ML20099K324

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Responds to 760329 Request for Addl Info Re Single Loop Operation.Info Provided by GE
ML20099K324
Person / Time
Site: Pilgrim, 05000000
Issue date: 04/12/1976
From: Andognini G
BOSTON EDISON CO.
To: Ziemann D
Office of Nuclear Reactor Regulation
Shared Package
ML20099K024 List:
References
FOIA-84-105
  1. 76-31, NUDOCS 8411290388
Download: ML20099K324 (8)
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4 BOSTON EDISCN COMPANY Ga=< mas. Cer. css acc Bovssrc= srasst B estc N. MassAce.usttis C:199 April 12, 1975 3ECo. Ltr. #76-31 Director of Nuclear Reactor Regulation AT*I: D. L. Ziemann, Chief Operating Reactors Branch !2 Division of Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Docket No. 50-293 License No. DPR-35 Additional Information Regarding Single Loop Operation Submittal

Dear Sir:

In a letter dated :tarch 29, 1976, you requested additional information con-cerning the request for approval for singla loop operation as submitted by Boston Edison on November 17. 1975. This information has been provided by General Electric Company and is hereby submitted. The questions and 29, responses 1976, letter.

are numbered as were the questions in Enclosure A to your tarch Question 1.

The additional information on the stability analysis pr2sented in Supplement 1 to NEDO-20999 is not clear. Provide a curve of Decay Racio versus Power similar to NEto-2')855-01 Figure 7-19 for single lo:p operation.

Response 1. A rewrite of Section 5 is given below.

5. STABILITY ANALYSIS The least stable power / flow condition attainable under normal conditions occurs at natural circulation with the control rods set for rated power and flow. This condition may be reached following the trip of both re-circulation pumps. Operation along the minimun forced recirculation line with one pump running at minirum
  • speed is more stable than operating with natural circu-lation flow only, but is less stable than operating with

~ both pumps operating at minimum speed. The core stabil-ity along the forced circulation, raced rod pattern line for single loop operation is the same as that for both loops operable except that rated power is not attainable.

Hence, the core is limited to maximum power for single pump operation and only manual flow control should be '

used. This is illustrated in Figure 5-1.

B411290388 840419

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1.2-ULTIMATE STABILITY LIMIT 1.0 1 SINGLE LOOP, PUMP ttINI!!UM SPEED 2 BOTH LOOPS, PUMPS MINIMUli SPEED O

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= FLOW 0.6 - NATURAL CIRCULATION CONTROL LINE LINE

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OPERATIONAL - -

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DESIGN GUIDE / l

0. 2- UPPER LIli1T l FOR SINGLE LOOP OPERATI0tl l
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80 100 0.0 20 40 60 PERCENT POWER FIGURE 5-1 DECAY RATIO VERSUS POWER CURVE FOR TWO-LOOP AND SINGL OPERATION.

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...:-.~ . . . _ . ~ . - - . _ , . - . . . . . . . . . . . . ., -

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BOSTON EDlSON COMPANY '

f Director of Nuclear Reactor Regulation

, ATTN: D. L. Zismann, Chief 3 April 12, 1976 Page 2-Question 2. At the NRC, GE, BECo. meeting on January 24, 1976, it was requested that more justification be provided to show that a small break could not be limiting one pump operation as in a non-jet pump plant. This has not been done. Please provide a discussion showing that reflood will take place l before the crossover period.

Response 2. Add the following after the fourth paragraph of Resconse to .

Question B8A of the BEco. Letter dated March 19, 1976:

"...which compares the reflooding time for a late reflooding 3WR versus PCT turnover time fer a non-jet pump BWR" Since reflooding for smaller breaks occurs much earlier than l

PCT turnover, the PCT for smaller breaks will be substantially less (at least 200-3000F less) for single-loop operation in jet pump plants than for a comparable break in s non-jet pump plant. Thus, the MAPLEGR for single-loop operation in jet pump BWR'S will be limited by the maximum size break, and not by a smaller break as is currently the case for non-jet pump plants.

For single loop operation, i= mediate (0.1 sec.) loss of nucleate boiling is assumed independent of break size. Thus, the initial temperature response is identical for breaks of different sizes.

The largar break uncovers earlier and therefore it has a higher temperature after the time of uncovery for the large break. Very late in the transient, the later spray initiation for the case of the s= aller break causes the temperature dif ference between the large and small to be reduced. Hewever, reflooding occurs at early enough times such that the larger break has the higher tem- ,

perature. Specific detailed calculations have shown this to be the case (see NE00-20099, Section 2.2.3).

Attached are plots (Table 2-1 and Figure 2-1) of peak cladding l temperature versus time that are calculated with the single loop l ECCS analysis for various sized breaks.

Question 3. The response to question B16 is incomplete. Provide a basis for increasing the core flow uncertainty to 6.0% for one loop operation.

If the method of establishing the core flow uncertainty differs from

( that used in NEDO-20340, provide an equivalent analysis for single i

loop operation.

