ML20072T045

Proposed Tech Spec Reflecting Changes to Reactor Core Design for Cycle 4

ML20072T045
Person / Time
Site:	Hatch
Issue date:	03/30/1983
From:	GEORGIA POWER CO.
To:
Shared Package
ML20072T041	List: ML20072T037 ML20072T045 ML20072T049
References
TAC-49989, TAC-51054, NUDOCS 8304070421
Download: ML20072T045 (26)

Text

.

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LIEAR HEAT GENERATION RATE ,

[IMITIN3 C0rOITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LIEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE. shall not exceed the limits shown in Figure 3.2.1-1 thru 3.2.1-8.  ;

APPLICABILITY: CONDITION 1, when THERMAL POWER 'p 25% of RATED THERMAL POWER. -

ACTION:

With an APLHGR exceeding the limits of Figures 3.2.1-1 thru 3.2.1-8, initiate' l corrective action within 15 minutes and continue corrective action so that APLHGR is within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREENTS 4.2.1 All APLHGRs shall . be verified to be equal to or less than the applicable limit determined from Figures 3.2.1-1 through 3.2.1-8:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Whenever THERMAL POWER has been increased by at .least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

HATCH - UNIT 2 3/4 2-1 Amendment No.

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MCPR LIMIT FOR 7X7 FUEL AT RATED FLOW FIGURE 3.2.3-3 HATCH - UNIT 2 3/4 2-7b Amendment No.

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MTCH - WIT 2 3/4 2-7e Amendment No. 21

. ' POWER DISTRIBUTION LIMITS

~ 3/4.2.4 LIEAR EAT GENERATION RATE ,

i

LIMITIPC C0tOITION FOR OPERATION I T

- 3.2.4 All LIEAR. EAT GENERATION RATES -(LHGRs) shall not exceed 13.4 Kw/ft 2

for 8X8R/P8X8R fuel or 18.0 Kw/ft:for 7X7 fuel.

~

, ' APPLICABILITY: ~ CONDITION 1, when' THERMAL POWER 2t25% of RATED THERMAL POWER.

• ~~

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ACTION: .

j

- With the LHGR of, any fuel- rod exceeding the. limit, initiate corrective action within 15 minutes and continue corrective action so~that the LHGR is within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 'or . reduce' THERMAL POER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4 2

SURVETlI APCE REQUIREENTS -

t . . .

j - 4.2.4 LHGRs shall be determined to be equal to or less than the limit;

. a. -At-least'once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. . When THERMAL ' POWER has been increased by at least 15% . of RATED THERMAL -' POWER : and steady state ' operating -conditions have been established, and:

i . e

c. Initially and at ' least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITIPC CONTRO ROD PATTERN FOR LHGR.

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. HATCH - UNIT 2 3/4 2-8 . Amendment No.

2

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5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall Le as shown in Figure 5.1.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone coincides with the exclusion area and is also shown in Figure 5.1.1-1.

5.2 CONTAIMENT CONFIGURATION 5.2.1 The primary containment is a steel structure composed of a series of vertical right cylinders and truncated cones which form a drywell. This drywell is attached to a suppression chamber through a series of vents. The

, suppression chamber is a steel pressure vessel in the shape of a torus. The

! primary containment has a total minimum free air volume of 255,978 cubic feet.

DESIGN TEM)ERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:

a. Maximum design internal pressure 56 psig.

b. Maximum allowable internal pressure 62 psig.

c. Maximum internal temperature 3400F.

d. Maximum external pressure 2 psig.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The initial core shall contain 560 fuel assemblies with each fuel assembly containing 62 fuel rods and 2 water rods clad with Zircaloy -2.

Each fuel rod shall have a nominal active fuel length of 150 inches and contain a maximum total weight of 3341 grams uranium. The initial core I loading shall have a maximum average enrichment of 1.87 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum average enrichment of 2.90 weight percent U-235. 7X7 fuel containing 49 fuel rods and no water rods may also be inserted.

HATCH - UNIT 2 5-1 Amendment No.

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 CONTROL RODS The specifications of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) the potential effects of the rod drop accident are limited. The ACTION statements permit variations from the basic requirements, but at the same time inpose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scramtime measurements ensure that any indication of systematic prcblem with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem; therefore, with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.

The control rod system is analyzed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than 1.07 during the l limiting power transient analyzed in Section 15 of the FSAR. This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications provide the required protection and MCPR remains greater than 1.07. The occurrence l of scram times longer than those specified should be viewed as an indication of a systematic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a , M etially serious problem. .

Control rode ei operable accumulators are declared inoperable and Specification L 3 .! n applies. This prevents a pattern of inoperable accumulators tha cue msult in less reactivity insertion on a scram HATCH - UNIT 2 8 3/4 1-2 Amendment No.

