ML20012C598

Cycle 4 Startup Rept. W/900314 Ltr

ML20012C598
Person / Time
Site:	Diablo Canyon
Issue date:	03/14/1990
From:	Shiffer J PACIFIC GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
DCL-90-074, DCL-90-74, NUDOCS 9003220261
Download: ML20012C598 (11)

Text

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' ; - , -. , . Pacific Gas and Electric Company 77 Beale Street Jane: D. Shiffer-e' San Francisco,CA94106 senior Vice President and OOI2 ar beneration 415/973-4684 March 14, 19901 PG&E Letter No. DCL-90-074 U.S. Nuclear Regulaitory Commission ATTN: Document Control Desk Hashington,.D.C. 20555 Re: Docket No. 50-275,'OL-DPR-80 Diablo Canyon Unit 1:

Unit 1 Cycle 4 Startup Report Gentlemen:-

Pursuant-to Diablo Canyon Power Plant Technical Specifications 6.9.1.1 and 6.9.1.3.- enclosed is a summary report of Diablo Canyon Unit 1-Cycle 4 startup and power escalation testing following installation of VANTAGE 5 fuel.

Kindly = acknowledge receipt of this material on the enclosed copy of this letter and return it in the enclosed addressed envelope.

incerely. -

IsL D. Shif cc:: A. P. Hodgdon J. B. Martin M. M.' Mendonca

'P.~ P. Narbut H. Rood CPUC Diablo Distribution Enclosure

~ 3069S/0080K/JilA/1881 9003220261 900314 i' PDR ADOCK 05000275 E P PDC {f 5

i ,

PG&E Letter No. DCL-90-074 i ENCLOSURE DIABLO CANYON PONER PLANT UNIT 1 CYCLE 4 STARTUP REPORT

SUMMARY

Diablo Canyon Power Flant (DCPP) Technical Specification (TS) 6.9.1.l(3) requires that a sumary report of plant startup and power escalation testing be submitted following installation of fuel that has a different design.

TS 6.9.1.3(2) requires that this startup report be submitted within 90 days following resumption of commercial power operation. Pursuant to these technical specifications, this enclosure provides the DCPP Unit 1 Cycle 4 startup report following installation of VANTAGE 5 fuel. This is the first cycle using VANTAGE 5 fuel.

l i

The DCPP Unit 1 Cycle 4 startup program extended from pre-critical testing through the completion of power ascension testing at 100 percent Rated Thermal l

Power (RTP). Significant milestones included:

Fourth cycle criticality: December 10, 1989 Commercial operation (breaker closure): December 15, 1989 Full power operation: December 31, 1989 The startup testing was divided into three phases: subcritical testing, low power physics testing, and power ascension testing. Results indicated little effect due to the change in fuel design.

UNIT 1 CYCLE 4 CORE.

L The DCPP Unit 1 Cycle 4 core is the first transition core leading to an all VANTAGE 5 fuel core in Cycle 6. Cycle 4 contains 137 standard Hestinghouse 17 x-17 fuel assemblies that were irradiated in previous fuel cycles and 56 fresh Hestinghouse VANTAGE 5 fuel assemblies (serial numbers F&FH series, see ,

Figure 1). Compared to the standard assemblies, the VANTAGE 5 assemblies have slightly smaller diameter fuel rods and guide tubes, and contain three. extra flow mixing grids, one each midway between' four of the five upper grids.

Cycle 4 VANTAGE 5' fuel assemblies also contain a burnable absorber coating on the outside of the fuel pellets in 2496 fuel rods, an axial blanket of natural uranium pellets at the top and bottom of each fuel column, and an improved bottom nozzle design known as the debris filter bottom nozzle.

The. VANTAGE 5 fresh fuel assemblies and irradiated standard fuel assemblies are arranged in an octant - symmetric pattern (Figure 1). Westinghouse fuel designers used a random quadrant fuel shuffling routine to design the core and smooth out power tilts from previous cycles of operation. The core was designed with low power assemblies in the corners to cause a reduced fast neutron flux (Iow leakage core) at the reactor vessel wall.

After refueling for Cycle 4, underwater cameras were used to videotape assembly serial numbers. Two independent observers verified that the Cycle 4 core was correctly configured.

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SUBCRITICAL TESTING Several plant tests were conducted prior to initial cycle criticality. One of these, rod-drop. testing, measured the drop times of control rods from the initiation of a reactor trip to the time of rods entering the guide tube dash

- pots. Because of the smaller. VANTAGE 5 guide tube diameters, the Technical-Specification limit was changed from a maximum of 2.2 seconds (for standard ;

fuel) to 2.7 seconds (for VANTAGE 5 or mixed cores). The average drop time that was observed was on the order of 1.4 to 1.5 seconds and showed little difference between the two types of fuel assembly (no more than about_10 percent).

