ML20245C315

Safety Evaluation Supporting Amend 9 to License DPR-66

ML20245C315
Person / Time
Site:	Beaver Valley
Issue date:	07/12/1977
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20245C312	List: ML20245C308 ML20245C313 ML20245C315 ML20245C317
References
TAC-6811, TAC-7861, NUDOCS 8710260061
Download: ML20245C315 (6)

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% UNITED STATES .4

.? NUCLEAR REGULATORY COMMIS510N /b '

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WASHINGTON, D. C. 20055 -

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMEN 0 MENT NO. 9 TO FACILITY OPERATING LICENSE NO. OPR-66 0UOVESNE LIGHT COMPANY d OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY BEAVER VALLEY POWER STATION UNIT NO. 1 .

DOCKET NO. 50-334 Introduction Bf' letter dated February 11,~1977,'Ouquesne' Light'Companj (DLC) requested ,

approval of proposed Technical Specification changes for Beaver Valley Power Stction Unit No.1 (BVPS-1) relating to the Total Nuclear Peakina Factor (F Q) and the Axial Power Distribution Monitoring System (APDMS)!

t 0iscussion I

On August 27, 1976, we issued an Order for Modification of License for BVPS-1 which modified the Technical Specification limit for the Fq to 2.22. The Order was issued because recent. operating data gathereo at the Connecticut Yankee facility indicated that reactor vessel upper l-head fluid temperatures were higher than reactor inlet water temperatures i for reactor vessels similar in design to that in operation at BVPS-1.

These higher upper head temperatures would have the effect of increasing the calculated peak clad temperature in the event of a loss of coolant -

accident (LOCA). Our Order required DLC to submit a reevaluation of t emergency core cooling system (ECCS) performance with an approved j Westinghouse Model with an appropriate correction for upper. head water  !

temperature.

Nuclear Peakina Factor By letter dated November 30, 1976, DLC submitted the requested re-evaluation of the ECCS performance. The reevaluation ysed the October F 1975 version of the Westinghouse ECCS Evaluation Modelli, 2, 3, 4) l 8710260061 770712 PDR ADOCK 05000344 ~

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and assumed the upper head fluid temperature equal to the fluid outlet  !

(hot leg) temperature. The reanalysis superspdas the previously per- ,"

formed ECCS evaluation for a large break LOCAtSJ which used the October 1975 version of the evaluation model b'ut which assumed that  !

j the upper head fluid temperature was equal to cold leg temperature, ' 4 A reanalysis of the small break LOCA was not required since the small breaks change 01 (gre The relatively insensitive.to reevaluation the upper of the ECCS head fluid performance temperature in Westinghouse q plants was required because recent experimental data had indicated that the actual temperature in the upper reactor vessel head was in the range of the cold leg temperature plus 50 to 75 percent of the difference j between vessel inlet and outlet temperatures (7). Westinghouse has l proposed to undertake a program designed to measure this temperature .i in the operating reactors. In the meantime, the NRC decided con- I servatively to request all the' Westinghouse plant licensees to reevaluate the ECCS performance with upper head fluid temperature equal to the l fluid outlet temperature. i NeECCSanalysisprovided'byDLCwasperformedwiththefollowing 4 input parameters: _j Licensed Core power' - 102% of 2652 MWt 102% of 12.07 kw/ft Peak Linear Power

- Nuclear peaking Factor - 2.32 .

j 1025 ft.3 each L Accumulator' Water Volume -

The analysis consisted of the evaluation of ECCS performance for double ended cold leg guillotine break (DECLG) with L discharge coefficient .

j CD of 0.4 and assumed one. percent uniform plugging of steam generator tubes. Also the rod bow linear power penalty, which was explicitly j incorporated in the previous analysis, has been removed. The removal of the rod bow penalty results from a recent Westinghouse study (3) which  !

shows that for burnups less than or equal to 24,000 MWD /MTU, the rod f bow penalty is equal to or less .than 5.7%. This red bow penalty is adequately accounted for statistically in the nuclear and engineering '

penalties incorporated into the original LOCA analysis. DLC identified the break with Cn = 0.4 as the critical break with respect to limiting values of peak cladding temperature and local zirconium water reaction from the previous break spectrum analysis which used the October 1975 [

version of the evaluation model but which was perform 9d with upper leg fluid temperature equal to the cold leg temperature (5). DLC justified 1

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the use of this analysis for determining critical break size by ,

referencing the generic sensitivity study performed by Westinghouse (6)  :

in which it was shown that the change in the upper head fluid temper-  !

ature does not affect the critical break size. The peak cladding i temperature and the maximum local zirconium water reaction corresponding to the critical break identified by DLC are 20al*F and 5.05 percent, respectively.