4 3/4.2 POWER DISTRIBUTION-BASES The specifications of this section assure that the peak cladding

' temperature following the postulated design basis loss-of-coolant accident will not exceed the 22000F limit specified in the Final Acceptance Criteria - (FAC) -issued. in June 1971 considering the postulated effects of fuel pellet densification.

3/4'.2.1 AVERAGE PLANAR LIEAR HEAT GEERATION RATE 4

- This specification assures that the' peak cladding temperature following

'the postulated design basis loss-of-coolant accident will not exceed the .

limit specified in 10 CFR 50,-Appendix K.

The peak cladding temperature ' (PCT) following a postulated

, loss-of-coolant accident is primarily a function of the average heat

. generation , rate of all the rods of a fuel assembly at any axial location and.

is dependent only secondarily on the rod-to-rod power distribution within an assembly. - The peak clad temperature is calculated assuming an LHGR for the

, highest powered rod which is equal:to or less than the design LHGR corrected

for Edensification. This LHGR times 1.02 is used in the heatup code along -

with the exposure -dependent steady ' state -gap conductance and tod-to rod

! local- peaking factor. The Technical Specification APLHGR is ~ this LHGR of

4. the ' highest powered rod divided by its local peaking factor. The limiting l, -value. for APLHGR is shown~ 1n the figures in Technical Specification 3/4.2.1.

The calculational procedure used to~ establish the. APLHGR shown in .the figures 'in Technical Specification 3/4.2.1, is based on a loss of coolant - -

, accident analysis. ~ The analysis was ~ performed,using General Electric (GE) calculational .models which.-are consistent with ' the'_' requirements 'of

, , Appendix K to 10 CFR 50. A complete discussion of each code employed'in the

- analysis is. presented in Reference 1. -Differences in this analysis compared to . previous analyses performed with. Reference 1 are: (1) the ? analysis assumes a fuel assembly planar. power consistent with -102% of the MAPLHGR shown in the; figures in Technical Specification 3/4.2.1; (2) fission' product

. decay is computed _ assuming an energy release ' rate of 200 EV/ fission;' (3) pool boiling (is . assumed after' nucleate boiling is lost during the flow L stagnation period; and --(4) the, effects. -'of core spray .entrainment and counter-current flow limitiation as ? described in Reference 2, are included in the reflooding calculations.:

k, . ' A list off the significant; plant. input -parameters 'to- t'he loss-of-coolant -

accident analysis presented'in bases Table B 3.2.1-1. ,

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• B 3/4 2-l' ' Amendment No.

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POWER DISTRIBUTION LIMITS BASES MINIh0M CRITICAL POWER RATION (Continued)

The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.1-6 that are input to a GE-core dynamic behavior transient computer program described in NEDO-10802(3). Also, the void reactivity coefficients that were input to the transient calculational procedure are based on a new method of calculation termed NEV which provides a better agreement between the calculated and plant instrument power distributions. The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic SCAT code described in NED0-20566(1). The principal result of this evaluation is the reduction in MCPR caused by the transient.

The purpose of the Kr factor is to define operating limits at other than related flow conditions. At less than 100% of rated flow the required MCPR is the product of the operating limit MCPR and the Kr factor.

Specifically, the Kr factor provides the. required thermal margin to protect against a flow increase transient. The most limiting transient initiated from less than rated flow conditions is the recirculation pump speed up caused by a motor-generator speed control failure.

For operation in the automatic flow control mode, the Kr factors assure that .the operating limit MCPR of Specification 3.2.3 will not be violated should the most limiting transient occur at less than rated flow.

In the manual flow control mode, the Kr factors assure that the Safety Limit MCPR will not be violated should the most limiting transient occur at less than rated flow.

The Kr factor values shown in Figure 3.2.3-4 were developed generically and are applicable to all BWR/2, BWR/3 and BWR/4 reactors. The Kr factors wre derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow.

For the manual flow control mode, the Kr factors were calculated such that the maximum flow rate, as limited by the pump scoop tube set point and the corresponding THERMAL POWER along the rated flow control line, the

' limiting bundle's relative power, was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPRs were calculated at different points along the . rated flow control line corresponding to different core flow. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the Kr.

HATCH - UNIT 2 8 3/4 2-4 Amendment No.

POWER DISTRIBUTION LIMITS BASES

. MINIHN CRITICAL' POWER RATION (Continued)

For operation in the automatic flow control mode, the same procedure was

' employed.except the initial power. distribution was established such that.the MCPR was equal to the operating limit MCPR at RATED THERMAL POWER and rated flow.

The Kr factors shown in Figure 3.2.3-4 are conservative for the

~ General Electric Plant . operation .because the operating limit MCPRs of' Specification 3.2.3 are greater than the original 1.20 operating limit MCPR

- used for the generic derivation of Kr.

