ML19260A114

Forwards Responses to First Round Questions Re TMI-1 Cycle 3 Reload Application

ML19260A114
Person / Time
Site:	Crane
Issue date:	03/31/1977
From:	Arnold R METROPOLITAN EDISON CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
GQL-0435, GQL-435, NUDOCS 7910290733
Download: ML19260A114 (13)

Text

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DESCRIPTION ENCLOSU R E Ltr. tr;.ns the following:

Consists of responses to r2C Round 1 questions Concerning the L I-l Cycle 3 Reload Application.....

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Three Mile Island Unit No 1 1480 004 RJL SAFETY FOR ACTION /INFORMATION FWTon ASSIGNED AD:

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REl ATORY DOCl(ET FILE CDPY.o s o'35 t',

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Director of Nuclear Reactor Regulation

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Attn: R. W. Reid, chief lll.

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Dear Sir:

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Three Mile Island Nuclear Staticn Unit 1 (DII-1)

Docket No. 50-289 Operating License DPR-50 Enclosed please find the responses to your Round 1 Questions con-cerning the DII-l cycle 3 Reload Application.

Sincerely,

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hold Vice President RCA:JJM:pg Encicsures 1480 00^6 p?d9C0/Sy

ROUND 1 QUESTICNS FOR THREE MILE ISLAND UNIT 1 CYCLE 3 RELOAD APPLICATION QUESTION 1.

Provide a low and high power XY power map for SCC 2.

Both measured and predicted asse=bly powers should be tabulated.

RESPONSE

Eoth measured and predicted fuel assembly to average fuel assembly power ratics (XY power map) are provided in Figures 1 and.2.

Figure 1 was taken at 40% power during BOC2 physics testing with Group 6 81% withdrawn and Group 7 9% withdrawn. Figure 2 was taken at 100% reactor power during EOC2 (@ 5 3 EFPD) physics testing with Group 6 at 81% withdrawn and Group 7 at 9% withdrawn.

QUESTION 2.

Provide tabulated values for the cutpute of any quarter-core sy==etric detector strings (i.e. sun =ed axially b2t not converted to assenbly power) corresponding to the two incore maps mentioned above.

RESPONSE

Figures 3 and k give the integral readings of 1/8 - core sy==etric detectors after corrections were applied for removal of background and the depletion of rhodium.

All valves are in nanoa=ps. Figures 3 correspcnds to the re-sultant radial power map shown in Attachment 1, as Attachment h corresponds to Attachment 2.

QUESTION 3.

Provide a numerial estimate of the uncertainty in the burnup figures given in Fig. 3-2.

RESPONSE

An estimate of the uncertainty in the burnup figures given in Figure 3-2 is 5% at the 95% confidence level.

QUESTION h.

Provide tr >..aasured and predicted BOC2 rod bank worths, by bank.

RESPONSE

Individual rod group integral worths were measured for Pcd Banks 5, c, and T (Regulating & Transient Groups) during the 30C2 physics test program.

The measured and predicted values are as follows :

1480 006

. Measured Predicted

(%Ak/k)

%Ak/k Group 7 772

.858 6

1.056 1.137 5

. 6 (5

.776

'he worth of safety rod ban:s. (Groups 1-4) was measured as one bank using the r os drop technique. The measured and predicted worth of Groups 1 h is as foi ows:

Measured Predicted

%ak/k

%ak/k Group 1-4 7.4h 6.93 CUESTION 5.

What is the maximum credible worth of an ejected APSR for cycle 3?

RESPONSE

The maximum credible worth of an ejected APSR for Cycle 3 is considerably less than the ejected rod criteria, i.e., 1.00% AK/K at HZP; 0.65% AK/K at EFP. The worth of the entire bank of APSR's, whether withdrawn in sequence or before the regulating banks, is within the ER criteria. Three-D PDQ cases for a siallar plant indicate that a single ejected APSR is worth approxLnately 10 to 20% of the entire bank worth.

QUESTION 6.

Provide 30C2 =easured values for critical boror concentration and moderator temperature coefficienL. State the power and Xenon conditions under which each measurement was taken.

RESP 3 HSE The measurederitical boron concentration at BOC2 was 1337 pp=3 with a rod configuration of Groups 1-6100% withdrawn, Group 7 63% withdrawn, and Group 815% withdrawn. Two te=perature coefficient =easurements were made at zero power. The measured valves for the moderator te=perature coeffi-cient at zero were 0 9h x 10-3 gagfoF at 1366 pp=B and -5.31 x 10-3 %Ak/oF k

k 3 llk9 pp=B.

