ML18151A551

to NE-1187, Surry Unit 1,Cycle 16 Startup Physics Tests Rept.

ML18151A551
Person / Time
Site:	Surry
Issue date:	02/10/1999
From:	Lawrence D, Lingler N, Twitchell R VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18151A553	List: ML18151A551 ML18152B547
References
NE-1187, NE-1187-R, NE-1187-R00, NUDOCS 9902230207
Download: ML18151A551 (72)

Text

e TECHNICAL REPORT NE-1187 - Rev. 0 SURRY UNIT 1, CYCLE 16 STARTUP PHYSICS TESTS REPORT NUCLEAR ANALYSIS AND FUEL NUCLEAR ENGINEERING AND SERVICES VIRGINIA POWER FEBRUARY 1999 PREPAREDBY: .;t)~Jb) L~~ Cf~~~.)111 N. A. W. Ling'ler Date REVIEWED BY: --l{/JJI-=*_J=:tt("""""---"'=.o=.."------

R. W. Twitchell z/"'/n Date REVIEWED BY: J~qi~~U- ~t

_it......,_,_,...t,.......

APPROVED BY:_*- - -~ - +- ' =- -_. ; __

• __~ z1,., /-;1

~e Date QA Category: Nuclear Safety Related Keywords: SPS1. S1Cl6, Startup

CLASSIFICATION/DISCLAIMER The data, techniques, information, and conclusions in this report have been prepared solely for use by Virginia Electric and Power Company (the Company), and they may not be appropriate for use in situations other than those for which they have been specifically prepared. The Company therefore makes no claim or warranty whatsoever, express or implied, as to their accuracy, usefulness, or applicability. In particular, THE COMPANY MAKES NOWARRANTY OF MERCHANT ABILITY OR FITNESS FOR A PARTICULAR PURPOSE, NOR SHALL ANY WARRANTY BE DEEMED TO ARISE FROM COURSE OF DEALING OR USAGE OF TRADE, with respect to this report or any of the d~ta, techniques, information, or conclusions in it. By making this report available, the Company does not authorize its use by others, and any such use is expressly forbidden except with the prior written approval of the Company. Any such written approval shall itself be deemed to incorporate the disclaimers of liability and disclaimers of warranties provided herein. In no event shall the Company be liable, under any legal theory whatsoever (whether contract, tort, warranty, or strict or absolute liability), for any property damage, mental or physical injury or death, loss of use of

• property, or other damage resulting from or arising out of the use, authorized or unauthorized, of this report or the data, techniques, information, or conclusions in it.

NE-1187 S1Cl6 Startup Physics Tests Report Page 1 of 56

e TABLE OF CONTENTS PAGE Classification/Disclaimer............................................................................. 1 Table of Contents ............................................................ ;............................ 2

• List of Tables .......................................... '..................................................... 3 List of Figures.............................................................................................. 4 Preface......................................................................................................... 5

__, Section 1. futroduction and Summary......................................................,. 6 Section 2 Control Rod Drop Time Measurements.................................. 15 Section 3 Control Rod Bank Worth Measurements................................ 20 Section 4 Boron Endpoint and Worth Measurements............................. 25 Section 5 Temperature Coefficient Measurement................................... 29 Section 6 Power Distribution Measurements ... ,...................................... 32 Section 7 References ............................................ :................................... 39 APPENDIX Startup Physics Test Results and Evaluation Sheets .......... 40 NE-1187 S 1C 16 Startup Physics Tests Report Page 2 of 56

LIST OFTABLES TABLE TITLE PAGE 1.1 Chronology of Tests............................................................................. 9 2.1 Hot Rod Drop Time Summary............................................................. 17 3.1 Control Rod Bank Worth Summary..................................................... 22

• .."-j 4.1 Boron Endpoints Summary.................................................................. 27 4.2 Boron Worth Coefficient. ................................................... 28

***~

5.1

• Isothermal Temperature Coefficient Summary.................................... 31 6.1 Incore Flux Map Summary.................................................................. 34 6.2 Comparison of Measured Power Distribution Parameters With Their Core Operating Limits ...................................................... .35 NE-1187 S1C16 Startup Physics Tests Report Page 3 of 56

~* .1

e LIST OF FIGURES FIGURE TITLE PAGE 1.1 Core I.,oadirig Map............................................................................... 10

