ML112970813

Application for Amend to Ol.Changes to Tech Specs Requested Per Encl Exhibits A,B,C & D

ML112970813
Person / Time
Site:	Monticello
Issue date:	02/06/1981
From:	Mayer L Northern States Power Co
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ML112970812	List: ML112970811 ML112970813 ML19345F249 ML19354C423
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Download: ML112970813 (12)

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UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 REQUEST FOR AMENDMENT TO OPERATING LICENSE NO. DPR-22 (License Amendment Request Dated February 6, 1981)

Northern States Power Company, a Minnesota corporation, requests authorization for changes to the Technical Specifications as shown on the attachments labeled Exhibit A, Exhibit B, Exhibit C, and Exhibit D.

Exhibit A describes the proposed changes along with reasons for the change.

Exhibit B is a set of Technical Specification pages incorporating the proposed changes. Exhibit C is the Supplemental Reload Licensing submittal for Cycle 9. Exhibit D contains the LOCA Analysis Report.

This request contains no restricted or other defense information.

NORTHERN STATES POWER COMPANY By __ _

Manager of Nuclear Support Services On this day of 1Luses4e /4/, before me a notary public in and for said County, personally appeared O Mayer, Manager of Nuclear Support Services, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof and that to the best of his knowledge, information and belief, the statements made in it are true and that it is not interposed for delay.

Betty D Notar Public - Minnesota Ramsey County My Commission Expires December 16, 1987 102 170 e,1 z

0 0 EXHIBIT A MONTICELLO NUCLEAR GENERATING PLANT License Amendment Request dated February 6, 1981 Proposed Changes to the Technical Specifications Appendix A of Operating License DPR-22 Pursuant to 10 CFR 50.59 and 50.90, the holders of Operating License DPR-22 hereby propose the following changes to Appendix A, Technical Specifications:

1. Page 82, Section 3.3.C, Scram Insertion Times Continue d Proposed Change Add new Section 3.3.C.3.

Reason for Change This section cross references the Scram Insertion Time section with the Minimum Critical Power Ratio section. Lower MCPR's are allowed depending upon the relationship between the cycle average scram insertion time and the adjusted analysis mean scram time.

Safety Evaluation General Electric has shown that lower MCPR'S can be allowed with shorter scram times (Reference 1).

2. Page 90 Bases for Section 3.3 and 3.4 Continued Proposed Change Add the definitions of the cycle average scram time and adjusted analysis mean scram time below the existing paragraph on this page.

Reason for Change These definitions explain the new terms used in section 3.3.C.3.

Safety Evaluation These definitions are the same as those used by General Electric. The equations were obtained from Attachment 1, Issue 2 titled "Scram Insertion Time Conformance Procedure" of Reference 1. The constants 0.710 sec and 0.053 sec were obtained from Attachment 1, Issue 1, Table 1-1 page 1-7 of Reference 1. The constant 1.65 is the statistical number associated with the 95% confidence level for one tail of a normal distribution.

3. Page 213 Section 3.1l.C, Minimum Critical Power Ratio (MCPR)

Proposed Change 3.3.C.1: Change the MCPR limits for fuel type P8x8R. Add the restric tion that the cycle average scram time must be below the adjusted analysis mean scram time to use these limits. If this condition is not met, higher MCPR limits are specified at the bottom of the page.

3.3.C.2: The MCPR limits for all fuel types should be changed to 1.48.

Reason for Change 3.3.C.l: This change will allow the lower MCPR limits for shorter scram times.

3.3.C.2: This limit reflects the accident analysis done for this cycle.

Safety Evaluation The new MCPR limits are determined from page 4, item 11, Exhibit C.

4. Page 214 Table 3.11.1 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)

Proposed Change A new fuel type has been added, P8DRB284LB. New values have been included for all fuel types.

Reason for Change A re-analysis has been done using newer LOCA evaluation models (REFLOOD and CHASTE) and accounting for the fact that some assemblies have holes drilled in their lower tie plates.

Safety Evaluation Exhibit D, Loss-of-Coolant Accident Analysis Report for Monticello Nuclear Generating Plant, NEDO-24050-1 Table 4A thru 4H.

5. Page 216 Bases for Section 3.11.C Proposed Change Following the paragraph beginning "The ECCS evaluation..." insert the paragraph discussing the ODYN options.

Reason for Change The paragraph provides the basis for the changes made to section 3.1l.C and describes the method to be used for linear interpolation when the cycle average scram time is larger than the adjusted analysis mean scram time.

