ML20211N540

Safety Evaluation Supporting Issuance of Amend 88 to License DPR-6

ML20211N540
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	02/17/1987
From:	Office of Nuclear Reactor Regulation
To:
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ML20211N497	List: ML20211N491 ML20211N518 ML20211N540 ML20211N549
References
NUDOCS 8703020110
Download: ML20211N540 (6)

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UNITED STATES '

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SAFETY EVALUAT1,0N BY THE OFFICE OF Nd[ LEAR REACTOR REGU1ATION

, SUPPORTINGAMEN0MENTNO.88g0FACIL'ITYOPERATINGLICENSENO.DPR-6 CONS'MERS U POWER COMPANY BIG HOCK POINT PLANT

D0CKET NO. 50-155 .

1.0 INTRODUCTION

l l By letter dated December 5,1986 (Ref.1) from K. W. Berry, Consumers Ptwer

{ Company (CPC), to the Director.uf NRR,'NRC, CPC proposed changes to the Big i Rock Point Plant (BRP) Technical Specification Section 5.1 to incorporate a new design of control blades. The new design provides for replacement of the  ;

top one quarter of the. boron carbide absorber rods with hafnium absorber rods l

in the outer two rods on each control blade ' wing and the use 'of a nu cladding-

! material intended to eliminate intergranular corrosion cracking of absorber i tubes containing boron carbid6. The new design is expected to increase control '

rod assembly life. In addition, the licensee proposed changes to the core 1 operating limits Tables 1 and 2 of Technical Specification Section 5.2.1 to ,

1 include Reload I-2 fuel for Cycle 22 operation. Thase proposed changes to l Section 5.2.1 of the Technical Specifications were found to be unacceptable

and were denied in our letter, dated bbruary 12, 1987. I l

During our preliminary revir.1 of Reference 14 the staff infor1ned the licensee of a need for more detailed inform tion regarding the hafnium hybrid control i

-l rod design and for additional information regardino the Cycle 22 reload and.

! safety analysis which concluded that core operatini; limits defined in the Technical Specifications 'will be satisfied. W licensee responded with i submittals dated January 20 and February 4,19e17 (Ref. 2 and 3) from CPC to i

the NRC. ,

) This report describes the staff's review of the above submittals and provides  !

the staff's safety evaluation of the CPC requested changes to Section 5.1 cf [

! the BRP Technical Specifications concerning the use of new hybrid control rods. L i  !

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2.0 BACKGROUND

i '2.1 Core Description .

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BRP is the oldest commercial boiling water reactor (BWR) still in operation.

having started up in 1962. The plant is a BWR-1 rated at 240 MWt or about l 72 MWe (gross), but has been restricted to a lower power level during much of l

' its operating histnry because of restrictive core thernel 1Imits.

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e F Although BRP is fundamentally similar to current BWR designs, there are some significant differences in design and mode of operation:

1. The BRP reactor core is small, the active region being about six

feet in height and six feet in diameter. An advantage of this is a very stable, leakage controlled power distribution as compared to s- modern plants whose core volume is about eight times larger. To 2 compensate for the high leakage associated with the small core, reactivity (K-infinity 1 and hence fuel enrichment must be higher than for most later plants.

2. Although shorter, PRP fuel assemblies are wider than modern plants (71" pitch versus ~ 6"1. The BPP 11 x 11 assembly is roughly the

same in rod diameter and pitch as the modern 8 x 8 RWR assembly.

Because of the larger assembly, the ratio of control rods to interior assemblies is one to two rather than the typical one to four, i.e. a "D" lattice.

! s, 3. BRP has external recirculation loops with constant velocity pumps, therefore flow control is not employed, and maneuvering is,done entirely with control rods. This is a disadvantage as far as plant flexibility, but greatly simplifies predictive physics analysis and j power distribution surveillance.

4. BRP has only 37 control rods, as opposed to around 200 in the large modern plants. Since the reactivity inventory is about the same as a larger plant, individual control rod worths are generally larger

for BRP. During operation, banking the control rods in groups of greater than two rods would result in unacceptable axial power shapes, so that X - Y symmetry is limited to half core rotational, rather than quadrant or octant.

