ML17318A778
| ML17318A778 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/13/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17318A777 | List: |
| References | |
| NUDOCS 8005280191 | |
| Download: ML17318A778 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATIOf< BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. lg TO FACILITY OPERATING LICENSE NO.
DPR-74 INDIANA AND MICHIGAN ELECTRIC COMPANY DONALD C.
COOK NUCLEAR PLANT UNIT NO.
2 DOCKET NO. 50-316 A.
Maximum Tav at Rated Thermal Power 1.0 Introduction In Reference 1, Indiana and Michigan Electric Company requested a
Technical Specification change for D. C. Cook Unit 2 which involves changing the maximum Tavg at Rated Thermal Power used for the overtemperature and overpower bT trips and DNB limits.
Supporting analysis for this change was submitted as Attachment A to Reference 2.
Revised pages and additional information were submitted in Reference 3.
2.0 Discussion The thermal overpower hT {OPbT) and overtemperature bT
{OT hT) reactor trips provide protection against departure from nucleate boiling (DNB) and fuel centerline melting (excessive kw/ft) respectively during postulated transients.
The methodology for derivation of the limiting safety system settings is presented in Reference 4.
2.1 Overtem eI.ature Tri OThT trip is designed to ensure operation within the DNB design basis and the hot leg boiling limit.
The setpoint is a function of a vessel bT, vessel average temperature primary system pressure and axial flux difference.
The static equation is:
BTsp
= kl - k2 {Tavg - T') + k3 {P - P') - F(hI) where:
GTsp
= OTBT setpoint (X of full power hT) kl = a preset, manually adjustable bias k2
= a constant based on the effect of temperature on design limits k3 = a constant based on the effects of pressure,
(/ full power/psi)
Tavg 'verage reactor coolant temperature (oF) 8OO5 ~So lfl
T' Nominal Tavg at full power
( F)
P
= Pr essurizer Pressure (psig)
P' Nominal reactor coolant system pressure (psig) f(hI) = A function of the neutron flux difference between upper and lower long ion chambers A dynamic term is also included to compensate for inherent instrument delays and piping lags between the core and the loop temperature sensors.
For the OTaT setpoint a lead/lag function
~~l + T S l is appl ied to the k2(Tavg - T' term.
1
+ TpS The OPBT trip is designed to prevent fuel centerline melting with an overpower setpoint of 118Ã of the nominal full power hT.
A compensating term is included since thermal power is not precisely proportional to hT, as well as a flux difference term.
The static equation is:
BTsp =
K4 - K6 (Tavg T )
F (aI )
where:
BTsp = overpower bT setpoint (X of full power hT)
K4 = a present, manually adjustable bias (A of full-power hT)
K6 = a constant that accounts for the effects of coolant density and heat capacity on the relationship between hT and thermal power (5 of full power hT/oF)
T' indicated average reactor coolant temperature at full power (oF)
Tavg
= average reactor-coolant temperature (oF)
F(h,I)
= a function of the neutron flux difference between upper and lower long ion chamber section (5 full power hT).
Increases in hT beyond a predefined deadband result in a decrease in the trip setpoint.
The dynamic term for the OPh T setpoint is
-Kg ~1+T S Tavg where:
K5 = a constance that compensates for piping and thermal time delay T3S/1+T3S
= rate/lag function Tavg
= average reactor coolant temperature 2.3 DNB Limits The DNB parameters limits in the T.S. assure that each parameter is maintained within the normal steady-state envelope of operation'ssumed in the safety analysis.
3.0 Pro osed Modifications The licensee has proposed the following changes:
a.
For OTET setpoint (pps.
2-7 and 2-8, Table 2.2-1 of T.S.)
1.
T'increased from 572.2 F to 573.8 F
2.
kl (4 loop operation) decreased from 1.36 to 1.334 3.
kl (3 loop operation) decreased from 1.142 to 1.116 b.
For'OPBT setpoint (p. 2-9, Table 2.2-1 of =T.S.)
1.
T'ncreased from 572.2 F to 573.8 F
c.