LON POWER PHYSICS TESTING PROGRAM Objective The objective of low power physics testing was to verify that the nuclear characteristics of the core were consistent with Westinghouse design calculations.

Methodology Testing was performed in accordance with the following general sequence:

1. Approach to and achievement of initial criticality: First, all shutdown banks were withdrawn and then control banks were withdrawn to a Control Bank D position of 170 steps. Then an RCS dilution was executed until the core was within about 100 PCM of criticality. Finally, control rods were withdrawn until criticality was achieved.

2. Determination of physics testing flux level.

3,- Reactivity computer on-line check.

4. All-rods-out (ARO) critical boron concentration measurement.

5. AR0-Isothermal Temperature Coefficient (ITC) measurement: ITC was based on the reactivity change resulting from an RCS temperature change.

Moderator Temperature Coefficient (MTC) was inferred from ITC.

6. Rod worth measurements using the " Rod Swap" method: Rod banks were interchanged and bank worths were inferred from the position of a

" Reference Bank." The reference bank was first moved to the bottom of the core by dilution of the soluble boric acid in the reactor coolant system.

7. Calculation of boron worth.

Results Because of a high gamma-to-neutron signal ratio due to the low leakage core design, a compensating current was applied to the neutron detectors to cancel the gamma component interference. This yielded test results that agreed well with predictions.

30695/0080K Low. power physics results are summarized in Table 1. All acceptance and review criteria.wera met. Results indicated little or no effect due to changing fuel design.

' POWER ASCENSION TEST PROGRAM Obiective The main objectives of power ascension testing were to monitor core power distribution at various power levels and to recalibrate instrumentation for important plant parameters. This ensured that normal, full power ope:3 tion could be reached and sustained while meeting all design requirements.

Methodoloav Testing was performed'at specified power plateaus. For Unit 1 Cycle 4, the primary test plateaus were 30, 50, 75, 90 and 100 percent RTP. Power level changes were governed by operating procedures and fuel pre-conditioning guidelines specified by Westinghouse, the fuel vendor.

To determine the steady state core power distribution, flux mapping was accomplished by use of the Movable Incore Detector System. To determine any limitations on further power ascension, the measured peaking factors were compared to Technical Specification limits. Also, thermal-hydraulic parameters, nuclear parameters, and related instrumentation were monitored throughout the power ascension. The major areas addressed were:

• Steam and feedwater flows: Flows were measured and correlated to-verify calibration of the steam flow and feedwater flow transmitters.

Subsequent corrections or calibrations were performed as necessary.

• RCS temperature: RCS temperatures and steam generator pressures were monitored to verify alignment of the RCS delta-T instrumentation Tavg instrumentation, and the Tref program. Subsequent instrument adjustments were made as necessary to assure that appropriate delta-T power level signals were sent to the overpower delta-T (OPDT) and overtemperature delta-T (OTDT) trip circuitry. The need to reset or maintain the Tref program was based on extrapolated, full power steam generator pressures.

• RCS-flow: A primary-side heat balance was performed concurrent with a secondary side heat balance and RCS flow was calculated from the data.

RCS flow transmitters were scaled as necessary.

• Reactor power level: Secondary plant heat balance calculations were performed on a daily basis (more often if needed). If necessary, the nuclear instrumentation system (NIS) power indications were adjusted to

-agree with the heat balance.

• Incore/excore calibration: Through rod motion and a subsequent xenon oscillation, the core was operated with a variety of axial power shapes.

Axial power distributions (in terms of axial offset and axial flux difference) calculated by the movable incore detectors then were related to currents from the upper and lower power range excore detectors. From this relationship, scaling factors were calculated and used to recalibrate the power range excore nuclear detectors.

3069S/0080K

. i e

Results During Unit 1 Cycle 4 power ascension, core performance was acceptable. Flux  !

maps were performed at 35, 50 and 100 percent power. The results are  !

summarized in Table 2.  !

i The initial flux map at 35 percent power showed an incore tilt of 1.028.  !

Westinghouse was consulted and attributed the tilt to small differences in i burnup and enrichment unrelated to the change in fuel design. They provided a revised position paper en core tilts. West'nghouse accepts core tilts above  :

i 1.02 and up.to 1.04 as being within analyzed conditions. They recommend a  :

normal ascension to full power as long as they evaluate any tilt over 1.02. l The enthalpy hot channel factor (FDHN), measured at 35 percent power, was high -

enough to establish 96 percent as the maximum allowable power level for i further ascension. All other flux map parameters were within limits.

