ML19347E753

Amend 5 to License DPR-22 Consisting of Changes to Tech Specs 3.0 & 4.0,incorporating Requirements of Odyn & New MAPLHGR & MCPR Limits W/Recent Analyses for Cycle 9 Operation

ML19347E753
Person / Time
Site:	Monticello
Issue date:	05/04/1981
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML19347E754	List: ML19347E753 ML19347E756
References
NUDOCS 8105130321
Download: ML19347E753 (15)

Text

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UNITED STATES f

NUCLEAR REGULATORY COMMISSION yg c(

WASHINGTON, D. C. 20555 7,,

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w NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDfENT TO FACILITY OPERATING LICENSE' Amendment No. 5 License No. DPR-22 s

1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Northern States Power Company (tile licensee) dated February 6,1981, as supplemented by letter dated March 19, 1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comissign's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission, i

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be ccnducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

8105130$W

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2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-22 is hereby amended to read as follows:

2.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. S

, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.,

FOR THE NUCLEAR REGULATORY COMMISSION f

omas A..Ippolito, Chief Operating Reactors Branch !2 Division of Licensing

Attachment:

Changes to the Technical Speci fications Date of Issuance: May 4,1981 l

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ATTACHMENT TO LICENSE ~ AMENDMENT NO. 5 FACILITY OPERATING LICENSE NO. DPR-22"- ~~.I

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DOCKET NO. 50-263

)

Remove the following pages and insert identically numbered pages:

26 82 89 90 213 214 216 217 I

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i

e 3.0 LlHITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIP.EMENTS 3.1. HEACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM.

'l gplicability:

Applicability:

[.

Applies to the instrumentation and associated Applies to the surveillance of the instruraentation devices which initiate a reactor scram, and associated devices which initiate reactor scram.

i.

Obejective:

Objective:

)

To assure the operability of the reactor To specify the type and frequency of surveillance protection system.

to be applied to the instrumentation that initiates a scram to verify its operability.

g Specification:

Specification:

A. The setpoints, minimum number of trip A. Instrumentation systems shall be functionally systems, and minimum number of instrument tested and calibrated as indicated in Tables channels that must be. operable for each 4.1.1 and 4.1.2, respectively.

i.,

position of the reactor mode switch shall he as given in Table 3.1.1.

The time from

!l initiation of any channel trip to the de-energization of the scram pilot valve solenoids l

,l shall not exceed 50 milliseconds. '

I ltl E

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26 1/4.1 i

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i Amendment No. 5 l

f

4.0 SURVEILLANCE REQUIREMENTS 1.0 1.lHITING CONDITIONS FOR OPERATION s

Any four rod group may contain a control rod which is valved j.

out of service provided the above requirements and Specification 3.3.A are met.

3. If the cycle average scram insert ion time (%v,), based on the de-energization of the scram pilot value solenoids at time sero, of all operable control rods in the reactor power bperation condizion at tlic 20% inserted position is larger than the adjust-d analysis mean scram time ( 'To

), a more restrictive MCPR g

t 1imit (see section 3.1I.C.!) shalI be used, p

i D. Control Rod kccumlators D. Control Rod Accumulators I

rod accumulator may be once a shift check the atatus in tha At all reactor operating pressures, inoperable provided that no other control rod in the nine'-

control room of the accumulators pressure l a

and level alarms.

rod square array around this rod has a:

1. Inoperable accumulato.

2. Direct ional control valve electrically disarmed wlille in a non-fully insert ed position.

I i

If a control rod with an inoperable accumulator is inserted

" full-in" and its directional control valves are electrically disarmed, it shall not be considered to have an ina.nerabe accumulator.

82 1.1/4.1 I

Amendment No. 5

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Bases Continued 3.3 and 4.3:

consequences of reactivity, accidents are functions of the initial neutron flux. The require-ment of at least 3 counts per second assures that any transient, should it occur, begins at or above the initial value of 10% of rated power used in the analyses of transients from cold conditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control r'od withdrawal.

A minimum of two operable

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l SRH's are provided as an added conservatism.

i-5.

The consequences of a rod block monit.or failure have been evaluated. These evaluations show I

that during reactor operation with certain limiting control rod patterns, the withdrawal of a designated single control rod could result in one or more fuel rods with HCPR's below the Safety Limit (T.S.2.1.A).

