (GE) letter to T. A. Ippolito (NRC) dated May 28, 1981.

10.

L. S. Rubenstein (NRC) memorandum for T. M. Novak (NRC) on " Extension of General Electric Emergency Core Cooling System Performance Limits,"

dated June 25, 1981.

11.

A. V. Morisi (Boston Edison) to T. A. Ippolito (USNRC), " Additional Information on Reload Application," BECo Ltr. #81-277, December 2,1981.

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