BECO-87-081, Application to Amend License DPR-35,consisting of Proposed Change 87-03,ensuring Safe Operation of Reload 7 Core Design During Cycle 8.Supporting Matl Encl.Fee Paid

Application to Amend License DPR-35,consisting of Proposed Change 87-03,ensuring Safe Operation of Reload 7 Core Design During Cycle 8.Supporting Matl Encl.Fee Paid

ML20214Q099
Person / Time
Site:	Pilgrim
Issue date:	05/22/1987
From:	Bird R BOSTON EDISON CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20214Q102	List: BECO-87-081, Application to Amend License DPR-35,consisting of Proposed Change 87-03,ensuring Safe Operation of Reload 7 Core Design During Cycle 8.Supporting Matl Encl.Fee Paid ML20214Q111 ML20214Q147 ML20214Q182 ML20214Q200 ML20214Q222
References
BECO-87-081, BECO-87-81, NUDOCS 8706040233
Download: ML20214Q099 (5)

Text

r i 10CFR50.90 1

$1 aosawasou 1 Executive Offices l 800 Boyiston Street Boston, Massachusetts 02199 BECo 87 081 Ralph G. Bird Proposed Change 87-03' ,

senior vice President - Nuclear May 22, 1987 U. S. Nuclear Regulatory Commission Document Control Desk Hashington, DC 20555 License DPR-35 l Docket 50-293 Reload 7 Licensina Submittal and Procosed Chanae to Technical Soecifications

Dear Sir:

Pursuant to 10CFR50.90, Boston Edison Company proposes the modifications to Appendix A of Operating License DPR-35 provided in Attachment 1. These proposed modifications to the Pilgrim Nuclear Power Station Technical Specifications will ensure safe operation of the Reload 7 core design during the next operating cycle.

These proposed changes must be approved prior to restart from Refueling Outage 7.

A transient and safety analysis unique to Reload 7 was performed as reported in the General Electric Supplemental Reload Licensing Submittal in Attachment 2. This report provides plant-specific information to supplement Topical Report NEDE-240ll-P-A-8, General Electric Standard Application for Reactor Fuel (GESTAR-II), dated May 1986. Together, these documents provide the justification for safe operation of Pilgrim during Cycle 8.

l l The Topical Report NED0-21696, Loss-of-Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, dated August 1977, has been updated and amended as described in Attachments 3 through 5 to examine the impact of several plant modifications l affecting emergency core cooling system (ECCS) performance, new knowledge concerning ECCS behavior, and the use of new fuel type BP80RB300 in the Reload 7 core.

In accordance with 10CFR170.12(c), a fee of one hundred and fifty dollars ($150.00) will be electronically transferred to your offices. I DMV/amm hd R. G. Bird '

I Attachments: See next page One original and 37 copies ~~

8706040233 870522 ..

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Commonwealth of Massachusetts) 0500g3 ,. ,

County of Suffolk ) .M,g q, d (g.y 4 :-

Then personally appeared before me, Ralph G. Bird, who, being duly sworn, d a 4,! N thatheisSeniorVicePresident-NuclearofBostonEdisonCompanyandthat{h ,y <-

duly authorized to execute and file the submittal contained herein in the nam'esan / ,

on behalf of Boston Edison Company and that the statements in said submittal aFe ?

true to the best of his knowledge and belief.

My commission expires: 3.em/ .3,/t'V2

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DATt NOTARY PUBLIC Nersy Pal;;

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• BOSTON EDISON COMPANY Page 2 Attachments:

1) Proposed Technical Specification Changes for Reload 7.

2) General Electric Boiling Water Reactor Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station, Reload 7, 23A4800, Revision 0, dated December 1986.

3) Pilgrim Nuclear Power Station Loss-of-Coolant Accident (LOCA) Analysis Update, General Electric Topical Report NEDO-30767, dated September 1984.

4) Errata and Addenda Sheet 4, dated August 1986, to Loss-of-Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, NED0-21696, dated August 1977.

5) Errata and Addenda Sheet 5, dated October 1986, to Loss-of-Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, ??EDO-21696, dated August 1977.

cc: U.S. Nuclear Regulatory Commission Regicn I 631 Park Avenue King of Prussia, PA 19406 Mr. R. H. Nessman, Project Manager Division of Reactor Projects I/II Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Senior NRC Resident Inspector Pilgrim Nuclear Power Station Mr. Robert M. Hallisey, Director Radiation Control Program Mass. Dept. of Public Health 150 Tremont Street F-7 Boston, MA 02111 I

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7 Attachment 1 Pronosed Technical Snecification Chances for Reload 7 Pronosed Chanaes It is proposed that the Pilgrim Nuclear Power Station (PNPS) Technical Specifications be changed as shown on the attached Technical Specification pages and described below. Affected Technical Specification pages include Pages 7, 8, 46, 53, 59a, 68, 203, 205A-1, 2058-2, 205C-3, 205E-1 through 205E-6, 205F, and 206m.

1. References to non-retrofit 8X8 fuel are removed from Technical Specification Pages 7, 8, 205A-1, 2058-2, 205C-3, and 206m because this

- fuel type is not to be used during Cycle 8 operation of PNPS.

2. The descriptions of low and low-low reactor water level setpoints with respect to the top of active fuel are revised on Technical Specificacion Pages 46, 53, 59a, and 68 to acknowledge the change in height of the top of active fuel of retrofit fuel. The actual reactor water level trip setpoints remain unchanged.

3. The operating limit minimum critical power ratio (MCPR) values in Technical Specification Table 3.11-1 (Page 2058-2) are revised for operation during Cycle 8 at PNPS.