Both of these moderator te=perature coefficient reasurements and the critical boron concentration measuremente were made at zero power, clean Xenon conditions.

1480 007

. The modarator temperature coefficient was also measured at 100% full power during ti.-

BOC-2 physics test program at equilibrium xenon conditions. The temperature coefficient at power was measured to be 10 95 x 10-3 %Ak/k/0F @

820 ppmB.

QUESTION T.

Table 5-2 gives ElJ values for the total red worth at 30C3 and EOC3. Provide analogous worths for the individual rod banks which total to the valites given in Table 5-2

RESPONSE

It is not clear why t..a individual bank vorths are necessary for shutdown analysis. Bank vorths are provided, however, in the Physics Test Manual for groups 7, 6, 5, and 1-4 collectively at the HZP condition. These are:

BOC EOC (2h6 EFPD)

P3. No.

Worth. % AK/K Worth, % AK/K 1-4 5.82 5.53 5

1.08 1.00 6

0 96 0.92 7

0.73 1.0h 6.59 6.49 QUESfION 8.

How does this reactor overceme the reactivity addition due to Xenon under-shoot f611cving shutdown? Does the scram system initiate boron injection?

RESPONSE

The reactivity addition due to Xenon burn-out following reactor shutdown is overcome by boration to hot shutdown boron concentration. The reactor trip system does not automatically initiate boron injection.

QUESTION 9

Several of the accident discussion in Section T take credit for the non-positive moderator temperature coefficient. This is appropriate, since the calculated MTC for Cyc] ? 3 is negative and thus bounded by the FSAR analyses.

However, Technical Specification 3.1.T allows operation with no restrictions on the moderator temperature coefficient at less than 95% power.

Scme acci-dents and transients (e.g. ejected rod) are not necessarily most limiting at 1005 power. Therefore, explain how the ncn-positive moderator temperature coefficient is enforced.

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RESPONSE

Section 7 of the reload report is based en the philosophy of " analyses by comparison." It is structured to show that the key accident parameters used in the FSAR analysis form the bounding cases for Cycle 3 operation.

For reactor operation less than 95% uover, the moderater temperature coefficient is restricted by analysis to a positive value of +0.5 x 10-4 Ak/k/F. Technical Specification 3.1.7 can be amended to state a specifi-cation limiting the moderator temperature coeeficient to a value $,10.5 x 10 h sk/k/F for operation at power levels <95% rated power.

Technical Specification can be amended by adding the following:

3.1.7.2 The moderator te=perature coefficient shall be 5, + 0.5 x 10 b Ak/k/F at power levels 5,95% of rated power.

QUESTION 10.

Provide an analysis of operation with a mis-loaded fuel assembly. If such operation cannot be safely acco==odated, vould the incere system detect the flux anomaly?

RESPONSE

Misloading of fuel pins in an assembly and misloading fuel assemblies in the core is prevented by loading controls and procedures. This is discussed in Volume 1 of the FSAR on page 3-16e as Amendment No.18 dated 3/22/T1.

QUESTION 11.

How =any control rod group vorths vill be measured in the startup program detailed on page 9-l? What criteria vill be used to evaluate the results?

Will control rod drop times be measured?

RESPONSE

During Cycle 3 startup physics testing, the integral worth of control rod groups 5, 6, and 7 vill be measured. This will be accomplished by exchanging reactivity between baron and control rods. When groups 5-7 are at 0% vithdrawn, controlrod groups 1 h vill be dropped into the core and the resulting negative reactivity addition will be calculated.

For all the regulating control banks individually, if any one bank vorth differs from the predicted value by more than 15%, or the sum of the worths of the regulating banks differs from the predicted value by more than 10%, the first shutdown bank should be measured. If the sum of the worths of the regulating 1480 009

- banks and the first shutdown bank differs from the predicted value by more than 10%, additional shutdown bank measurements should be performed tc verify tech cal specification shutdown margin.

Control rod drop times will be measured at hot shutdown with full reactor coolant flow.

QUESTION

12. State hov many zero power moderator temperature measurements are scheduled for the cycle 3 startup program. State the planned rod configuration for each test.

RESPONSE

Two moderator temperature coefficient measurements will be made during zero power physics testing. The first measurement vill be performed with control rod groups 1 through 6 at 100%w.d. group 7 at a,75%v.d. and group 8 at 37 5%w.d.

The boron concentration at this condition vill be appro-ximately 1255pp=B. The second =ensurement will be performed with control rod groups 1-h at 100%v.d., groups 5-7 at 0%v.d. and group 8 at 37.5%v.d.