Safety Evaluation Reference 1, Attachment 1, Issue 2, page 2-3.

6. Page 217 References for the Bases for 3.11 Reference 4 has been updated to reflect the new Analysis Report.

Reference 7 has been added. These changes are administrative in nature and have no effect on safety.

References:

1) Letter "Response to NRC Request for Information on ODYN Computer Model" Buchholz (GE) to P S Check (USNRC), September 5, 1980.

EXHIBIT B License Amendment Request dated February 6, 1981 Docket No. 50-263 License No. DPR-22 Exhibit B consists of revised pages of Appendix A Technical Specifications as listed below:

Pages 82 90 213 214 216 217

T 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS t

Any four rod group may contain a control rod which is valved out of service provided the above requirements and Specification 3.3.A are met.

3. If the cycle average scram insertion time (1'Av), based on the de-energization of the scram pilot value solenoids at time zero, of all operable control rods in the reactor power operation condition at the 20% inserted position is larger than the adjusted analysis mean scram time ('re ), a more restrictive MCPR limit (see section 3.1l.C.1) shall be used.

D, Control Rod Accumlators D. Control Rod Accumulators At all reactor operating pressures, a rod accumulator may be Once a shift check the status in the inoperable provided that no other control rod in the nine control room of the accumulators pressure rod square array around this rod has a: and level alarms.

1. Inoperable accumulator.

2. Directional control valve electrically disarmed while in a non-fully inserted position.

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 inoperabe accumulator.

3.3/4.3 82 REV

Bases Continued 3.3 and 4.3:

estimated This is adequate and conservative when compared with the typical time delay of about 210 milli-seconds of the time interval results from the sensor from scram test results. Approximately the first 90 milliseconds de-energized. Approximately 120 milliseconds and circuit delays; at this point the pilot scram solenoid is rod motion is not later control rod motion is estimated to begin. However, to be conservative, control value was included in the transient analyses and is assumed to start until 200 milliseconds later. This included in the allowable scram insertion times of Specifications 3.3.C.1 and 3.3.C.2.

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

38) of all the operable.

is the weighted cycle average scram time to the 20% insertion position (- notch T'..v rods measured at any point in the cyce.

where: n = the number of surveillance tests performed

'nb to date in this cycle.

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

'average scram time to the 20% insertion

=-4 position of all rods measured in the ith test.

ryb is the adjusted analysis mean scram time where: N1 = total number of active rods measured in to the 20% insertion position. the first test following core alterations.

0.710 = the mean scram time used in the analysis.

(

0.0875 = 1.65x0.05 3 where 1.65 is the appropriate statistical number to provide a 95%

0.710 + 0.0875 confidence level and, 0.053 isthe standard deviation of the distribution for average scram insertion time to the 20% position, that was used in the analysis.

90 3.3/4.3 BASES REV

I 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS I.

C. Minimum Critical Power Ratio (MCPR) C. Minimum Critical Power Radio (MCPR)

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

> 1.41 for 8x8 and 8x8R fuel, > 1.45 for P8x8R fuel at power operation at > 25% rated thermal power rated power and flow, provided -y% > 14* (see section and following any change in power level or I distribution which has the potential of 3.3.C.3), If at any time during operation it is determined that the limiting value for MCPR is being bringing the core to its operating MCPR Limit exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. If the steady state MCPR 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 Operating MCPR Limit shall be the above applicable MCPR value time K where K is as shown in Figure 3.11.3.

f

2. If the gross radioactivity 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 MCPR Limits specified in 3.11.C.1 shall be adjusted to

>1.48 for all fuel types, times the appropriate K . I Subsequent operation with the adjusted MCPR values shall be per paragraph 3.11.C.I.

• If %46 >Y6 , the operating MCPR Limit shall be a linear interpolation between the limits in 3.11.C.1 and 1.49 for 8x8 and 8x8R fuel and 1.53 for P8x8R fuel.