Based on our review,'we conclude that the results of the ECCS reanalysis, using the October 1975 version of the Westinghouse ECCS Evaluation Model with upper head temperature equal to the outlet (hot leg) temperature, are conservative relative to the 10CFR50.46 criteria. However, the ECCS i analysis was limited to one percent plugging of steam generator tubes and an ECCS reevaluation by DLC would be required in order for BVPS-1 to operate above this limit. Also, the removal of the linear power i penalty put the upper limit on-fuel burnup at 24,000 MWD /M U. With i tne above mentioned considerations, we find the results of the re- .!

analysis submitted by OLC acceptable, and since the Fg assumed in the r,eanalysis was 2.32, this value is acceptable and should replace the value of 2.22 presently in the Technical Specifications.

Ailal Power Distribution Monitorine System By their letter dated February 11,19 77, DLt, also requested that:the .

BVPS-1 Technical Specifications be amended to delete the required use -

of. the APDMS oelow an average core burnup of 3500 MWD /MTU. I The basis for this request is an increase of the value of the Fg at l rated power to 2.32. The increased permissible linear heat rate allowed j l from increasing Fg to 2.32 is about 4". and corresponds, for invariant ,

radial peaking factors, to an increase of the permissible axial peaking j factor (F g) of about a%. DLC states .that, using a constant axial control  :

strategy, the core axial power distribution can be controlled to the less  !

restrictive value of F, using excore instrumentation alone; hence, there  !

is adequate justification for deletion of incore monitoring. requirements  !

using the APDMS. l I

Constant a ial control (CAC) strategy has been described in the open literatura 10) and deferced in Westinghouse proprietary topical report WCAPS385[11). Validity of the CAC strategy a' be I been confirmed by work performed at Brookhavenfl2) . Theseginning studies of cycle has  !

show that using CAC (e.g. , a target axial offset and a 5% flux difference i l

~ control band about the target axial offset), the realizable reactor j power shapes, and hence magnitude and location of limiting conditions, i l

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i will be less severe than the initial conditions used in the safety j analysis. The studies address base load and load follow operation, i The studies show on a generic basis that presuming use of CAC strategy, 3 monitoring the reactor by excere instrumentation alone will adequately j assume that the limiting value of Fg of 2.32 will not be exceeded.

A previous ECCS analysis .for BVPS-1(9) supported a value of F less than or equal to 2.23. (Asdiscussedabove,thisvaluewasrehuced by an Order to 2.22.) Based on an F less than or equal i plant specific CAC analysis for cycl 1 was submitted (13)to 2.23, aand reviewed (9) .

The plant specific analysis assumed a + 7% flux difference monitoring i band. The analysis supported use of tiie CAC strategy with monitoring ,

on excore instrumentation alone for core average fuel exposure greater i For core burnup less than 3500 MWD /MTV  !

than or equal to 3500 MWD /MTU.

additional monitoring using the APDMS was required. l We concluded above that an Fn of 2.32.is acceptable for SVPS-1. The revise value of Fg is consiYtent with the generic Westinghouse topical -

^ report 11Lwhich justifies monitoring on excores alone presuming a '

i 5% flux difference monitorjne band. The plant specific analysis _.;

which we previously reviewedt9) assumed a i 7% flux difference monitor- #

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4 ing band. The extended monitoring band increase: the realizable power ,

distributions, most notably the axial power shapes, that must be con- .

sidered as initial conditions to the safety analysis. This plant specific analysis for fuel exposure greater than 3500 MWO/MTV has ex-

. plicitly verified the safety of the expanded monitoring band nresuming excore monitoring alone. For core exposure less than 3500 MWD /MTU, a 7% flux difference monitoring band and a value of Fg of 2.23, this analysis showed that monitoring of the axial power distribution using ...,.

the incore APDMS would be required above 95.5% of rated power. Review of this analysis shows sufficient margin with the revised value of Fg of 2.32 to justify deletion of required use of the APCMS at exposures f j

less than 3500 MWD /MTU.