4. The spelling of the acronym "MFLPD", maximum fraction of limiting power density, is corrected on Technical Specification Page 8.

5. Technical Specification Pages 203, 205B-2, 205E-1 through 205E-6, and 205F are revised to clarify that the unit of measure used at PNPS for fuel exposure is megawatt days per standard ton (MHD/ST).

6. Technical Specification Bases 3.11.C on Page 205C-3 are revised to add a reference to Technical Specification Section 2.1 that was inadvertently deleted in Amendment 54.

Reasons for Changes One significant feature of the Reload 7 core design at PNPS is that for the first time, only retrofit fuel will be loaded into the core. All remaining non-retrofit 8x8 fuel at PNPS will be discharged to the spent fuel pool. For this reason, Proposed Change 1, above, requests that all references to specific fuel types be removed from these Technical Specification sections.

The maximum average planar linear heat generation rate (MAPLHGR) limits in Technical Specification Figures 3.11-1 through 3.11-7 will continue to specify the fuel types that have been approved for use at PNPS. The MAPLHGR limits for non-retrofit 8x8 fuel are to remain in Technical Specification Figures 3.11-1 through 3.11-3 because of their possible reinsertion into the PNPS core during future plant operating cycles.

Because the top of active fuel for retrofit fuel is 1.24 inches higher than the earlier non-retrofit 8x8 fuel, Proposed Change 2, above, requests that the relationships between reactor water level trip setpoints and the top of active fuel be revised accordingly. The actual reactor water level trip setpoints remain unchanged.

_ . . . .- .- --= __ - -- . -

8 Proposed Change 3, above, requests that the operating limit MCPR values in Technical Specification Table 3.11-1 be revised to permit operating ,

flexibility during Cycle 8 operation while maintaining design safety margins.  :

The minor editorial changes in Proposed Changes 4 through 6, above, are  !

requested to correct previous errors and to exactly define which unit of 1 measure is used at PNPS for fuel exposure.

l Safety Evaluation and Determination of No Sianificant Hazards Considerations In accordance with 10CFR50.91, the following analysis has been performed using the standards in 10CFR50.92, concerning the issue of significant hazards consideration.

1. Operation of PNPS in accordance with the proposed amendment would not

, involve a significant increase in the probability or consequences of an accident previously evaluated.

The Reload 7 core design at PNPS has been analyzed using the NRC-approved methods contained in NEDE-24011-P-A-8, GESTAR-II, dated May 1986. PNPS specific information and Reload 7 core loading are contained in the Supplemental Reload Licensing Submittal for PNPS, Reload 7, General

Electric Report 23A4800, Revision 0, Dated December 1986. Together, these documents provide justification that operation during Cycle 8 at PNPS <

using the operating limit MCPR values in revised Technical Specification Table 3.11-1 will not significantly increase the probability or '

consequences of an accident previously evaluated.

. The Reload 7 core design at PNPS marks the first time that only retrofit fuel will be used in the core. An analysis has been performed to provide E justification that the use of this previously approved retrofit fuel in ,

the core does not require a change in any reactor water level trip setpoints. For this reason, it is concluded that the change in the relationships betwet.. the reactor water level trip setpoints and the top of active fuel does not involve a significant increase in the probability 1 or consequences of an accident previously evaluated. The analysis is

provided-in General Electric Report NED0-21696, Loss-of-Coolant Accident (LOCA) Analysis Report for PNPS, dated August 1977, as updated and amended by the following:

! a. General Electric Topical Report NED0-30767, PNPS LOCA Analysis Update, dated September 1984.

b. Errata and Addenda Sheet 4, dated August 1986, to LOCA Analysis Report for PNPS, NED0-21696, dated August 1977.

c. Errata and Addenda Sheet 5, dated October 1986, to LOCA Analysis Report for PNPS, NED0-21696, dated August 1977.

2. Operation of PNPS in accordance with the proposed amendment would not 4

create the possibility of a new or different kind of accident from any 4

accident previously evaluated.

i

The replace ent fuel to be inserted into the core for Reload 7 at PNPS is substantially of the same design as existing fuel in the core and has been previously reviewed and approved by the NRC for use at PNPS. Thus, it is concluded that operation of the Reload 7 core using the MAPLHGR limits provided in Technical Specification Figures 3.11-1 through 3.11-7 will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Operation of PNPS in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

As described above, the Reload 7 core design has been analyzed using NRC-approved methods, and meets or exceeds all safety design margins.

Specifically,

a. The minimum shutdown margin through Cycle 8 is >l.0% delta k/k, which exceeds the Technical Specification requirement of R+0.25% delta k/k, where R for Cycle 8 only includes the 0.04% delta k/k allowance for inverted tubes in the old control blades.

b. Technical Specification Table 3.11-1 is proposed to be revised to reflect the MCPR value of 1.44 for the limiting transient of load rejection without bypass.

c. The peak vessel pressure resulting from MSIV closure with flux scram is 1315 psig, which remains below the 1375 psig ASME over-pressurization limit.

d. The pressure margin to spring safety valve actuation is 78.3 psid, which exceeds the 60 psid minimum margin requirement.

e. The core loading meets the Technical Specification requirement of having at least one burnt fuel bundle with exposure >1000 MWD /ST in each four-bundle cell.

For these reasons, it is concluded that the proposed amendment will not significantly reduce a margin of safety.

This modification to PNPS Technical Specifications does not present an unreviewed safety question as defined in 10CFR50.59. It has been reviewed and approved by the Operations Review Committee and reviewed by the Nuclear Safety Review and Audit Committee.

Schedule of Change It is requested that the proposed amendment become effective immediately upon receipt of approval by the Commission.

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