During use of such patterns, it is judged that testing of the RBH system prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur.

It is the responsi'bility of the Engineer, Nuclear, to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable rods in other than limiting patterns.

C.

Scram Insertion Times The control rod system is designed to bring the reactor subcritical at a rate f ast enough to prevent fuel damage; i.e., to prevent the HCPR from becoming less than the Safety Limit (T.S.2.1.A).

i This requires the negative reactivity insertion in any local region of the core and in the

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overall core to be equivalent to at least the scram reactivity curve used in the transient analysis. The required average scram times for three control rods in all two by two arrays and the required average scram tim'es for all control rods are based on inserting this amount of negative reactivity at the specified rate locally and in the overall core. Under these conditions, the thermal limits are never reached during the transients requiring control rod scram. The j

limiting operational transient is that resulting from a turbine stop valve closure with failure

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of the turbine bypass system. Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above j

Specification, provide the required protection, and MCPR remains above the Safety Limit (T.S.2.1.A).I c;

3.3/4.3 BASES 89 Amendment No. 5 i

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Itases Continued 3.3 and 4.3:

The analysis assumes 50 milliseconds for Reactor Protection System delay, 200 milli seconds from de-energization of scram solenoids to the beginning of rod motion, and 175 milliseconds later the rods are at the 5% position.

Section 3.3.C.3 allows a lower MCPR limit to be used if the cycle average scram time (% ) is less than the adjusted analysis mean scram time ('Ts) (see Reference 7, of Section 3.11) i'

't;,, is the weighted cycle average scram t ime to the 20% insertion position (~ notch 38) of all the operable, rods measured at any point in the cyce.

where:

n = the nurober of surveillance testa performed i

to date in this cycle.

l; N Tg I

g i = number of control rods measured in the i=1 th test.

I, g

I

'Y

= average scram time to the.20% insertion l.

N i

position of all rods measured in the ith j

i=1 test.

q,is the adjusted analysis mean scram time where:

N

= total number of active rods measured in j!

g to the 20% insertion position.

the first test following core alterations.

I 0.710 = the mean scram time used in the g

analysis.

I N

0.0875 = 1.65x0.053 where 1.65 is the appropriate i

g

'r = 0.710 + 0.0875 statistical number to provide a 95%

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confidence level and, 0.053 is the

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y j

standard deviation of the distribution i

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for average scram insertion time to the

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q 20% position, that was used in the analysis.

3.3/4.3 BASES 90 Amendment No. 5 j

3.0 LittlTitlG CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREHENTS C. Ilinimum Critical Power Ratio (MCPR)

C. Minimum Critical Power Radio (MCPR)

1. Diring power operation the Operating MCPR Limit shall be HCPR shall be determined daily during reactor

> 1.43 for 8x8 and 8x8R fuel, > 1.47 for P8x8R fuel at power operation at > 25% rated thermal power rated power and flow, provided

'i; >[U.,* (see section and following any cEange in power level or 3.3.C.3).

If at any time during operation it is

'istribution which has the potential of det ermined that the limiting value for HCPR is being b.inging the core to its operating HCPR Limit, exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

Surveillance and corresponding act ion shall continue until reactor operation is within the prescribed limits.

If the steady state flCPP is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

For core flows other than rated the Operat ing HCPR 1,imit shall be the above applicable HCPR value time K wh ere f

K is as shown in Figure 3.11.3.

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2. If the gross radioact ivity release rate of noble gases at the steam jet air ejector monitors exceeds, for a period greater than 15 minutes, the equivalent of 236,000 uci/sec following a 30-minute decay, the Operating itCPR I.imits specified in 3.11.C.1 shall be adjusted to

>l.48 for all fuel types, times the appropriate K.

l Subsequent operat ion with the adjusted MCPR valuek shall f

he per paragraph 3.ll.C.I.

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'lin > 1'., the operating HCPR Limit shall be a linear interpolation between the limits in 3.11.C.1 and 1.48'for 8x8 and 8x8R fuel and 1.52 for P8x8R fuel.

j 213 3.11/4.11 Amendment flo. 5

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Itani n Continued Itinimum Critical Power Ratio (HCPR)

C.