• Because of uncertainty about the core tilt and xenon and core stability.during  ;

the first flux map, a second flux map was performed the next day at 35 percent i power. Results confirmed the earlier observations, j Power ascension continued toward 75 percent power. However, due to quadrant '

power tilt ratio (QPTR) alarms and calculated excore tilts of over 1.02, operators were forced to reduce power to below 60 percent power. Prior to this, just enough time was spent at 73 percent power to allow one set of data  ;

to be gathered on feedwater and steam flows, RCS loop temperatures and flows, and steam generator pressures. From these data, it was noted that feedwater and steam flows were in agreement, RCS loop flows met all limits, loop  ;

delta-T's were scaled conservatively high by a few percent, and steam ,

, generator pressures were~close to optimum. None of the associated instruments i required immediate recalibration. There was no unexpected effect on the plant  ;

due to the change in fuel design.

A full core flux map was performed at 50 percent power as part of the  ;

incore excore detector calibration test. Results are shown in Table 2. The i data from the test were used to recalibrate excore nuclear instruments.

following recalibration, the OPTR alarm cleared and the unit core conditions  ;

! were considered acceptable for power escalation. The full core flux map t ir.ticated that there was still an incore tilt of about 1.03 (Table 2).

However, all peaking factors were within limits, and power ascension could continue as high as full power with no fuel-related limits.

The power level was escalated to 90 percent and held there for a few hours. A heat balance was performed and feedwater and steam flows, RCS loop temperatures, and steam generator pressure were checked. Results of the RCS loop temperature testing revealed that power level could be escalated to 97 percent without challenging delta-T protection setpoints. At g7 percent power, the RCS loop temperatures were determined and used to scale full power loop delta-T's. Once the delta-T instruments were rescaled, the unit was brought to full power.

I At full power, a final set of data was gathered on feedwater and steam flows,  ;

RCS loop flows, RCS loop temperatures, and RCS Tref - first stage turbine pressure and steam generator pressure scalings. Results indicated that steam p

30695/0080K -.- - _ - __ -

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. .. I flow transmitters had to be rescaled to a lower indicated flow by 2 to 5 l percent. Also, as expected, the calculated RCS flow was slightly higher than I in Cycle 3.

( In addition, all RCS loop temperatures were in specification after the j recalibration at 97 percent power. Finally, steam generator pressures were observed to be within a few psi of the optimum of 830 psia, and therefore rescaling of RCS Tref - first stage turbine pressure was not needed. Selected RCS thermal hydraulic data are shown in Table 3. Results were similar to previous cycles of operation.

A full power flux map revealed that incore tilt was reduced to 1.023 (Table 2). The power distribution (Figure 2) revealed that the tilt was generalized, asymmetric, and not related to the fuel design change. A follow-up map a few days later onfirmed the 100 percent power distribution. 1 One peaking factor was in excess of limits, however. Total peaking factor l (Fq) was in excess of its limit in the lower part of the core (Figure 2). The i Technical Specifications permitted continued. full power operation with a l concurrent reduction in AFD limits to compensate for the elevated Fq. Fq for  !

the first full power flux map exceeded its limits by a maximum of 2.14  :

percent. AFD limit curves were each redrawn closer together by about this l amount. Two days later, the second flux map was performed to evaluate the accuracy of the previous map. Again Fq exceeded its limit, this time by a maximum of 2.73 percent. Again AFD limit curves were redrawn closer together. Since the maximum Fq(z)/K(z) ratio had increased in the second flux .

map. Technical Specifications required the surveillance frequency for flux  !

mapping to be increased from every 31 effective full power days (EFPD) to t every 7 EFPD.

Westinghouse was consulted concerning the above-limit Fq. They explained that the operation of DCPP Unit 1 to the high end of the allowable Cycle 3 burnup caused the Cycle 4 core design to be right at Fq limits with no design margin at the beginning of cycle. The problems with Fq were unrelated to the change in fuel design, but rather were a function of the overall cycle core design.

Fq was predicted to be fairly constant but decreasing slowly over the first few months of full power operation. Flux mapping is being used to monitor the  !

changes in Fq to ensure that full power operation can continue within Fq limits.