..-\ 1.2 Beginning of Cycle Fuel Assembly Burnups ...................................... 11 1.3 lncore Thimble I.,ocations ........................................................... 12 1.4 Burnable Poison and Flux Suppression Insert I.,ocations ................... 13 1.5 Control Rod I.,ocations ........................................................................ 14 2.1 Typical Rod Drop Trace ...................................................................... 18 2.2 Rod Drop Time - Hot Full Flow Conditions....................................... 19 3.1 Control Bank B Integral Rod Worth - HZP ..................................... 23 3.2 Control Bank B Differential Rod Worth- HZP ................................. 24 6.1 Assemblywise Power Distribution - 28% Power. ............................... 36 6.2 Assemblywise Power Distribution - 69% Power. ............................... 37 6.3 Assemblywise Power Distribution -99% Power ................................ 38 NE-1187 S 1C 16 Startup Physics Tests Report Page 4 of 56

PREFACE This report presents the analysis and evaluation of the physics tests, which were performed to verify that the Surry Unit 1, Cycle 16 core could be operated safely, and

- m_akes an initial evaluation of the_ performance of the core. It is not the intent of this report to discuss the particular methods of testing or to present the detailed data taken. t. j Standard testing techniques and methods of data analysis* were used. The test data, results and evaluations, together with the detailed startup procedures, are on file at the ,Surry Power Station. Therefore, only a cursory discussion of these items is included in this report. The analyses presented include a brief summary of each test, a comparison of the.

test results with design predictions, and an evaluation of the results.

The Surry Unit "I, Cycle 16 startup physics test results and evaluation sheets are included as an appendix to provide additional information on the startup test results.

Each data sheet provides the following information: 1) test identification, 2) test ,;;_* ...

conditions (design), 3) test conditions (actual), 4) test results, 5) acceptance criteria, and

6) comments concerning the test. These sheets provide a compact summary of the startup test results in- a consistent format. The design test conditions and design values (at design conditions) of the measured parameters were completed prior to the startup physics testing. The entries for the design values were based on the calculations performed by Virginia Electric and Power Company's Nuclear Analysis and Fuel Group 1* During the tests, the data sheets were used as guidelines both to verify that the proper test conditions were met and to facilitate the preliminary comparison between measured and predicted test results, thus enabling a quick identification of possible problems occurring during the tests.

NE-1187 S1C16 Startup Physics Tests Report Page 5 of 56

SECTION 1 INTRODUCTION AND

SUMMARY

On October 19, 1998 Unit No. 1 of the Surry Power Station shutdown for its sixteenth refueling. During this shutdown, 57 of the 157 fuel assemblies in the core were replaced with 56 fresh assemblies and one once-burned assembly. The Cycle 16 core consists of 6 sub-batches of fuel: two once burned batches from Cycle 15 (batches 17A and 17B); two twice-burned batches from Cycle 14 (batches 16A and 16B); and two fresh batches (batches 18A and 18B). The single once-burned assembly is from Cycle 14 (batch 16A).

The core loading pattern and the design parameters for each_ sub-batch are shown in Figure 1.1. Beginning of cycle fuel assembly bumups are given in Figure 1.2. The

• incore *thimble locations available during startup physics testing are identified in* Figure 1.3. Figure 1.4 identifies the location and number of burnable poison rods and flux suppression insert locations for Cycle 16, while figure 1.5 identifies the control rod locations.

The cycle 16 core achieved initial criticality at 0527 on November 19, 1998. Prior to and following criticality, startup physics tests were performed as outlined in Table 1.1.

A sununary of the physics test results follows:

NE-1187 S1C16 Startup Physics Tests Report Page 6 of 56 _

1.

The measured drop time of each control rod was within the 2.4 second limit of L

Technical Specification 3.12.C.l.

2. The reference control rod bank was measured with the dilution method, and the result was. within 0.2% of the design prediction. Individual control rod bank ,..

worths were measured using the rod swap technique 2*3 and the resul~ were within

-2.1 % of the design predictions. The sum of the individual measured control rod bank worths was within -0.8% of the design prediction. All results were within the design tolerance of +/-15% for individual bank worths (+/-IO% for the rod swap reference bank worth) and the design tolerance of +/- 10% for the sum of the individual control rod bank worths.

3. Measured critical boron concentrations for two control bank configurations were within 13 ppm of the design predictions. The all-rods-out (ARO) result was within the 50 ppm design tolerance, and met the Technical Specification 4.10.A criterion that the overall core reactivity balance shall be within +/- 1% L1k/k of the design prediction. The reference bank in critical boron concentration was within its design tolerance.