The boron concentration at this condition vill be approximately 1005pp=B.

QUESTION

13. State your schedule for submitting to NRC a brief su==ary report of physics startup tests. This report should include both measured and predicted values.

If the difference between measured and predicted values exceeds the acceptance criterion, the report should discuss the actions taken and justify the adequacy of these actions.

RESPONSE

Section 6.91. A. of Technical Specificaitons require a su==ary report of plant startup and pcwer escalation testing shall be suc=itted following 1. )

receipt of an operating license, 2) amendment to the license involving a planned increace in power level, 3) installation of fue' that has a different design or has been manufactured by a different fuel supplier, and k) =cdificaitons that may have significantly altered the nuclear thermal, or hydraulic performance of the plant. We do not feel the start-up progran at 30C3 falls into any of the above areas, therefore no start-up su==ary is planned to be submitted.

GUESTION

14. The LOCA analysis describea in Section 7.1h refers to BAW 10103.

It is the understanding of the staff that the B&W generic ECCS model was revised to eliminate return to nucleate boiling, as described in EAW 1010hA Rev. 1.

Provide or reference a new analysis using the currently approved model, as requested in our letter of Decc=ber lk,1976.

RESPONSE

Refer to the proposed revision to BAW 1010h as referenced in the URC's 1480 010

. February 18, 1977 letter frc= Mr. S. A.Varga of the NRC staff to Mr. K.E.

Suhrke of the Babcock and Wilcox Company.

QUESTION 15 The revision to Technical Specification 1.6.1 eliminates a statement concern-ing the minimum permissable number of operable excore detectors. How many excore detectors are needed for the bases of Technical Specification 3.5.2.4.c to remain valid?

RESPONSE

The required minimum number of pcVer range channels is defined in Table 3 5.1 (Instrument Operating Conditions) Ite: 2.

QUESTION 16.

Revised Technical Specification 3.5.2.7 states that power maps shall be taken every 30 full pcuer days. Is it the intent that "30 full pcver days" vill mean "up to 30 effective full power days?" As it stands, this specification vould not require a power map to be taken until exactly 30 calendar days of continuous operation at exactly 100% power had passed.

RESPONSE

Technical Specification 3.5.2.7 can be revised as follevs:

3.5.2.7 A power map shall be taken at intervals not to exceed 30 effective full power days using the incore instrumentation detection system to verify the power distribution is within the limits shown in Figure 3.5-2J.

CUESTION 17 The proposed revision to Technical Specification 3.5.2.7 requires appropriately that an incere-measured peak liner heat rate (Kv/ft) be periodically ec= pared with the LOC A 11-

's of Fig. 3.5-2J.

However, the revision also eliminates the requirement co cc= pare measured with expected power distributions. It is tne intent of the licensee not to verify predicted power distributions? If not, how would gec=etrical changes and/or loose parts within the core be detected?

_R_ESPONSE Measured power distributiens are cc= pared to predictions during tne power escalatice test program. Acceptance criteria for the pcVer distributions

=ust be satisfied before the plant is escalated to full power. Changes to core power distributions are continually monitored via the axial imbalance and quadrant tilt systems.

1480 o'~

-T-QUESTION 18.

Ecv is FAH =cnitored in this reactor?

RESPONSE

There is no Technical Specification requirement for monitoring FAH in TMI-1.

Radial power distributions are measured and compared to predicted values during the power escalation test program at the start of each operating cycle.

Acceptance criteria cust be satisfied at that time. However, in B&W reactors,

'he methodology used in setting the LSSS and LCO is such that no continuous comparisons of measured Fag to some mnximum allowable value are necessary.

QUESTION 19.

Other B&W plants are required to measure power distribution prior to operation 6bove 75% rated thermal power and periolically thereafter. The proposed Technical Specification would allev a month of operation before the first power cap is done. Provice justification for the discrepancy.

RESPONSE

Core power distributien measurements are taken at h0%, 75% and 100% full power during the BOC physics testing pro 6 ram. Measured values are compared to predicted results 9.t each plateau.

QUESTION 20 The bases on Technical Specification pp. 3-35 and 3-35a explain that operation in the restricted region of the following figures is permitted for up to four hours. Yet this restriction is not enforced in an actual specification. Explain how the four hour time limit is enforced.

RESPCNSE Technicel Specification 3.5.2.5.6 enforces this time limit.

QUESTION 21 It is not clear what the value of the power level cutoff is in revised figures 3.5-2 A, P, & C.

What is the numerical value of the power level cutoff?

PESPCNSE The pcVer level cut-off is 905 throughout Cycle 3.

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