213 REV 3.11/4.11

O TABLE 3.11.1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE vs. EXPOSURE Exposure MAPLHGR FOR EACH FUEL TYPE (kw/ft) 8DB219L 8DRB265L P8DRB265L 8DRB282 P8DRB282 P8DRB284LB MWD/STU 8DB262 8DB250 11.2 11.4 11.5 11.6 11.2 11.2 11.4 200 11.1 11.5 11.6 11.6 11.2 11.2 11.4 1,000 11.3 11.3 11.9 11.7 11.8 11.6 11.8 11.8 5,000 11.9 11.9 12.0 1118 11.9 11.7 11.9 11.9 10,000 12.1 12.1 11.9 11.7 11.9 11.7 11.8 11.9 15,000 12.1 12.1 11.8 11.6 11.8 11.5 11.7 11.7 20,000 12.0 11.9 11.3 11.3 11.3 11.3 11.3 11.4 25,000 11.6 11.5 10.2 10.7 10.7 11.1 11.1 10.8 30,000 10.5 10.6 9.7 10.2 10.2 10.4 10.4 10.2 35,000 9.8 9.6 9.0 9.1 9.6 9.6 9.8 9.8 9.5 40,000 8.9 45.000 8.0*

50,000 7.3*

• For extended burnup program test bundles 214 REV 3.11/4.11

Bases Continued C. Minimum Critical Power Ratio (MCPR) 6 assumed the steady state MCPR prior to the The ECCS evaluation presented in Reference 4 and Reference normal and reduced flow. The Operating postulated loss-of-coolant accident to be 1.24 for all fuel types for discussed in Bases Sections 2.1 and 2.3. By maintain MCPR Limit is determined from the analysis of transients event of the Safety Limit (T.S. 2.1.A) is maintained in the ing an operating MCPR above these limits, the most limiting abnormal operational transient.

time to be a factor in determining the MCPR Use of GE's new ODYN code Option B will require average scram code Option A), the (Reference 7). In order to increase the operating envelope for MCPR below MCPRA(ODYN Bases 3.3.C). If rYTug, is below the adjusted analysis cycle average scram time ('ms- ) must be determine4 (see must be used to determine the scram time, the MCPRB Limit can be used. If Y& > a linear interpolution appropriate MCPR. For example:

MCPR = MCPRB + (MCPRA-MCPR )

B 0.9- 'YiAs accident analyses.

MCPRA and MCPRB have been determined from the most limiting MCPR Limit is adjusted by multiplying the above For operation with less than rated core flow the Operating analysis done at rated conditions encompasses the limit by Kf. Reference 5 discusses how the transient reduced flow situation when the proper K factor is applied.

are indicative of fuel failure. Since the failure Noble gas activity levels above that stated in 3.11.C.2 Operating MCPR Limit must be applied to account mode cannot be positively identified, a more conservative for a possible fuel loading error.

Section 14.5, which result in an automatic reactor Those abnormal operational transients, analyzed in FSAR MCPR limits in such cases need not be reported.

scram are not considered a violation of the LCO. Exceeding 216 3.11 BASES REV

Bases Continued References 8,

1. "Fuel Densification Effects in General Electric Boiling Water Reactor Fuel," Supplements 6, 7,.and NEDM-10735, August, 1973.

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

Docket

3. Communication: VA Moore to IS Mitchell, "Modified GE Model for Fuel Densification,"

50-321, March 27, 1974.

Plant," NEDO

4. "Loss-of-coolant Accident Analysis Report for the Monticello Nuclear Generating 24050 -1, December, 1980, L 0 Mayer (NSP) to Director of Nuclear Reactor Regulation (USNRC),

February 6 1981.

1, November

5. "General Electric BWR Generic Reload Application for 8x8 Fuel," NEDO-20360, Revision 1974.

to

6. "Revision of Low Core Flow Effects on LOCA Analysis for Operating BWR's," R L Gridley (GE)

DG Eisenhut (USNRC), September 28, 1977.

to P S Check

7. "Response to NRC Request for Information on ODYN Computer Mode," R H Buchholz (GE)

(USNRC), September 5, 1980.

Bases 4.11 rod movement have caused The APLHGR, LHGR and MCPR shall be checked daily to determine if fuel burnup, or control only a few control rods are removed daily, a changes in power distribution. Since changes due to burnup are slow, and value to occur below 25% of rated thermal power, an daily check of power distribution is adequate. For a limiting rod sequences. In which is not the case for operating control unreasonably large peaking factor would be required, are made which have the potential addition, the MCPR is checked whenever changes in the core power level or distribution of bringing the fuel rods to their thermal-hydraulic limits.

217 4.11 BASES REV

0 0 EXHIBIT C License Amendment Request dated February 6, 1981 Docket No. 50-263 License No. DPR-22 Exhibit C is the Supplemental Reload Licensing submittal for Monticello Nuclear Generating Plant Reload 8 (Cycle 9).

EXHIBIT D License Amendment Request dated February 6, 1981 Docket No. 50-263 License No. DPR-22 Exhibit D is the Loss-of-Coolant Accident Analysis Report for Monticello Nuclear Generating Plant.