For DNB Parameters (Table 3.2-1 of T.S.)
4 loop operation Tavg limit increased from 576.2 F to 578 F
2.
3 loop operation Tavg limit increased from 569.8oF to 570oF
- 4. 0 Evaluation S.l A~nal sls Revised analyses were provided for transients potentially affected by the proposed changes.
These events include:
Rod"Withdrawal Loss of Flow - 4 pump coastdown Loss of Flow -
1 pump coastdown, 4 loops in operation Startup of an Inactive Loop Loss of Load Loss of Feedwater Excessive Heat Removal due to Feedwater Systems Malfunctions Excessive Load Increase Steam Line Rupture The input parameters, codes and evaluation models used were the same as those used in the FSAR (Reference
- 5) except as discussed below.
The OT<T and OPgT setpoints are as given in Section 3.0.
The maximum Tavg was increased from 576.2'F {572.2'F {nom) + O')
to 578'F (573.8'F
{nom)
+ O', rounded).
The three loop maximum Tavg was 570'F.
Nominal Tavg (used in analysis of most events with the Westinghouse Improved Thermal Design Basis (Reference 7))was 573.8'F.
The loss of load event was analyzed with the Improved Basis for DNB evaluations and also with the maximum temperature and pressure to ensure that the peak pressure remains below the limit.
The results show that only small changes occur for most events The minimum DNBR occurs for the rod withdrawal event, with a MDNBR of 2.0.
The design limit for Cook 2 with the WRB-1 correlation (Refer-ence
- 6) is 1.8.
The steam line break analysis was also reviewed to ensure that the return to power following the scram does not violate any core thermal limits.
The effect of a temperature increase of 1.8'F (578-576.2'F) on steady-state DNBR is a reduction of
.04, so that the steady-state DNBR at the maximum allowable temperature remains well above the limit.
4.2 'echnical S ecification Chan es For the static OT~T equation, the proposed changes to T 'nd Q
balance each other so that the ~Ts is unchanged.
The dynamic effect hs a higher esp for any particular value of Tavg.
However, since the event starts at a higher average temperature, the hT rise before reaching the trip is less than previously and, in fact, the analysis shows that a reactor trip occurs slightly earlier into the transient than it did with the existing setpoints.
The OPNT setpoint change allows operation at the 118Ã overpower limit for Tavq up to 573.8'F, with the setpoint decreasing after that point.
This is consistent with the safety analysis for full power operation at a nominal core average temperature of 573.8'F.
The DNB limits are simply rounded off values based on nominal temperatures plus 4'P for indication error, Periodically, plant conditions as determined by instrument readout are compared to these limits, and adjusted as required so that steady-state operation is maintained within the bounds of the anal-ysis.
The effect of the roundof+ (.2'F) is negligible.
5;0 References 1.
Letter from J.
E.
Dolan (AEP) to H.
R.
Denton (NRC), Serial Number AEP:
NRC:
- 00297, November 2, 1979.
2.
Letter from J.
E.
Dolan (AEP,) to H.
R.
Denton (NRC), Serial Number AEP:
NRC:
00297A, December 11, 1979.
3.
Letter from J.
E.
Dolan (AEP) to H.
R.
Denton (NRC), Serial Number AEP:
NRC:
- 00297D, March 18, 1980.
4.
WCAP-8745 "Design Bases for the Thermal Overpower hT and Thermal Overtemperature bT Trip Functions,"
March 1977.
I
5.
Indiana and Michigan Power Company "Donald C.
Cook Nuclear Plant Final Safety Analysis Report," w'ith Amendments 1-79.
6.
WCAP-8762 "WRB-1 Critical Heat Flux Correlation 7.
WCAP-8567 "Improved Thermal Design Procedure" B.
Deletion of License Conditions In Supplement No.
7 to the D. C. Cook Nuclear Plant, Unit 2 Safety Evaluation Report dated December 1977, the Staff's evaluation concluded that two license conditions were required to support the Staff's findings of adequate design and operation.