5. LPRMs are present, but they are not part of the reactor protective

system. A high flux trip is provided by three excore detectors.

6. In-core power distribution measurements are i

activation of flux wires, rather than a movabrovided by the There le TIP detector. ,

i are only eight measurement locations arranged in four symmetrically l located pairs. These are employed to verify calculated axial power 1

shapes, but because of the small number of locations, they are not

considered useful for radial power measurements.

< 7. The primary coolant system is pressurized to 1350 psi versus the typical 1000 psi. Maximum exit void fractions are about 55%, which is much lower than modern plants.

8. There is no on-line power distribution monitoring system comparable

, to later plants. The LPRMs are used to monitor changes in power

! distribution, but there is not an on-line thermal margin calculation.

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The Cycle _22' core loading consists of 20 fresh I-2 assemblies (identical to I-1), 20 once-burned I-1 assemblies, 20 twice-burned H-4 assemblies, 18 5 twice-burned H-3 and H-2 assemblies, two four-times-burned H-1 assemblies and four reconstituted H-2 assemblies which were irradiated earlier for two cycles. A descriptfon of the fuel comprising the Cycle 22 reload is provided e

inTa.ble1(fromRef.2). ,

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, ' TABLE I ,

CYCLE 22 CORE LOADING t  ;

FUEL NO. CYCLES NO.0F INITIAL U235 1HITIAL INITIAL BOC 22 **

TYPE IN CORE ASSEMBLIES ENRICHMENT FUEL DENSITY RADIAL GAP (MIL) EXP(GWD/T)

H-1 4 2 3.15 .93.5 9.5 19.6 i H-2 3 2 3.43 93.5 9.5 16.1 H-3 3 16 3.43 94.0 9.5 16.6 A

H-2 ,2 4 3.43 93.5 9.5 9.6 7

I H-4 2' ,

20 3.43 94.0 9.5 12.0

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, I-I 1 20 3.43 94.0 7.5* 5.7 l' I-2 'O 20 3.43 ;94.0 ;7.5 0 s

• Excludes Special Test Rods

• • f Based upon outage starting 10/24/86 i

3 A Reconstituted Bundle; H201, H202, H204, H205 l 3.0 EVALUATION OF PAFNIUM HYBRID CONTROL BLADE DESIGN The major design changes of the control blades are as follows:

1. Replacement of thei top one quarter of the outer two poison rods in each control blade wina with solid hafnium rods. The solid hafnium rods are seventeen (17) inches long and have the same outside diameter as the stainless steel boron carbide (B4 C) filled tubes being replaced.

2. Use of an improved B4 C absorber rod cladding from presently used 304 l

I stainless steel to high purity 348 stainless steel.

In addition, there are other associated material changes of the pin / roller materials from Haynes 25/ Stellite 3 to PH13-8 Mo/Inconel X-750 to eliminate cobalt-bearing stellite materials (Ref. 4).

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Although there exist material differences and slight differences in nominal sheathing thickness between the NUCOM design and GE design control blades (0.055" vs 0.056"), the sub-component and overall dimensions are the same.

Also, the NUCOM control blade is slightly ligher (approximately 2 lbs) than

- the GE control blade, making scram times for the NUCOM blade approximately the same as or slightly faster than scram times for the GE blade. Therefore, scram time requirements of the Bio Rock Point Technical Specifications 5.2.2.(a) and 5.1.3 will not be altered. Hafnium exhibits no irradiation induced swelling and proper tolerances ensure that the hafnium will not s induce any stresses in the stainless steel sheathing during irradiation.

The use of hafnium in control rod blades has previously been approved for test blades in Pilgrim Nuclear Power Station Unit 1, Peach Bottom and Dresden Unit 3. The staff is unaware of any material problems associated with the use of hafnium and, therefore, this aspect of the design is acceptable.