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TA8tE 1 L(Bt P(MER PMYSICS TEST RESULTS: INIIT 1 CYCLE 4 TEST PREDICTED REVIEM ACCEPTANCE '

STP ITEM UNITS VALUE VALUE DIFFERENCE CRITERIA CRITER_IA i

R-30 ES Boron at Criticality PPM 1792 1301 -9(-71 PCM) N/A 1 000 POI 1

(Bank D at 185 steps)

R-6 Ak0 Boron Endpoint PPM 1804.5 1817 -12.5 N/A 50 R-6 Bank D-in Boron Endpoint PPM 1684 1691 - 7 150 N/A R-7A ITC (ARO) PCM/*F +.09 +.73 - .64 2 N/A l MTC (ARO) PCM/*F +1.61 +2.25 - .64 N/A <+5  !

I R-6 Boron Morth (ARO) PCM/ PPM -8.19 -7.98 -31 N/A 1151 R-31 Control Bank D Morth PCM 986 %3 +21 101 15%  !

i R-31 "SMAP" Morths >

Shutdown Bank A PCM B67 886 -21 i Shutdown Bank 8 PCM 795 765 +4% l i 5% 301

^

Shutdown Bank C PCM 345 340 +1.51 1 51 1 301 ,

Shutdown Bank D PCM 435 406 +71 15% 1301 I Control Bank A PCM 7% 801 - 11 1151 1301 )

Control Bank 8 PCM 458 434 + 5.5% 1 51 1 1301 Control Bank C PCM 736 705 +41 r151 2301 Total PCM 5418 5300 +21 < +101 > -101 f

i 30695/0080K

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TABLE 2 .

POMER ASCENSION FLUX MAP RESULTS: UNIT 1 CVCLE 4 Date of Nap 12-17-89 12-18-89 12-23-89 01-03-90 01-05-90 Power Level (1) 35 35 51 100 100 i Bank D Position (Steps) 196 190 204 228 228 RCS Boron (PPM) 1512 1486 1411 1259 1252 Incore Tilt 1.028' 1.029' 1.0305* 1.023* 1.024*

FDHN (V5/Std.) 1.609'*/1.364 1.606'*/1.372 1.577/1.382 1.542/1.374 1.540/1.368 Fq 2.389 2.413 2.242 2.080+ 2.093++

Fz 1.34? 1.337 1.230 1.230 1.230

Mestinghouse informed when greater than 1.02; no impact on safety analyses.

f Power level limit - 961

+ Fq over surveillance limit; AFD limits made more restrictive by 2.21.

++ Fq over surve111ance limit and increasing; AFD limits made more restrictive by 2.81 and flux map surveillance frequency increased from 31 EFPD to 7 EFPD.

30695/0080K

TABLE 3 . ,

FULL P0MER THOWIRL - NYDRAULIC OATA: UNIT 1 CYCLE 4 ,

Reactor Coolant System average temperature values *:

Loop 1: 571.7'F Loop 2: 571.4*F Loop 3: 572.5'F ,

i Loop 4: 571.5*F Average: 571.8'F

Reactor Coolant System core delta-T values

1 Loop 1: 62.1*F Loop 2: 60.0*F I Loop 3: 63.7'F Loop 4: 61.3*F Average: 61.8'F ,

l Reactor Coolant System flows **: ,

Loop 1: 93,750 gym  ;

. Loop 2: 91.370 gym  ;

loop 3: 93,740 gym ,

Loop 4: 94,010 gpa r

TOTAL: 372.870 gym (Tech. Spec. minimum 359.200 gpm)

Cycle 3 reference temperature 572.8'F. This temperature maintains steam generator

pressures at approximately 830 psia at full load. Final Tavg adjustments may be based on future turbine performance optimization.

4

• • All thimbie pings removed from Unit i Cycle 4 core to compensate for slightly higher pressure drop and icwer flow in VANTAGE-5 fuel assembites.

30695/0080K 4

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BCPP UNIT 1 1/ 3/CS 3337.8 MM C79. MND/MT 228.9 IMAP919329RUW494 C1/03/99 12:91818 BCIS7 ,

INCORE 3.9.3.Il PAGE- 249 COPYRIOMT NESTINGMOUSE ELECTRIC CORP. - .

MEASURED AND ped 2ENT. DIFF. OF FDNN R P N M L K J M G F E D C. 3 A ,  ;

. 8.315. 9.455. 0.550. .

. . 0.%43. 8.546. 8.451. 8.396.

1 . 3.7. 1.8. 0.7. -0.6. -0.1. 1.9. 1.0.

. S.391. 9.559. 1.292. 1.279. 1.277. 1.973. 1.259. 1.242. 1.146. 9.518. 8.289.

2 . 4.7. 4.6. 3.7. 2.9. 0.8. -0.2. -0.5. -0.9. -1.1. -1.4. -2.7. ,

. 9.296. 1.945. 1.319. 1.198. 1.189. 1.949. ..1 . 36 8 . 1.023. 1.113. 1.123. 1.255. 9.989. 9.276.