At THERMAL POWER levels :less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and'

~

the moderator void content will- be very small. For all designated control rod patterns which may -be employed at this point, operating plant experience .

Indicated that the resulting MCPR value is in excess of requirements by .a considerable margin. With ;this low void content, any inadvertent core flow

increase would only. place operation in a more conservative mode relative to

- MCPR. During-initial: startup testing of the plant, an MCPR evaluation will be .made at 25% .of ; RATED THERMAL POWER with minimum recirculation pump

,~

speed. The . MCPR margin will thus be demonstrated such that future MCPR

_ evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 25% of RATED THERMAL POWER is sufficient 'since power: distribution shifts are very slow when. there have ~ not

~been significant -power: ~or control rod changes. .The requirement for

. calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change _ in THERMAL POWER or power shape,

-regardless of magnitude that could place operation at'a thermal limit.

3/4.2.4 LIEAR HEAT GENERATION RATE t

, The LHGR specification assures that the linear heat generation rate in any rod is .less thanf the ~ design linear heat generation even if fuel pellet .

densification;is. postulated.

' HATCH - UNIT 2' B 3/4 2-5 Amendment No.

m

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ATTACHENT 3 NRC DOCKET 50-366 OPERATIto LICENSE PPF-5 EDWIN I. HATCH NUCLEAR PLANT UNIT 2 PAGE-BY-PAGE DESCRIPTION OF CHANGES 3/42-1 -

Add new Figure numbers 3/4 2-2 -

No Change: Better quality graph 3/4 2-3 -

Extend MAPLHGR curve to higher exposure 3/4 2-4a - Change page number 3/4 2-4b -

Extend MAPLHGR curve to higher exposure; change page number 3/4 2-4c - Extend MAPLHGR curve to higher exposure; change page number l 3/4 2-4d - Add Unit 1 7x7 MAPLHGR curve '

3/4 2-4e -

Add new Unit 2 fuel MAPLHGR curve 3/* 2-4f - Add Unit I reconstituted fuel MAPLHGR curve 3/4 2-6 -

Add' new figure number; reflect new Kr figure number; add reference to 7x7 fuel 3/4 2-7b - Add MCPR limit curve for 7x7 fuel 3/4 2-7c - Change figure number and page number 3/4 2-8 -

Add LHGR for 7x7 fuel 3/4 5-1 -

Change " reactor. core" to " initial core";

correct "Zircaloy-4 to "Zircaloy-2"; add sentence about 7x7 fuel B 3/4 1 Reflect new safety limit of 1.07 8 3/4 2 Condense reference to APLHGR figures B 3/4 2 Reflect change in Kr curve figure number B 3/4 2 Reflect change in Kr curve figure number m

2 4

4 l

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i ATTACHENT 4 NRC DOCKET 50-366 i

OPERATIMI LICENSE M'F-5 EDWIN I. HATCH NUCLEAR PLANT UNIT 2 REFERENCE DOCUENTATION Y1003J01A57,-" Supplemental Reload ' Licensing Submittal for Edwin I. Hatch Nuclear Plant Unit 2, Reload 3 (Cycle 4)", January 1983.

LMQ:83-018, " Hatch 2 ECCS MAPLHGR Calculations for Hatch 1 Initial Core and Reconstituted Hatch 1 Reload 2 Fuel", February 22, 1983 LMQ:83-022, " Hatch 2 Cycle 4 OLMCPR", February 24, 1983.

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G E N E R A L 'h E L E CT R I C AND SPE CI AL GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 PROJECTS D1 VISION February 24, 1983 LMQ:83-022 cc: B. E. Hunt L. K. Mathews H. C. Nix Mr. R. D. Baker Georgia Power Company P.O. Box 4545 Atlanta, GA 30302

Subject:

Hatch 2 Cycle 4 OLMCPR

References:

1) " Supplemental Reload Licensing Submittal for

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Edwin I. Hatch Nuclear Plant Unit 2 Reload 3

\ (Cycle 4)", Y1003J01A57, Rev. O, 1/83

2) " General Electric Boiling Water Reactor Load Line Limit Analysis for Edwin I. Hatch Nuclear Plant Unit 2", NEDO-24295, 10/80, as amended

3) " Safety Review of Hatch Nuclear Power Station Unit No. 2 at Core Flow Conditions above Rated Core Flow Throughout Cycle 2," NEDO-24292, Rev, 2 i 10/81

Dear Mr. Baker:

The operating limit minimum critical power ratios (OLMCPR's) documented in Reference 1 for Hatch 2 Cycle 4 are based on operation at rated flow and final feedwater temperature and assume a full-arc turbine control valve configuration in the transient analyses. These results bound both operation above the 100% power /100% flow load line at rated final feedwater temperature and a partial-arc turbine control valve configuration.