4. The boron worth coefficient measurement was -within 3.4% of the design prediction, which is within the design tolerance of+/- 10%.

NE-1187 S1Cl6 Startup Physics Tests Report Page 7 of 56

5. The measured isothermal temperature coefficient (ITC) for the all-rods-out configuration was within 0.92 pcmfF of the design prediction. This result is within the design tolerance of +/-3 pcmfF. The measured ITC was -0.83 pcml°F.

When the Doppler temperature coefficient (-1.70 pcml°F) and a 0.5 pcml°F uncertainty are accounted for in the +6.0 pcml°F MTC limit of Core Operating

. Limits Report (COLR) 2.1, the MTC acceptance criteria is satisfied as long as the ITC is less positive than +3.80 pcml°F.

6. Measured core power distributions were within established acceptance criteria and

~ i',:

COLR limits. The average relative assembly power distribution measured/predicted percent difference was l.4% or less for the three initial power ascension flux maps. The heat flux hot channel factors, F-Q(Z), and enthalpy rise hot channel factors, F-DH(N), were within the limits of COLR Sections 2.3 and 2.4, respectively.

In summary, all startup physics test results were acceptable. Detailed results, specific design tolerances and acceptance criteria for each measurement are presented in the following s~ctions of this report.

NE-1187 S 1C 16 Startup Physics Tests Report Page 8 of 56

e Table 1.1

• ..r.,.

SURRY UNIT I -CYCLE 16 STARTUP PHYSICS TESTS CHRONOLOGY OF TESTS Reference Test Date Time Power Procedure Hot Rod Drop-Hot Full Flow 11/18/98 1150 HSD l-NPT-RX-014 Zero Power Testing Range 11/19/98 0640 HZP l-NPT-RX-008 Reactivity Computer Checkout 11/19/98 0729 HZP 1-NPT-RX-008 Boron Endpoint - ARO 11/19/98 0800 HZP 1-NPT-RX-008 Temperature Coefficient - ARO 11/19/98 1027 HZP l-NPT-RX-008 BankB Worth 11/19/98 1114 HZP 1-NPT-RX-008 Boron Endpoint - B in 11/19/98 1515 HZP 1-NPT-RX-008 Bank D Worth- .Rod Swap 11/19/98 1550 HZP 1-NPT-RX-008 Bank C Worth - Rod Swap 11/19/98 1613 HZP 1-NPT-RX-008 Bank A Worth- Rod Swap 11/19/98 1630 HZP 1-NPT-RX-008 Bank SB Worth - Rod Swap 11/19/98 1700 .HZP . 1-NPT-RX-008 Bank SA Worth - Rod Swap 11/19/98 1735 HZP 1-NPT-RX-008 Flux Map - 28% Power 11/21/98 1512 28% l-NPT-RX-002 Peaking Factor Verification 1-NPT-RX-005

& Power Range Calibration l-NPT-RX-008 Flux Map - 69% Power 11/26/98 0300 69% 1-NPT-RX*002 Peaking Factor Verification 1-NPT-RX-005

& Power Range Calibration 1-NPT-RX-008 Flux Map - 100% Power 12/02/98 1000 100% l-NPT-RX-002 Peaking Factor Verification l-NPT-RX.:005

& Power Range Calibration 1-NPT-RX-008 NE-1187 S1Cl6 Startup Physics Tests Report Page 9 of 56