! Response 3. The uncertainty analysis procedure used to establish the core flow L

uncertainty for one pump operation is basically the same as for two pump operation, except for some extensions. The core flow un-certainty analysis is described in References 1 and 2. The analysis of one pump core flow uncertainty can be summarized as follows:

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, , 3 .s TABLE 2.1 F.

i COMPARISON OF KEY y:

i I

PARAMETERS FOR VARIOUS BREAK L

SIZES FOR SINGLE-LOOP ECCS ANALYSIS

  • l' DBA (LARGE 1.0 FI2 (LARGE 1.0 FT (SMALL .07 FT (SMALL PARAMETER BREAK MODEL) BREAK MODEL BREAK MODEL) BREAK MODEL)

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F Boiling Transition 0.10 0.10 0.10 0.10 f Time (sec)

Uncovery Time 25 85 85 290 (sec)

Core Spray Cool- 34 95 95 -

ing Time (sec) l Reflooding Time 107 154 160 457 (sec)

Peak Cladding 2200 1925 1900 1830 Temperature (OF) l kw

  • All calculations performed for the same plant with MAPLHGR = 12.7 gg l

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EQsTON EDISQN COMPANY Director of Nuclear Reactor Regulation l ATTN
D. L. Ziemann, Chief l April 12, 1976 Page 3 P

l a. During one pump operation, the core flow is measured by the

following formula
[ Tota 1 Active Loop f Inactive Loop h

j lCore\ ,

= ~

I i

Indicated Flow)i \ ndicated Flow)

(Flow)

( AP Th I Kforwardh h [ AP Thl 1

I l .

Kforward (Kraverse/ kKforward/

l The constant "C" is required to modify the inactive loop flow indication since the jet pump diffuser flow coefficient is dif-farent for reverse flow compared with the forward flow coefficient used for the core flow instrumentation calibration.

b. The core flow uncertainty analysis must now account for the I

uncertainty in "C". The value of "C" has been decernined I analytically, using a conservative bounding analysis; there-fore, the core flow input to the process computer during one pu=p operation has a conservative bias, since "C" was analyzed in a conservative nanner. However, the following uncertainty l analysis is based on the uncertainty in the true (or nominal)'

l value of "C", not the uncertainty in the conservative value I of "C" used in the reactor flow measurement.

"C" can be defined as:

K h ,

forward -

! C *{p averse /

where:

Eforward = The forward flow loss coefficient resulting from in-reactor calibration tests assumed for the analytical derivation of "C".

I e l ,

Note: K s (flow)* (AP)

K reverse

= The loss coefficient calculated for reverse ft,,,

1 Combining the uncertainties in Kforward "" reverse'

, be shown that

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j J' .BasTON FAHSON COMPANY Director of Nuclear Reactor Regulation i~ ATTN: D. L. Ziemann, Chief April 12, 1976 Page 4

. c. Now the effect of this reverse flow coefficient uncertainty must be related to total core flow uncertainty. Assuming that 33%*of the flow in the' active (forward flowing) jet pumps backflows through the inactive pumps, it can be shown that:

2 2 2 .

o ao + 0.33 2 27 WA 1-0.33 o C

where:

o = the uncertainty in the total core flow.

W T

o = the uncertainty in the active loop flow.

WA To produce a conservative, bounding analysis, assume c = 4.0%.

Then: W A

2

  1. 0.33 2 W

=

(4.0%)2 4 (3.42) = (4.34%)

T l-0.33 When the effect of 4.1% core bypass flow uncertainty at 12%

(bounding case) bypass flow fraction is added to the above total core flow uncertainty, the active coolant flow uncer- .

tainty is:

o 2

active

, (4,34g)2 # 0.12 ) (4,1 )2 , (4.3g )2

_1-0.12 /

coolant

' This verifies the assumption of core flow uncertainties of 6%.

Actually, the core flow accuracy is expected to be much better, i as shown above.

In summary, core flow during one pump operation is measured in

. a conservativa way, its uncertainty has been conservatively -

evaluated, and the not effect on MCPR is insignificant.

  • Note: _This value can vary from about 20% to 30%, depending on plant
  • type and operating conditions. 33% is a conservative bounding value.

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I, 50STON EDISON COMPANY

. Director of Nuclear Reactor Regulation ATTN: D. L. Ziemann, Chief April 12, 1976

  • j Page 5 .

(

REFERENCES:

1. Letter to Walter R. Butler'(AEC/NRC),

Subject:

Response to the Third Set of AEC Questions on the General Electric Li-censing Topical Reports NEDO-10958 and NEDE-10958, " General Electric BWR Thermal Analysis (CETAB): Data, Correlation and Design Application", July 11, 1974.

2. J. F. Carev, " Process Computer Performance Evaluation Accuracy", June 1974, (NEDO-20340)

In addition, your staff requested confirmation that the information contained in a July 24, 1974 letter from Mr. Eines of General Electric to Mr. Butler of the NRC was applicable to Pilgrim. General Electric Co. has confirmed that the appropriate portions of this letter are applicable to Pilgrim, and specifically that at an MC?R of 1.01 there are 1.8 fuel pins which are expected to experience bo111ng transition.

We believe that this information should be sufficient to allow issuance of your approval of single loop operation for Pilgrim I. However, if you do require additicaal information, please advise us.

Very truly yours,

_f )j'y q t{ch-;n 'o

'. Carbhdognin't i Manager Nuclear Operations a

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