The OLMCPR's in Reference 1 are also sufficiently conservative to bound operation in Cycle 4 before nominal end-of-cycle at increased core flow (up to 105% of rated) and/or final feedwater temperature reduction (up to 63 F AT, on or below the 100% power /100% flow load line) conditions on an intermittent basis, full-arc or partial-arc turbine control valves.

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h GENER AL $ ELECTRIC Mr. R. D. Baker February 24, 1983 This conclusion is based on the results of analyses performed by General Electric for Cycle 4, final documentation of which will be included in a revision to Reference 1. If you have any questions on this subject, please do not hesitate to contact us.

Very truly yours,

k. w c.--

L. M. Quintana

,. Fuel Project Specialist Hatch 1 & 2 M/C 174; -(408) 925-2026 Pm t- C. If Nilf &Lo

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GENER AL h ELECTRIC AND SPECI AL GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CAllFORNIA 95125 PROJECTS D1 VISION February 22, 1983 LMQ:83- 018 cc: R. D. Baker '

B. E. Hunt H. C. Nix R. T. Schellinger Mr. L. K. Mathews Southern Company Services P. O. Box 2625 Birmingham, AL 35202 Subj ect: Hatch 2 ECCS MAPLHGR Calculations for Hatch 1 Initial Core and Reconstituted Hatch 1 Reload 2 Fuel

Dear Mr. Mathews:

\ Per Mr. K. S. Folk's verbal request, General Electric Company has

\ performed an analysis of the ECCS MAPLHGR limits for the Hatch 1 initial core and reconstituted Hatch 1 Reload 2 fuel bundles when inserted in Hatch 2. This analysis was required because, although the Hatch 1 and Hatch 2 plants are similar, the Hatch 2 plant ECCS analysis is slightly more severe due to a different ibniting break size and associated core uncovery and reflood times. The results of this analysis are documented in the attached tables.

The three Hatch 1 7x7 initial core fuel types' MAPLHGR's and PCT's are represented in the tables as one fuel type since bounding calculations were made to cover all three. These calcultions were made for 80 mil channels only since local peaking data for these 7x7 lattice with 100 mil channels are not available. However, the trend observed from the GDRB265-6G3.0 results for 80 versus 100 mil channels shows less than a 0.1 Kw/ft MAPLHGR difference between the two, with the 80 mil channel results being slightly worse. It is concluded that the MAPLHGR limits for the 7x7 l bundles based on 80 mil channels are applicable to 100 mil channels as well, should the thicker channels be required.

The un-reconstituted local peaking factors for the 8DRB265-6G3.0 bundle were used for this analysis since the reconstituted bundle has slightly better ECCS performance. (The reconstituted bundle has better thermal conductivity and local peaking characteristics than the original bundle.) Although the results are not very sensitive to 80 or 100 mil channels, the reported 8DRB265-6G3.0 MAPLHGR results bound the use of either channel.

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] GENERAL $ ELECTRIC If you have any questions concerning this subject, please do not hesitate to contact us.

Very truly yours,

- 4L-L. M. Quintana F' el Project Specialist Hatch 1 & 2 M/C 174; (408) 925-2026 Pm Enclosures

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2 GENERAL ELECTRIC NUCLEAR ENERGY ENGINEERING DIVISION

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HATCH-2 R3/C4 ECCS REV.1

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, I MAPLHGR TABLE FOR BUNDLE TYPE: HATCH-1 I.C. TYPE 1,2,3

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EXPOSURE MAPLHGR PCT LOCAL OXIDATION (GWD/ST) (GWD/MT) (KW/FT) (DEG-F) (FRACTION)

.20 .22 14.30 2198. 0.036 1.0 1.1 14.40 2198. 0.035 5.0 5.5 14.70 2197. 0.032

10. 11. 14.40 2199. 0.031

15. 17. 14.00 2198. 0.057

20. 22. 13.80 2199. 0.057

25. 28. 14.00 2198. 0.052

30. 33. 12.70 2013. 0.017

35. 39. 11.50 1886. 0.010 j \

40. ,. 44. 10.20 1729. 0.006 II MAPLHGR TABLE FOR BUNDLE TYPE: 8DRB265-6G3-80M/100M '

EXPOSURE MAPLHGR PCT LOCAL OXIDATION (GWD/ST) (GWD/MT) (KW/FT) (DEG-F) (FRACTION) i

.20 .22 11 50 2164. 0.031 1.0 1.1 11.60 2167. 0.031 5.0 5.5 11.90 2199. 0.034 i

L 10. 11. 11.90 2200. 0.033

15. 17. 11 90 2200. 0.034

20. 22. 11 70 2198. 0.034

~' 0.029

25. 28. 11 30 2147.

30.- 33. 10.70 2067. 0.022

. 35. 39. 10.20 1994. 0.017

• 0.012

40. 44. 9.50 1904.

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