(

Figure 1.1 SURRY UNIT 1 - CYCLE 16 CORE LOADING MAP

-.-- R p N M L K J H G F E D C B A I6B 16A I6B 57A 21A 37A 16A l6B I8B 17B 18B 16B 16A 03A 42A 36C SIB 54C 52A ISA 2 16B 18A 18B 17A 18A 17A 18B 18A 16B 55A 24C 45C OIB 13C 13B 56C 32C SIA 3 16B 16B 18B 17A 18A 17A 18A 17A 18B 16B 16B 63A 38A 55C 09B 27C 07B 08C 16B 40C 45A 47A 4 16A 18A 18B 17B 18A 17A 18A 17A 18A l7B 18B 18A 16A 33A 21C 34C 35B llC l7B 07C 31B 16C 57B 48C 30C 22A 5 16B 18B 17A 18A l7B 17B 17A l7B 17B 18A I7A 18B 16B 53A 43C 18B ISC 54B 42B 03B 34B 40B 23C llB 38C 39A 6 l6B 18B 17A 18A 17A 17B l7B 17B 17B 17B 17A 18A 17A 18B l6B' 41A 49C 28B 26C 27B 53B 41B 48B 46B 52B 30B 04C 05B 44C 58A 7 16A 17B 18A 17A 18A 17A l7B 16A 17B 17A 18A 17A 18A ': 17B 16A 12A 49B 29C 04B 20C 12B 36B 16A 47B 19B 3IC 32B OIC 43B I IA 8 I6B 18B 17A 18A 17A l7B l7B 17B 17B I7B 17A 18A 17A 18B 16B 62A SIC 06B osc 26B 39B 56B 58B 44B 45B 25B 12C 20B soc 44A 9 16B 18B 17A 18A I7B 17B 17A 17B l7B 18A 17A l8B I6B 40A 53C 22B 19C 38B 33B 08B 55B 37B 14C 21B 46C 48A 10 16A 18A 18B 17B 18A 17A 18A 17A 18A 17B l8B 18A 16A 09A 06C 47C 60B 22C 24B 02C 29B 28C 59B 35C 09C 17A 11 16B l6B 18B 17A 18A 17A 18A 17A 18B 16B I6B 46A 64A 39C 23B 25C I5B 03C IOB 42C 54A 56A 12

\;'

16B 18A 18B 17A 18A 17A l8B 18A 16B 50A 17C 37C l4B 18C 02B 33C lOC 6IA 13 16A l6B l8B l7B l8B 16B 16A D ===>Batch

==> Assembly ID 31A 60A 52C l6B 59A SOB 16A 27A 41C l6B 49A 43A 26A 14 15 FUEL ASSEMBLY DESIGN PARAMETERS SUB-BATCH 16A 168 17A 178 18A 188 INITIAL ENRICHMENT (W/0 U-235) 3.81 4.01 3.81 4.01 4.11 4.26 BURNUP AT BOC 16 37996 38168 21779 21040 0 0 (MWD/MTU)

ASSEMBLY TYPE ISxlS ISxlS 1Sxl5 15xl5 15xl5 15xl5 NUMBER OF ASSEMBLIES 13 28 32 28 32 24 FUEL RODS PER ASSEMBLY 204 204 204 204 204 204 NE-1187 S1Cl6 Startup Physics Tests Report Page 10 of 56

• * \.

Figure 1.2 SURRY UNIT 1 - CYCLE 16 BEGINNING OF CYCLE FUEL ASSEMBLY BURNUPS R p N M L K J H G F E D C B A 57A 21A 37A 38.61 40.18 38.08 l 03A 42A 36C SIB 54C 52A ISA 1¥.,=1 38.84 32.81 0.00 21.77 0.00 33.33 38.46 2 55A 24C 45C OIB l3C 13B 56C 32C SIA 42.96 0.00 0.00 17.59 0.00 17.99 0.00 0.00 42.37 3

_,1*.1 63A 38A 55C 09B 27C 07B 08C 16B 40C 45A 47A 42.27 37.17 0.00 23.37 0.00 22.83 0.00 23.40 0.00 37.33 42.38 4 33A 21C 34C 35B llC l7B 07C 31B I6C 57B 48C 30C 22A 38.39 0.00 0.00 22.13 0.00 23.79 0.00 23.64 0.00 21.57 0.00 0.00 38.25 5  ;*,.

53A 43C 18B . l5C 54B 42B 03B 34B 40B 23C llB 38C 39A ...

33.90 0.00 23.38 0.00 21.83 18.46 22.59 18.27 21.61 0.00 22.80 0.00 33.04 6

• ~* J 41A 49C 28B 26C 27B 53B 4IB 48B 46B 52B 30B 04C 05B 44C 58A 38.53 0.00 18.20 -o.oo 23.54 18.64 21.96 22.57 21.65 18.45 23.41 0.00 17.50 0.00 38.65 7 12A 49B 29C 04B 20C l2B 36B 16A 47B 19B 3IC 32B OIC 43B llA f-40.43 21.75 0.00 22.56 0.00 22.54 23.32 23.56 22.61 21.99 0.00 22.67 0.00 22.01 40.53 8 62A SIC 06B 05C 26B 39B 56B 58B 44B 45B 25B 12C 20B soc 44A . -**