The ECCS evaluation presented in Reference 4 and Heference 6 assumed the steady state HCPR prior to the'the Operat'i to be 1.24 for all fuel types for normal and reduced flow.

accident By maintain-root.ulated loss-of-coolantis determined frtxa the analysis of transients discussed in Bases Sections 2 flCrit I.iinit ing an operating flCPR above these limits, the Safety I.imit (T.S.

limiting abnormal operational transient.

most to be a factor in determining the HCPR itse of CF.'s new ODYH code Option B will reqisire average scram time (Reference 7).

In order to increase the operating envelope for HCPR below HCPRg (ODYH code Option A), the is below the adjusted analysis cycle average scram time (Gar ) must be determined (see Bases 3.3.C).

If %s Ta. W,, a linear inurpolution must be used to M ermine tlie if scram time, Llie HCPR, I.imit can be used,

appropriate HCPR.

For example:

k US tlCrit = HCPR +

-(HCPR -HCPRB)

A a

g,9_.g ticrit and HCPR have been determined f roni the most limiting accident analyses.

D than rated core flow the Operating MCPR I.imit is adjusted by multiplying the above For operation with less Reference 5 discusses how the transient analysis done at rated conditions encoinpasses the limit by K.

g factor is applied, reduced flow situation when the proper Kg stated in 3.II.C.2 are indicative of fuel f ailure. Since the failure tioble gas activity levels above that mode cannot be positively identified, a more conservative Operating HCPR Limit must be applied to account for a possible fuel loading error.

in an autosaatic reactor those abnorinal operat ional transients, analyzed in FSAR Section 14.5, which result scram are not considered a violation of the I.CO.

Exceeding HCPR limits in such cases need not be reported.

,216

1. I I n A:;i.:;

ilmendnent flo. 5

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p.i s e s Continued 1

lieferences

1. " Fuel Densification Effects in General Electric Boiling Water Reactor Fuel," Supplements 6, 7..and 8, l

HEI)H-10735, August, 1973.

2. Supplement I to Technical Report on Densification of General Electric Reactor Fuels, December 14, 1974 (USAEC Regulatory Staff).

3. Communicat ion : VA Hoore to IS Hitchell, "Hodified CE Hodel for Fuel Densification," Docket 50-323. Harch 27, 1974.

4. " Loss-of-coolant Accident Analysin Heport for the Honticello Nuclear Generating Plant," HEDO-24050 -1, December, 1980, I. O Hayer (HSP) to Director of N,uclear Reactor Regulation (USNRC),

February 6 1981.

5. " General Elect ric BWR Cencric Reload Application for 8x8 Fuel," NEDO-20360, Revision I, November 1974.

6. " Revision cf 1.ow Core Flow Ef fects on I.0CA Analysis for Operating BWR's," R L Gridley (CE) to D G Eisenhut (USHRC), September 28, 1977.

7. " Response to NRC Request for Informat ion on ODYN Computer Hode," R 11 Buchholz (CE) to P S Check (USNRC), September 5, 1980.

panes 4.11

'ihe API.llGR, 1.llGR and MCPR aliall be cliccked daily to determine if fuel burnup, or control rod movement have caused c h any,e s in power distribution. Since changes due to burnup are slow, and only a few control rods are removed daily, a daily clicck of power distribut ion is adequate.

For a limiting value to occur below 25% of rated thermal power, an unreasonably large peaking factor woiId be required, wInich is not tiie case for operating control rod sequences.

In aildi t ion, tlic HCPR is checked winenever changes in the core power level or distribut ion are made which have the potent ial' <

of bringing the fuel rods to their thermal-hydranslic limite.

217 4.11 HASES J

l Amendment No. 5

o UNITED STATES NUCLEAR REGULATORY COMMISSION f

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,. E WASMNGTON, D. C. 20555 7,; Dg2nK / f o

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>> w SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l

SUPPORTING AMENDMENT NO. 5 TO FACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING' PLANT

, DOCKET NO. 50-263 1.0 Introduction Northern States Power Company (the licensee) has proposed changes to the

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Technical Specifications of the Monticello facility in Reference 1.

The pro-posed changes relate to the core for Cycle 9 operation at power levels up to 1670 Mit (100% power).

In support of the ceload application, the licensee has i

enclosed proposed Technical Specification changes in Reference 1 and the GE BWR supplemental licensing submittal (Reference 2). The licensee has also proposed changing the allowable RPS delay time from 100-milliseconds to 50 milliseconds.