H There is no direct opqrating experience to demonstrate acceptable performance with new control blade design. At the request of the staff, the licensee has

, defined a surveillanc6 program to confirm the Cycle 22 performance of the six

hybrid control rod blz. des inserted for Cycle 22 operation . The licensee

, surveillance testing program for the NUCOM control rods will include measure-ment of the control red drive scram times prior to start-up and verification

that these scram times meet the Technical Specification acceptance criteria.

At the end of the cycle, visual examinations will be performed on all six of

- the control rod blades to verify overall integrity. The visual examinations will be conducted underwater with a self-lighted TV camera looking for indications of swelling, cracks, corrosion or any other abnomalities.

Depending on the results of the visual observations, dimensional measurements of the blade outer thickness may also be performed.

Since centrol rod performance is affected by residence time and burn-up, the licensee has also committed to define a continuing surveillance l program for future cycies based on results of the Cycle 2? inspections and i to submit the program for NRC approval prior to continuing operation with these control blades after Cycle 22 operation. Further discussions on the design lifetime, including reactivity and mechanical design criteria of the new hafnium hybrid control blades, will also be provided at that time.

On this basis, we conclude that the use of the hafnium hybrid control blades is acceptable for Cycle 22 operations.

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e 4.0 ,P_R0 R POSED CHANGES TO TECHNICAL SPECIFICATION Section 5.1.1 Current Proposed fontrol Rods - Either B4 C filled Control Rods:

304 SS Tubes and Solid 304 SS Neutron Absorbing Solid Hafnium (Hf)

Rods or B4 C filled 304 SS Tubes Material or Boron Carbide with 304 SS Tubes in Cruciform- B C Powder 4

Shaped 304 SS Sheaths B4 C Cladding Material 304 or 348 SS Dummy Rod / Tube Material 304 or 348 SS Assembly Sheathing 304 SS Material Section 5.1.2 Current Proposed The reactor core shall contain The reactor core shall contain 32 32 control rod assemblies each control rod assemblies each composed consisting of a cruciform array of a cruciform array of empty or of empty or solid stainless solid stainless steel tubes, boron steel tubes containing approximately carbide (B4 C) powder filled stainless 68 inches of boron carbide, tubes and/or solid hafnium (Hf) rods 84C, powder surrounded by a surrounded by a crucifom shaped cruciform-shaped stainless stainless steel sheath. The individual steel sheath, poison rod composition made up of principal materials boron carbide and/or hafnium shall have an effective poison length of approximately 68 inches.

5.0 ENVIRONMENTAL CONSIDERATION

l This amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in l the types, of any effluents that may be released offsite and that there is no i significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement nor environmental assessment need be prepared in connection with the issuance of this amendment.

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6.0 CONCLUSION

The staff has reviewed the Technical Specification change request and the additional information submitted by CPC in response to staff requests. On the basis of its review, the staff has concluded that the proposed changes to Technical Specification Sections 5.1.1 and 5.1.2 to permit the inclusion of hafnium hybrid control rods in the Cycle 22 core is acceptable. The staff also accepts the licensee's commitment to reevaluate the acceptability of these control rods by inspection after Cycle 22 operation and to submit an acceptable surveillance program for future cycles at that time.

The staff has found the proposed changes to Technical Specification Section 5.2.1, Tables 1 and 2 (defining the operating limits for the new reload I-2 fuel),tobeunacceptable. The staff's evaluation and denial of this proposed

, change is contained in a separate licensing action.

The staff has concluded, based on the considerations discussed above, that:

1 (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common' defense and security nor to the health and safety of the public.

7.0 REFER.ENCES

1. Letter from K. W. Berry, Consumers Power Company, to the Director, Nuclear

Reactor Regulation, NRC, December 5, 1986. .

2. Letter from R. R. Frisch, Consumers Power Company, to the Nuclear Regulatory Commission Document Control Desk, January 20, 1987

3. Letter from K. W. Berry, Consumers Power Company, to the Nuclear Regulatory Commission Document Control Desk, February 4,1987.

4. EPRI NP-2329 BWR Control-Rod Cobalt-Alloy Replacement, General Electric Co., San Jose, CA, Nuclear Engineering Division, March 1982.

i Principal Contributors: U. Cheh and C. Hinson Dated: February 17, 1987 I

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