38.33 0.00 17.60 0.00 23.55 18.43 21.36 22.52 21.97 18.57 23.68 0.00 18.01 0.00 38.52 9 ~?."' ~

40A 53C 22B l9C 38B 33B 08B 55B 37B 14C 2IB 46C 48A 33.20 0.00 23.26 0.00 - 21.88 18.46 22.37 18.37 21.21 0.00 23.20 0.00 33:77 IO 09A 06C 47C 60B 22C 24B 02C 29B 28C 59B 35C 09C 17A V,..*)

38.41 0.00 0.00 21.75 0.00 23.31 0.00 23.50 0.00 22.04 O;OO 0.00 38.26 11 46A 64A 39C 23B 25C l5B 03C IOB 42C 54A 56A 42.90 37.89 0.00 22.98 0.00 22.80 0.00 23.25 0.00 37.37 42.26 12 50A l7C 37C l4B l8C 02B 33C IOC 61A 42.22 0.00 0.00 17.64 0.00 17.61 0.00 0.00 42.84 13 31A 60A 52C SOB 41C 43A 26A 38.27 33.69 0.00 21.62 0.00 33.45 38.70 14 59A 27A 49A 38.65 40.36 _38.44 15 r--J==> Assembly ID LJ==> Assembly Bumup (MWD/MTU)

NE-1187 S 1C 16 Startup Physics Tests Report Page 11 of 56

Figure 1.3 SURRY UNIT I-CYCLE 16 INCORE THIMBLE LOCATIONS R p N M L K J H G F E D C B A IT*

IT 2

IT IT IT IT 3

IT IT* IT* ,,

\(;, 4 \,!

1, IT IT IT IT* IT* IT \l,.

5 ,,

,)

IT IT IT 6

IT IT IT IT* IT 7

IT IT IT IT IT IT 8

IT IT IT IT 9

IT IT IT* IT '.,~\'

• • , IO i)t IT IT IT IT 11 IT IT IT IT 12 IT IT 13 IT IT 14 IT 15 IT - Incore Thimble

• - Unavailable Location NE-1187 S 1C 16 Startup Physics Tests Report Page 12 of 56

• e Figure 1.4 SURRY UNIT 1 - CYCLE 16 BURNABLE POISON AND FLUX SUPPRESSION INSERT LOCATIONS ~- <

R p N M L K J H G F E D C B A FSIS FS005 3P 3P BP643 BP650 2 ...1:-i FSIL SP 20P 20P 20P SP FSIL FS014 BP676 IBP659 BP631 BP666 BP675 FS007 3 FSIL 20P 20P 20P 20P FSIL *,c;':.\'

FS008 BP660 BP638 03P635 BP665 FS009 4 ~*

SP 20P 20P 20P 20P 20P SP '-'.

BP678 BP661 BP654 BP63C BP657 BP664 BP682

. 20P 20P 20P 20P 5

j-i_ ~

• 03P662 BP655 BP656 BP663 6 3P 20P 20P 3P BP644 BP642 BP641 BP649 7 FSIS 20P 20P 20P 20P FSIS FS002 BP634 BP629 BP628 BP633 FS003 8 3P 20P 20P 3P BP645 BP639 BP637 BP648 9 20P 20P 20P 20P r..:1 BP674 BP658 BP653 BP667 10 SP 20P 20P 20P 20P 20P

• SP c"

BP677 BP673 BP651 BP627 BP652 BP668 BP681 11 '!!.:,,

FSIL 20P 20P 20P 20P FSIL FS010 BP672 BP636 BP640 BP669 FS01 l 12 FSIL SP 20P 20P 20P SP FSIL FS012 BP680 BP671 BP632 BP670 BP679 FS013 13 3P 3P BP646 BP647 14 FSIS ... -~

FS004 15 3P - 3 BURNABLE POISON ROD CLUSTER 5P - 5 BURNABLE POISON ROD CLUSTER 20P - 20 BURNABLE POISON ROD CLUSTER FSIL - FLUX SUPPRESSION INSERT (LONG)

FSIS - FLUX SUPPRESSION INSERT (SHORT) xxP - #OF BP RODS BP###, FS### BP ASSEMBLY ID, FLUX SUPPRESSION INSERT ID NE-1187 SIC16 Startup Physics Tests Report Page 13 of 56