This reload involves loading of prepressurized GE 8x8 retrofit (P8x8R) fuel having drilled lower tieplates. This is the same type of fuel as was loaded during the last reload. The description of the nuclear and mechanical designs of 8x8 retrofit is contained in References 3 and 4.

Reference 3 also contains a complete set of references to topical reports which describe GE's analytical methods for nuclear, theralshydraulic, transient and accident calculations, and infonnation regardi g the applicability of these methods to cores containing a mixture of fuel. The use and safety implications of prepressurized fuel have i

been found acceptable per Reference 4.

The conclusions of Reference 5 found that the methods of Reference 3 were generally applicable to prepressurized fuel.

Therefore, unless otherwise specified, Reference 3, as supported by Reference 5,

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is adequate justification for the current application ^ of prepressurized fuel.

2.0 Evaluation 2.1 Reactor Physics The reload application follows the procedure described in NEDE-240ll-P,.

" Generic Reload Fuel Application." We have reviewed this-application and the censequent Technical Specification changes. The transient analysis input parameters are typical for BWRs and are acceptable.

Core wide transient analysis results are given for the limiting transients and the required operating limit values for MCPR are given for each fuel type.

The revised MCPR limits are required by the reload and they are acceptable.

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2.2 Themal Hydraulics As stated in Reference 3, for BWR cores which relcad with GE's retrofit 8x8R fuel, the safety limit minimum critical power ratio (SLMCPR) resulting from either core-wide or localized abnomal operational transients is equal to 1.07.

When meeting this SLMCPR during a transient, at least 99.9" of the fuel rods in the core are expected to avoid Boiling transition.

To assure that the fuel cladding integrity SLMCPR will not be violated during any abnomal operational transient or fuel misloading, the most limiting events have been reanalyzed for this reload by the li:censee, in or' der to determine which event results in the largest reduction in the minimum critical power ratio. These events have been analyzed for '

the expcsed fuel and fresh fuel.

Addition of the largest reductions in critical power ratio to the SLMCPR was used to establish the operating limits for each fuel ' type.

We have found the methods used for this analysis consistent with.previously approved past practice (Reference 3). We have found the results of this analysis and the corresponding Technical Specification changes acceptable.

2.3 ECCS Appendix K Input data and results for ECCS analysis have been given in References 1 and 2.

The informatdon presented fulfills the requirements for each analysis outlined in Reference 3.

We have reviewed _the analyses and infonnati.on submi.tted for the reload aad conclude that the Ebnticello plant will be in conformance with all requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50.46 when it is operated in accordance with the Technical Specifications we are issuing with this amendment.

2.4 RPS Delay Time The licensee has proposed to change the allowed time between a channel sensir3 a trip condition and the deenergization of the scram pilot valve solenoids from 100 milliseconds to 50 milliseconds. This change will bring the Technical Specifications into a reement with the value assumed in the licensing analysis. The change is in the more con-servative direction; the licensee has indicated that the 50 millisecond RPS delay time has been demonstrated to be attainable in previous tests.

We find the licensee's proposed change acceptable.

3.0 Environmer.tal Considerations We have determined that tne amendm'ent does not authorize a change in

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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 51.5(d)(a) that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared.in connection with the issuance of the amendment.

3-4.0 Conclus ion

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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 amend-ment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the gublic will not be endangered by operation in the proposed manner, and (3) sbEii activities will be conducted in compliance with the Commission's regu-lations 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.

Dated:

Ma; 4,1981 S

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REFERENCES l.

(a)

Letter, NSP to Office of Nuclear Reactor Regulation (USNRC),

dated February 9,1981 (b)

Letter, NSP to Office of Nuclear Reactor Regulation (USNRC),

dated March 19, 1981.

2.

" Supplemental Reload Licensing Submittal for Monticello Nuclear Generating Plant, Reload 8 (Cycle )", dated March 1981.

3.

" General Electric Boiling Water Reactor Generic Reload Application",

NEDE-240ll-P-A, May 1977.

4.

Letter, R. E. Engel (GE) to U. S. Nuclear Regulatory Commission, dated January 30, 1979.

5.

Letter, T. A. Ippolito (USNRC) to R. Gridley (GE), April 16, 1979, and enclosed SER.

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