The license conditions were as follows:
2.C.(3)(i)
Leak Rate Testin of Containment Isolation Valves Indiana and Michigan Power Company shall install prior to star tup, following the first regularly scheduled refueling outage, test connections to allow Type C leak testing of containment isolation valves.
Indiana and Mjchigan Power Company shall modify the containment isolation valves in the component cooling water system which are identified by and associated with the following containment penetration numbers:
CPN 25( 1), 25(2), 25(3), 25(4),
72(l), 72(2), 72(3) and 72(4).
Indiana and Michigan Power Company shall modify to allow pneumatic leak rage testing the isolation valves identified by and.associated with the following containment penetration numbers:
CPN 38, 39 and 56.
2 C (3)(m) 600 Volt Containment Power penetrations Indiana and Michigan Power Corroany shall modify the 6pp volt containment electrical power penetration circuits to meet the requirements of Regulatory Guide 1.63 prior to startup following the first regularly scheduled refueling outage.
This modification consists of the installation of redundant circuit breakers in the 600 volt containment electrical power penetration circuits to protect the penetration seals by a trip of the 600 volt switchgear breakers in the event of a failure of the molded case circuit breakers.
Further discussion and evaluation of these requirements is not required beyond that which is included in the Safety Evaluation Report, Supplement No. 7.
In the inspection report number 50-315/80-02; 50-316/80-.02 dated February 11, 1980, the Office of Inspection and Enforcement Resident Inspector reported that the license conditions had been met as evidenced by his review and inspection of the required modification and installations.
Therefore it is recommended that the license conditions 2.C.(3)(i) and (m) be deleted.
C.
Environmental Consideration We have determined that the amendments do not authorize a change in 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 amendments involve an action w'Hich is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or ne'gative declaration and environmental impact appraisal need not be prepared in con-nection with the issuance of these amendments.
D.
Conclusion We have concluded that the proposed changes to Tave and to the Technical Specifications are acceptable and the license conditions 2.C.(3)(i) and (m) should be deleted.
We have also concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a siqni-ficant increase in the probability or consequences of accidents previously considered and do not involve a,significant decrease in a safety margin, the amendments do 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 Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Oate:
May 13, 1980
0-01 UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO., 50-316 INDIANA AND MICHIGAN ELECTRIC COMPANY NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERA ING LICENSE 1
The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No; 19 to Facility Operating License No.
DPR-74 issued to Indian and Michigan Electric Company (the licensee),
which revised Technical Specifications for operation of Donald C.
Cook Nuclear Plant, Unit No.
2 (the facility) located in Berrien County, Michigan.
The amendment is effective as of the date of issuance.
The amendment revises the maximum Tavg at Rated Thermal Power used for.the overtemperature and overpower delta T trips and departure from nucleate boiling (DNB) limits.
The amendment also deletes license conditions which required modifications for leak testing certain containment isolation valves and installation of qualified 600 volt containment power penetration circuits.
These modifications and installations have been completed.
The application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations.
The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments Prior public notice of the amendment was not required since the amendment does not involve a 'significant hazards consideration.
- The Commission has determined that the issuance of the amendmept will not.result in any significant environmental impact and that pursuant to 10 CFR 551.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with issuance of the amendment.
For further details with respect to this action, see (1) the application for amendment dated November 2, 1979 as supplemented December ll, 1979 and March 18, 1980, (2) Amendment No. 19 to License No. DPR-74, and
('3) the Commission's related Safety Evaluation.
All of these items are available for public inspection at the Commission's Public Document
- Room, 1717 H Street, N.W., Washington, D.
C.
and at the Maude Reston Palenske Memor'ial Library, 500 Market Street, St.
- Joseph, Michigan 49085.
A copy of items (2) and (3) may be obtained upon request addressed to the U.
S. Nuclear Regulatory Commission, Washington, D.
C.
20555, Attention:
Director, Division of Licensing..
Dated at Bethesda, Maryland, this 13th day of May, 1980.
0 THE NUCLEAR REhULATORY COtBISSION i
Operating Reactors 8
ch Pl Division of Licensing