'.).,

Figure 1.5 SURRY UNIT I-CYCLE 16 CONTROL ROD LOCATIONS R p N M L K J H G F E D C B A A D A 2 SA SA 3 C B B C 4 SB SB 5 A B D C D B A SA SB SB SA --,

D C C D .. 8 SA SB SB SA 9

./

A B D C D B A IO SB SB 11 C B B C 12 SA SA 13 Absorber Material A D A 14 Ag-In-Cd 15 Function Number of Clusters Control Bank D 8 Control Bank C 8 Control Bank B 8 Control Bank A 8 Shutdown Bank SB 8 Shutdown Bank SA 8 NE-1187 S 1C 16 Startup Physics Tests Report

• Page 14 of 56

SECTI0N2

• ~*.

CONTROL ROD DROP Tfi\,IE MEASUREMENTS The drop time of each control rod was measured at hot full-flow reactor coolant system (RCS) conditions (Tavg of 547 +/- 5° F) in order to verify that the time from initiation of the rod drop to the entry of the rod into the dashpot was less than or equal to the maximum allowed by Technical Specification 3.12.C.1.

The rod drop times were measured by withdrawing three banks to their fully withdrawn position and dropping all 24 control rods within the three banks by opening th~ reactor trip breakers. This allowed the rods to drop into the core as they would during a plant trip with the exception that during a plant trip all six banks would drop simultaneously. The Individual Rod Position Indication (IRPD secondary coil voltage

~--  :

signals were recorded for each rod. in the bank to determine each rod's drop time. A . \,, .. J stationary gripper coil voltage was also .measured as confirmation of the initiation of the reactor trip breaker opening. This procedure was repeated for the remaining three banks.

As shown on the sample rod drop trace in Figure 2.1, the initiation of the rod drop is indicated by the decay of the stationary gripper coil .voltage when the reactor trip breakers reopened. As the rod drops, a voltage is induced in the IRPI secondary coil.

The magnitude of this voltage is a function of control rod velocity. As the rod enters the

,, .l dashpot region of the guide tube, its velocity slows causing a voltage decrease in the IRPI NE-1187 S 1C 16 Startup Physics Tests Report Page 15 of 56

~ *-* ...

coil. This voltage reaches a minimum when the rod reaches the bottom of the dashpot.

Subsequent variations in the trace are caused by rod bouncing.

The measured drop times for each control rod are recorded on Figure 2.2. The slowest, fastest, and average drop times are summarized in Table 2.1. Technical Specification 3.12.C. l specifies a maximum rod drop time from loss of stationary gripper coil voltage to dashpot entry of 2.4 seconds with the RCS at hot, full flow conditions.

The test results satisfy this limit.

• • 1:**

NE-1187 S 1C 16 Startup Physics Tests Report Page 16 of 56

Table 2.1 SURRY UNIT 1 - CYCLE 16 STARTUP PHYSICS TESTS HOT ROD DROP TIME

SUMMARY

ROD DROP TIME TO DASHPOT ENTRY SLOWEST ROD FASTEST ROD AVERAGE TIME

---

4----

F-8 1.39 C-9/F-2 1.23 sec. 1.29 sec.

NE-1187 S1C16 Startup Physics Tests Report Page 17 of 56

e

~**-.,,

Figure 2.1 SURRY UNIT 1 -CYCLE 16 STARTUP PHYSICS TESTS TYPICAL ROD DROP TRACE Beginning Of Dashpot Bottom Of Dashpot Initiation Of Rod (Begimwig' Of. First . (Begimdng Of First Drop Event Mark Downturn in 'l'race) Upturn in 'l'race)

I \ I

~..................--..................'--......--+-----------tStationary Gripper C~il Voltage Trace 60Hz IRPI Primary Coil nf~~:J.,,Q,Al.',A,~ Voltage Trace ROD DROP TIME MEASUREMENT NE-1187 S*IC16 Startup Physics Tests Report Page 18 of 56

Figure 2.2 SURRY UNIT 1 - CYCLE 16 STARTUP PHYSICS TESTS ROD DROP TIME - HOT FULL FLOW CONDITTONS R p N M L K J H G F E D C B A 1.30 1.27 1.23 1.28 1.28 3 1.27 1.26 1.29 1.31 4

• ~_,..',

1.29 1.29 5 1.26 1.26 1.28 1.30 1.27 1.28 1.36 1.27 1.31 1.27 1.28 7 1.29 1.31 1.39 1.32 8 1.31 1.28 1.30 1.23 9 1.27 1.29 1.28 1.31 1.31 1.30 1.30 10 L27 1.29 11 1.27 1.26 1.28 1.32 12 1.25 1.29 13 1.32 1.27 1.35 14 15 Q => Rod drop time to dashpot entry (sec.)

