ML17324A791
| ML17324A791 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/22/1986 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:0916T, AEP:NRC:916T, NUDOCS 8604250108 | |
| Download: ML17324A791 (17) | |
Text
REGULATGRYINFGRNATIGN-DISTR IBUTI.GN-SYSTEM-<-R-I DS)
ACCESSION NBR: 8604250108 DOC. DATE: 86/04/22 NOTARIZED-NO DOCNET ¹ FACIL:50-316 Donald C.
Cook Nuclear Polller Planti Unit 2.
Indiana 5
05000316 AUTH. NAME AUTHOR AFFILIATION ALEXICHiN. P.
Xndiana 5 Nichigan Electric Co.
REC IP. NANE RECIPIENT AFFXLIATXON DENTONi H. R.
Office of Nuclear Reactor Regulation>
Director (post 851125
SUBJECT:
Provides listed addi info to close out several outstanding items re Unit 2i Cycle 6 reload. Response to Question 5 re resulting reactor coolant f lofti core exit enthalpg 5
DNBR enc l.
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INDIANA8 MICHIGAN ELECTRIC COMPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 AEP:NRC: 0916T Donald C.
Cook Nuclear Plant Unit No.
2 Docket No. 50-316 License No.
DPR-74 ADDITIONAL INFORMATION RELATED TO THE D.
C.
COOK UNIT 2 CYCLE 6 RELOAD Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.
C.
20555
Dear Mr. Denton:
The purpose of this letter is to close out several outstanding items related,to the Unit 2, Cycle 6 reload.
The information is provided below.
l.
Additional Information Re ested b
the Reactor S stems Branch on Januar 8
1986 On January 8,
1986, your staff transmitted to AEPSC a request for additional information related to the Unit 2, Cycle 6 reload analyses performed by the Exxon Nuclear Company (ENC).
The response to questions numbered 1,
2, 3,
4, 7, and ll were transmitted to you by ENC on March 5, 1986 in their letter RAC:017:86.
Responses to questions numbered 6, 8, and 10 were submitted by ENC in their letter RAC:022:86, dated March 14, 1986.
ENC's response,to,Queption 10.was,a commitment to provide additional analyses related to the control rod ejection accident.
These analyses have been completed and were transmitted to you by ENC in their letter GNW:053:86, dated April 14, 1986.
We transmitted our response to Question 9 in our letter AEP:NRC:0916P, dated March 27, 1986.
Our response to Question 5, which concerns natural circulation flow from conditions of 20~ power, is included as Attachment 1 to this letter. This response includes an analysis of DNBR, which was performed by ENC at your staff's request.
Upon your acceptance of the information related to Questions 5 and 10, we believe that all open times identified in your January 8,
1986 letter should be closed.
8604250i08 860422 PDR
- DOCK 050003i6 P
PDR'
Mr. Harold R. Denton AEP:NRC:0916T 2.
Additional Information Re uested b
the Reactor S stems Branch on A ril 3 1986 3.
On April 3, 1986 we received an additional six questions related to the Cycle 6 reload.
These questions were discussed with your staff on April 7 and 10, 1986.
During those discussions we were informed that Questions 1, 2, 3,
and 6 would require no written response and would be closed out based on the verbal discussions.
The response to Questions 4 and 5 were transmitted to you by ENC in their letter GNW:055:86, dated April 18, 1986.
Assuming your acceptance of these responses, no open items remain concerning your April 3, 1986 questions.
LOCA Anal sis 4 ~
At the request of the NRC reviewer, Exxon is incorporating additional penalties in their Fuel Cooling Test Facility (FCTF) correlation factors.
A reanalysis incorporating these penalties is to be performed.
It is anticipated that the results will be submitted to the NRC by May 2, 1986.
Revisions to Pro osed DNB Technical S ecification 3 4.2.5.2.
In our letter AEP:NRC:0916I, dated March 15,
- 1986, we proposed to add a new Technical Specification (T/S) 3/4.2.5.2 (DNB Parameters-Modes 2 and 3).