3 . 2.7. 2.7. 3.7. 3.7. 3.7. 9.6. -0.2. -1.0. -2.9. -2.1. -1.4. -2.9. -4.1.

. 9.535. 1.296. 1.198. 1.165. 1.141. 1.889. 1.146. 1. 06 5. 1. e92. 1.121. 1.141. 1.231. s.593. -

4 . 1.7. 1.8. 2.1. 1.4 2 1.4 8.S. -0.9. -1.5. -2.9. -2.4 -2.8. -3.3. -4.2.

. 9.311. 1.185. 1.168. 1.179. 1.976. 1.335. 9.986. 1.391. 9.959. 1.271. 1.926. 1.197. 1.139. 1.146. 9.319.

5. 2.5. 2.3. 1.8. 1.9. 1.9. 2.1. 9.8. -1.7. -2.0. -2.8. -2.8. -3.6. -1.5. -1.1. Zee.

. 9.464. 1.298. 1.164. 1.143. 1.335. 1.215. 1.223. 1.291. 1.199. 1.182. 1.293. 1.996. 1.1 34 . 1.244 . S.4 54 .

6 . 3.8. 2.9. 1.7. 1.6. 2.9. 1.9. 8.7. -2.9. -1.3. -1.7. -1.2. -2.6. -1.8. -0.6. 2.9.

. . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . i

! . 0.566. 1.388. 1.957. 1.197. 1.992. 1.238. 1.993. 1.382. 1.982. 1.219. s.975. 1.073. 1.035. 1.275. 0.553. l 7 . 3.7. 3.3. 2.3. 2.4. 2.4. 1.9. 1.4. 1.2. 9.4. 9.4. -0.3. -0.6. 8.1. 8.7. 1.3.  :

. 9.46 9. 1.117. 1.419. 1.188. 1.359. 1.249. 1.385. 1.986. 1.365. 1.219. 1.315. 1.145. 1.368. 1.967. S.443.

8 . 4.9. 3.9. 3.5. 2.7. 2.6. 1.9. 1.5. 9.4. 9.9. -9.5. -0.7. -1.9. -9.2. -0.8. -1.7. l

. S.569. 1.318. 1.975. 1.116. 1.919. 1.247. 1.181. 1.381. 1.972. 1.293. 9.961. 1.962. 1.923. 1.254 S.535.

9 . 4.1. 4.1. 4.0. 3.3. 3.2. 2.7. 2.1. 1.2. -8.5. -1.9. -1.8. -1.7. -1.9. -1.8. -2.9.

. S.459. . . 1.289. 1.198. 1.129. 1.316. 1.299. 1.232. 1.234..1.228.. 1.182. 1.277. 1.997. 1.199. 1.225. W.437.

19. 2.6. 2.8. 4.5. 9.4 9.6. 9.5. 1.4. 9.7. 8.5. -1.7. -2.3. -2.5. -3.3. -2.3. -2.3.

1 . S.396. 1.164. 1.156. 1.159. 1.953. 1.311. 0.986. 1.329. 9.975. 1.278. 1.919. 1.197. 1.997. 1.152. 9.392.

11. 9.8. e.7. e.8. e.2. -e.2. e.2. e.7. s.4 -e.4. -2.3. -3.5. -3.6. -4.4 -0.6. -9.5.

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. s.537. 1.329. 1.167. 1.125. 1.193. 1.988. 1.157. 1.979. 1.998. 1.195. 1.128. 1.285. 0.519.

12 . 2.1. 3.7. -0.6. -2.1. -2.0. 8.7. 3.1. -0.1. -2.4. -3.7. -3.9. -5.3. -2.9.

. 9.296. 1.947. 1.286. 1.127, 1.134. 1.034. 1.368. 1.932. . . 1.141.

. 1.157. 1.193. 9.957. 9.273.

13 . 2.9. 2.9. 1.1. -1.8. -1.1. c.1. -0.3. -0.1. -0.5. -3.5. -4.3. -5.9. -5.1. '

l . 9.296. 9.539. 1.142. 1.244 1.267. 1.974. 1.273. 1.277. 1.128. S.489. 9.271.

14 . 2.8. 9.7. -1.4. -0.8. 9.1. -0.1. 0.6. 1.9. -2.6. -7.9. -6.9.

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. 9.392. 9.447. S.543. 9.447. 9.547. 9.454 s.296. . MEA 5 .

15 . -8.3. -0.1. -9.6. -0.7. 9.1. 1.7. -2.6. . DIFF .

t FIGURE 2 RADIAL POWER DISTRIBUTION (ARD, 1001 RPT) 4 1

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