NE-1187 S 1C 16 Startup Physics Tests Report Page 19 of 56

• SECTION 3 e

CONTROL ROD BANK WORTH MEASUREMENTS Control rod bank worths were measured for the control and shutdown banks using the rod swap technique 2*3

• The initial step of the rod swap method diluted the predicted most reactive control rod bank (hereafter referred to as the reference bank) into the core and measured its reactivity worth using conventional test techniques. The reactivity changes resulting from the reference bank movements were recorded continuously by the reactivity computer and were used to determine the differential and integral worth qt the reference bank. For Cycle 16, Control Bank B was used as the reference bank.

(\\

After the completion of the reference bank reactivity worth measurement, the reactor coolant system temperature and boron concentration were stabilized with the reactor near critical and the reference bank near full insertion. Initial statepoint daJa for the rod swap maneuver were obtained by moving the reference bank to its fully inserted position with all other banks fully withdrawn and recording the core reactivity and moderator temperature. From this point, a rod swap maneuver was performed by withdrawing the reference bank several steps and then one of the other control rod banks (i.e., a test bank) was inserted to balance the reactivity of the reference bank withdrawal.

This sequence was repeated until the test bank was fully inserted and the reference bank was positioned such that the core was just critical or near the initial statepoint reactivity.

This measured critical position (MCP) of the reference bank with the test bank fully NE-1187 S 1C 16 Startup Physics Tests Report Page 20 of 56

inserted was used to determine the integral reactivity worth of the test bank. The core reactivity, moderator temperature, and the differential worth of the reference bank were recorded with the reference bank at the MCP. The rod swap maneuver then was repeated in reverse such that the reference bank again was fully inserted with the test bank fully withdrawn from the core. This rod swap process was then repeated for each of the other control and shutdown banks.

A summary of the test results is given in Table 3.1. As shown in this table *and the Startup Physics Test Results and Evaluation Sheets given in the Appendix, all of the individual measured bank worths for the control and shutdown banks were within the design tolerance (+/- 10% for the reference bank, +/-15 % for test banks of worth greater than 600 pcm, and +/-lQO pcm for test banks of worth less than or equal to 600 pcm.). The sum of the individual measured rod bank worths was within -0;8% of the design prediction.

This is well within the design tolerance of+/- 10% for the sum of the individual control rod

.bank worths.

The integral and differential reactivity worths of the reference bank (Control Bank B) are shown in Figures 3.1 and 3.2, respectively. The design predictions and the measured data are plotted together in order to illustrate their agreement. In summary, the measured rod worth values were satisfactory.

NE-1187 S 1C 16 Startup Physics Tests Report Page 21 of 56

• Table 3.1 e

SURRY UNIT 1 - CYCLE 16 STARTUP PHYSICS TESTS CONTROL ROD BANK WORTH

SUMMARY

MEASURED PREDICTED PERCENT WORTH WORTH *DIFFERENCE(%)

BANK (PCM) (PCM) (M-P)/P X 100 B-Reference Bank 1374 1371 0.2 D 1070 1077 -0.6 C 712 727 -2.1 A 286 290 -1.4*

SB 1129 1129 0 SA 1010 1032 -2.1 *'.;.

Total Worth 5581 5626 -0.8 ..  ;

• Difference is less than 100 pcm.

.**,,.~

NE-1187 S1C16 Startup Physics Tests Report Page 22 of 56

Figure 3.1 SURRY UNIT 1 - CYCLE 16 STARTUP PHYSICS TESTS CONTROL BANK B INTEGRAL ROD WORTH - HZP ALL OTHER RODS WITHDRAWN 1600 1400 IL..

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1200 \'"' .

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H 1000 -.~. ~

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0 50 100 150 200 250 Bank Position (steps)

NE-1187 S1C16 Startup Physics Tests Report Page 23 of 56

Figure 3.2 SURRY UNIT 1 - CYCLE 16 STARTUP PHYSICS TESTS CONTROL BANK B DIFFERENTIAL ROD WORTH - HZP ALL OTHER RODS WITHDRAWN 12

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.... * -... ....... .... -Predicted i

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~

r-1

• ri i&.