This proposed T/S extended requirements for DNB-related parameters from the current Mode 1 to include restrictions in Modes 2 and 3 also.
During discussions with your staff on April 10,
- 1986, we were informed of their desire to extend the requirements to include restrictions in Modes 4 and 5.
In response to'his, we have included in Attachment 2 a revised version of T/S 3/4.2.5.2, which incorporates the additional restrictions.
It is noted that incorporation of these changes also required changes to T/S Table 3.2-2, T/S 3.4.1.3, the Index section, and the Action statement of T/S 3.2.5.2.
We have included revised versions of these pages in Attachment 2.
Changes to the original submittal are indicated by an additional bar in the right-hand margin of the affected pages.
Mr. Harold R.
Dent, on AEP:NRC:09l6T These revised changes have been reviewed by the Plant Nuclear Safety Safety Review Committee (PNSRC) and will be reviewed by the Nuclear Safety and Design Review Committee at their next regularly scheduled meeting.
This document has been prepared following Corporate procedures which, incorporate a reasonable set of controls to insure its accuracy and completeness prior to signature by the undersigned.
Very truly yours, Ale ich
~ n lp'ice Pre ident Attachments cc:
John E. Dolan W.
G. Smith, Jr. - Bridgman R.
C. Callen G.
Bruchmann G. Charnoff NRC Resident Inspector Bridgman
ATTACHMENT 1 TO AEP: NRC 0916T
RESPONSE
TO QUESTION 5
ON JANUARY 8 g 1 9 8 6 TRANSMITTAL
AEP:NRC: 0916T Attachment 1
Page 1 of 3
RESPONSE
TO QU1HTXON 5 OF NRC TRANSMZITALON 8 JANUARYI 1986 estion 5 Page 71 of XN-NF-85-28, Supplement 1 states that the P-7 setpoint is 11%.
with a 9% un~inty allowance so that loss of f1m'ithout reactor trip would be limited to conditions of about 20% pcarer.
At other Westinghouse designed plants P-7 limits power under loss of flmr conditions to 10% with 1%
uncertainty.
Justify that 20% pcarer under natural circulation conditions could be acconanodated without bulk boiling in the core or loss of DNB margin.
Justify the power peaking factors used in the analysis by carnparison to Technical Specification limits at 20% power.
Provide the resulting reactor coolant flow, core exit enthalpy and DNBR.
R~BS I1SB Pursuant to our discussions we have reviewed possible scenarios in an effort to identify those which might result in the condition indicated above.
Our investigation has determined that the only credible scenario which can result in a loss of all reactor coolant flow belier 10% power without a reactor trip is a Loss of Offsite Power.
However, this transient willde~ergize the electrical buses supplying the Reactor Coolant~ and will establish a Blackout condition.
A blackout condition automatically initiates a specific sequence of events including automatic shedding of various loads.
The Control Rod Drive Motor Generator, (MG) Sets will be shed from the 600 volt safety bus 2 seconds after the blackout: is established by opening of the MG set 600 volt supply breakers which are safety grade.
The MG sets willthen coast down.
These MG sets are designed incorporating a flywheel which prevents rod drops resulting from minor upsets in the MG set power supply.
There is an interlock associated with the MG set output breaker which willtrip the output breaker when the input breaker trips.
When this output breaker is tripped, the control rods will fall into the core.
The interlock mechanism and the MG output breaker are not safety grade and no credit is taken in the analysis of this accident for their action.
For purposes of the analysis, the time span after the MG sets are tripped from their power supply to the point at which the control rods fall into the core is determined by the coast down characteristics of the MG sets and the Rod Control System.
An inquiry to our NSSS vendor, Westinghouse Electric Corporation, concerning the MG coastdown characteristics was answered with the following.
"The flywheel is designed to provide power for one second with all rods out and two overlapped banks moving.
Considering conservative design this time could be extended to as long as five seconds.
Holding the rods stationary takes less pmer than moving them; therefore, if the control rods are withdrawn and stationary, the flywheel coastdown time could be increased by a factor of 10 to 20."