I Based on review of available analysis, we conclude that operation of  !

BVPS-1 during cycle 1 using a constant axial control strategy and a  !

7% flux difference monitoring band with excore monitoring alone is acceptable,  !

and that use of the incore APDMS is no longer necessary and its required ,

use may be deleted from the Technical Specifications. l I -

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Our review of Reference 11 is nearing completion. Assuming accept-ance of this topical report, operation beyond cycle 1 with a 5% flux -

difference mnitoring band could be approved on a generic basis. ,

However, if DLC wishes to operate BVPS-1 beyond cycle 1 with a 7%  !

flux difference mnitoring band, a plant and cycle specific analysis  ;

will be required.

Technical Specifications l We have reviewed the Technical Specification changes relating to raising the FO to 2.32 and deleting the APOMS which DLC proposed by letter dated February 11, 1977. We have made some changes to DLC's submittal and DLC has agreed to the changes. The proposed Technical Specifications reflect a testing and surveillance program along with the limiting conditions for operation that provides assurance that the Fg upper bound limit of 2.32 will not be exceeded. We also con-clude that the proposed Technical Specifications are acceptable and -

consistent with those of other facilities operating with simDar systems and found acceptable by the staff.

O Environmental Consideration [i

I We have determined that the amendment does not authorize a change U in effluent types or total amounts nor an increase in power level  !

and will not result in any significant. environmental impact. Having I 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 551.5(d)(4), that an environmental impact statement, or negative declaration and environ-mental impact appraisal need not be prepared iri connection with the ,

issuance of this amendment. -

l Conclusion l' We have concluded, based on the considerations discussed above, that: [

(1)becausetheamendmentdeesnotinvolveasignificantincreasein  :

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 Connission's 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. i Dated: July 12,1977

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REFERENCES }

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1. WCAP-8339, " Westinghouse ECCS Evaluation Model - Sumary," July 1974.

2. WCAP-8471 (Proprietary), WCAP-8472 (Nonproprietary), " Westinghouse ECCS Evaluation Model - Supplementary Information;" April 1,1975. ,

3. WCAP-8622 (Proprietary), WCAP-8623 (Nonproprietary) " Westinghouse ECCS .

Evaluation Model - October 1975 Version," November 1975. ,

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4. ' Letter NS-CE-924, C. Eicheldinger (Westing' house) to D. B. Vassallo (NRC) dated January 23, 1976. ..

5. FSAR, Beaver Valley Power Station, Unit 1, Amendment 21,' June 1976.  :

6. WCAP-8853, " Westinghouse ECCS - Three-Loop Plant (17x17) Sensitivity i Studies," dated September 1976.

Letter NS-CE-1163, C. Eicheidinger (Westinghouse) to V. Stello (NRC),

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dated August 13, 1976.

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8. NRC '" Interim Safety-Evaluation R'ep~o'rt on We'itin~ghouse Fuel ~ Bowing," ,, ,

dated April 1976. _ , . ,. ,

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9. Supplement No. 3 FSAR, Beaver Valley Power Station, Unit 1 Ei March 19,1976 .,,

10. R. J. Sipush, et. al . ," Lead-Follow Demons trations Employing Constant

, Axial Offset Power-Distribution control Procedures ," Nuclear Technolosy, Vol . 31, October 1976,

11. T. Mori ta , et . al . ,'ITopical Report Power Distribution Control and Lead Following Procedures", Westinghouse WCAP-8385, September, 1974.

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12. A. Buslick, et.al., " Power Peaking During Load Following Using Westinghouse Constant Axial Offset Power-Distribution Control,"

BNL-NUREG-22 77, January,1977. c

13. Letter to D. B. Vassallo (NRC) from E. J. Woolever (Duquesne Light I Company) February 26,1975, "Seaver Valley Power Station-Unit No.1 Power Distribution Limits. " ,

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