Based on the above we would expect that the control rods would fall into the core a maximum-of 102 seconds following the Blackout.
AEP:NRC:09l6T Attachment 1
Page 2 of 3 System Flaw Coast Dawn Measurements performed at D. C. Cook Unit 2 have demonstrated that follcaring loss of all four coolant pumps, there is still approximately 8% fleer after 102 seconds.
Exxon Nuclear Corporation has performed a calculation for this accident assuming no operator action and the control rods being inserted 102 seconds after the ass of Flow accident is initiated.
Their analysis follaws and indicates that DNB is not strongly challenged, and hence after the safety grade trips occur, reactor operators may take positive action.
A Minimum DNBR of 1.8 was calculated using the XCOBRA-IIIC computer code and the EPRI-1 (EPRI-NP-2609, "Parametric Study of CHF Data", Electric Power Institute, Palo Alto, CA, September 1982) critical heat flux correlation at core conditions which conservatively bound those expected to occur for the four pump coastdawn event fram below the.P-7 interlock setpoint.
This result indicates that DNB is not stmngly challenged during the event.
Core bound;~
conditions employed are:
- 1) a reactor paver of 20% of rated,
- 2) a coolant flaw rate of 6.9% of design,
- 3) an inlet temperature of 560 degrees F, and 4) a core exit pressure of 2202 psia.
The Technical Specification limit on delta H for zero pawer of 1.86 was employed along with the axial power profile given in Figure 15.0. 3-1 of XN-NF-85-64 (P), Rev. l.
Question 5 inquires about conditions and pcarer levels related to the P-7 setpoint.
P-7 is an interlock which is developed fram either the P-10 interlock or the P-13 interlock.
P-10 is in turn developed fram the Pcarer Range Nuclear Instnmemtation System and is calibrated to a setpoint of 10%
reactor pawer.
P-13 is developed fram a turbine power signal and is calibrated to a setpoint equivalent to 10% turbine power.
The phrasing of the question dealing with the 9% uncertainty tolerance indicates that the concern is with the establishment of P-7 as developed fram P-10.
The initial condition prescribed; i.e., the reactor critical and pawer below the P-10/P-7 setpoint, constrains this accident to a transient operating condition, either during turbine roll and generator paralleling operations or during shutdown operations.
During these transient conditions, two reactor operators and a supervisor are required to be present in the control roam; typically a third operator is also present to assist in the operations conducted at these times.
Since operation in these transient modes requires the active participation of the operating staff present and not merely the monitoring of plant conditions, particularly close attention is paid to all conditions and indications. If a total loss of flow occurred, it would be noted and the appropriate action instituted including tripping the reactor.'ripping the reactor is an action specified in the applicable procedures.
We believe that the increased number and awareness of the operators while in these traEent modes ensures that the reactor will be tripped very quickly follawing the lass of Flaw transient and therefore reactor power willnot increase substantially during the accident.
Our response to this question is sunmnarized as follaws:
1)
This prescribed condition constrains this accident to transient operating conditions.
AEP:NRC:0916T Attachment 1
Page 3 of 3 2), The only credible cause of this accident is a Loss of Offsite Poorer which will result in a Blackout condition.
3)
This Blackout will result in a non-safety grade reactor trip approximately two (2) seconds following the Blackout initiation.
4)
Xf no credit is taken for the non-safety grade trip, 'the contml rods willfall into the core a maximum of 102 seconds following the Blackout initiation.
5)
There will be approximately 8% coolant flow 102 seconds after the Blackout initiation.
6)
Exxon has performed an analysis which demonstrates that ENB is not strongly challenged during this 102 ~nd period following the blackout.
7)
The plant operating conditions during the times this accident could occur are such that typically there are three reactor operators and one s~~isor present.
They would notice a Blackout condition and respond in accordance with approved p~ures which specify ensuring the control rods are inserted.
ATTACHMENT 2 TO AEP:NRC:09l6T PROPOSED REVISED TECHNICAL SPECIFICATION PAGES
INDEX LIMIT NG CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 34 2
3/4.2.1 3/4.2.2 3/4.2.3 3/4.2.4 3/4 '.5 3/4.2.6 POWER DISTRIBUTION LIMITS Axial Flux Difference Heat Flux Hot Channel Factor Nuclear Enthalpy Hot Channel Factor Quadrant Power Tilt Ratio DNB Parameters MODE 1 MODES 2, 3,
4 and 5
Allowable Power Level 3/4 2-1 3/4 2-5 3/4 2-9 3/4 2-13 3/4 2-15 3/4 2-17 3/4 2-19 343 INSTRUMENTATION 3/4.3.1 3/4.3.2 3/4.3.3 3/4.3.4 REACTOR TRIP SYSTEM INSTRUMENTATION ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation Movable Incore Detectors Meteorological Instrumentation Remote Shutdown Instrumentation Post-Accident Instrumentation Fire Detection Instrumentation Radioactive Liquid Effluent Instrumentation Radioactive Gaseous Process and Effluent Monitoring Instrumentation TURBINE OVERSPEED PROTECTION 3/4 3-1 3/4 3-14 3/4 3-34 3/4 3-38 3/4 3-39 3/4 3-42 3/4 3-45 3/4 3-50 3/4 3-53 3/4 3-58 3/4 3-65 3 4 4 REACTOR COOLANT SYSTE 3/4.4.1 REACTOR COOLANT LOOPS Normal Operation 3/4.4.2 SAFETY VALVES -
SHUTDOWN 3/4.4.3 SAFETY VALVES - OPERATING 3/4.4.4 PRESSURIZER 3/4.4.5 STEAM GENERATORS 3/4 4-1 3/4 4-4 3/4 4-5 3/4 4-6 3/4 4-7 D.
C.
COOK - UNIT 2 IV AMENDMENT NO.
POWER DISTRIBUTION LIMITS DNB PARAME ERS MODES 2
3 4 and 5
LIMITING CONDITION FOR OPERATION 3.2.5.2 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-2:
a.
avg'.
Pressurizer Prcssure'PPLICABILITY.
MODES 2 3*
4* and 5*
ACTION:
MODES 2 and 3*
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or open the reactor trip system breakers within the next hour.
MODES 4* and 5*
Within one hour either open the reactor trip system breakers or render the control rod drive system incapable of rod withdrawal.
SURVEILLANCE RE UIREMENTS 4.2.5.2 Each of the parameters of Table 3.2-2 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal.
D.
C.
COOK - UNIT 2 3/4 2-17 AMENDMENT NO.
TABLE 3 2-2 DNB PARAMETERS
~PAEAEETE Reactor Coolant System Tavg Reactor Coolant System Tavg Pressurizer Pressure LIMIT
< 549.2 F.
(Reactor Subcritical)
< 576.3 F.
(Reactor Critical) 0
) 2176 psig Reactor coolant loop operational requirements are contained in Specifications 3.4.1.1, 3.4.1.2.c and 3.4.1.3.c.
D.
C.
COOK - UNIT 2 3/4 2-18 AMENDMENT NO.
REACTOR COOLANT SYSTE SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a.
The coolant loops listed below shall be OPERABLE and in operation as required by items b and c:
1.
Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump,*
2.
Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,*
3.
Re'actor Coolant Loop 3 and its associated steam generator and reactor coolant pump,*
4.
. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump,*
5.
- East, **
6.
- West **
b.
At least two of the above coolant loops shall be OPERABLE and at least one loop in operation if the reactor trip breakers are in the open position, or the control rod drive system is not capable of rod withdrawal.***
At least three of the above reactor coolant loops shall be OPERABLE and, in operation when the reactor trip system breakers are in the closed position and the control rod drive system is capable of rod withdrawal.
APPLICABILITY:
MODES 4 and 5
ACTION:
a.
With less than the above required loops OPERABLE, immediately initiate'orrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
b.
With no coolant loop in operation, suspend all operations involving a reggggion in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.
D.
C.
COOK - UNIT 2 3/4 4-3 AMENDMENT NO.
0 fl l.
e C,
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