,IJ I -

s::

~

-~ GI 1,1 I

GI

~

~

\

• ri 4 A

.. ~

\

Ij '

2 **

I1 '

'/ I i

I

\

0 0

' 50 100 150 200 250 Bank Position (steps)

NE-1187 S 1C 16 Startup Physics Tests Report Page 24 of 56

e SECTI0N4 BORON ENDPOINT AND WORTH MEASUREMENTS

_: I Boron Endpoint With the reactor critical at hot zero power, reactor coolant system .(RCS) boron concentrations were measured at selected rod bank configurations to enable a direct comparison of measured boron endpoints with design predictions. For each critical boron concentration measurement, the RCS conditions were stabilized with the control banks at or very near a selected endpoint position. Adjustments to the measured critical boron concentration values were made to account for off-nominal control rod position and moderator temperature, if necessary.

The results of these measurements are given in Table 4.1. As shown in this table and in the Startup Physics Test Results and Evaluation Sheets given in the Appendix, the

• f.., *.

measured critical boron endpoint values were within their respective design tolerances.

The all-rods-out (ARO) endpoint comparison to the predicted value met the requirements of Technical Specification 4.10.A regarding core reactivity balance. In summary, the boron endpoint results were satisfactory.

Boron Worth Coefficient The measured boron endpoint values provide stable statepoint data from which the boron worth coefficient or differential boron worth (DBW) was determined. By relating each endpoint concentration to the integrated rod worth present in the core at the NE-1187 S 1C 16 Startup Physics Tests Report Page 25 of 56

e

• time of the endpoint measurement, the value of the DBW over the range of boron endpoint concentrations was obtained.

A summary of the measured and predicted DBW is shown in Table 4.2. As indicated in this table and in the Appendix, the measured DBW was well within the design tolerance of +/-10%. In summary, the measured boron worth coefficient was satisfactory'.

I NE-1187 S 1C 16 Startup Physics Tests Report Page 26 of 56

e Table 4.1 SURRY UNIT 1 -CYCLE 16 STARTUP PHYSICS TESTS BORON ENDPOINTS

SUMMARY

Measured Predicted Difference Control Rod Endpoint Endpoint M-P Confi~ration (ppm) (ppm) (ppm)

ARO 1952 1939 13 l:.,.,i B Bank In 1766 1760* 6

• The predicted endpoint for the B Bank In configuration was adjusted for the difference between the measured and predicted values of the endpoint taken at the ARO configuration as shown in the boron endpoint Startup Physics Test Results and Evaluation Sheet in the Appendix.

~,. *'

NE-1187 SlC16 Startup Physics Tests Report Page 27 of 56

Table 4.2 SURRY UNIT 1 - CYCLE 16 STARTUP PHYSICS TESTS BORON WORTH COEFFICIENT Measured Predicted Percent Boron Worth Boron Worth Difference (%)

(pcm/ppm) (pcm/ppm) (M-P)/P x 100

-7.38 -7.14 3.4 NE-1187 S 1C 16 Startup Physics Tests Report Page 28 of 56

e SECTION 5 TEMPERATURE COEFFICIENT MEASUREMENT

.:-i The isothermal temperature coefficient (ITC) at the all-rods-out condition is measured by controlling the reactor coolant system (RCS) temperature through varying the steam generator blowdown flow, establishing a constant heatup or cooldown rate, and monitoring the resulting reactivity changes on the reactivity computer. This test sequence includes a cooldown followed by a heatup.

Reactivity was measured during the RCS cooldown of 3.0°F and RCS heatup of 3.0°F. Reactivity and temperature data were taken from the reactivity computer and strip chart recorders. Using the statepoint method, the temperature coefficient was determined by dividing the change in reactivity by the change in RCS temperature. An X-Y plotter, which plotted reactivity versus temperature, confirmed the statepoint method in calculating the measured ITC.

The predicted and measured isothermal temperature coefficient values are compared in Table 5.1. As can be seen from this summary and from the Startup Physics Test Results and Evaluation Sheet given in the Appendix, the measured isothermal temperature coefficient value was within the design tolerance of +/-3 pcml°F. Accounting for the Doppler temperature coefficient (-1.70 pcmf'F) and a 0.5 pcmf'F uncertainty, the NE-1187 S1C16 Startup Physics Tests Report Page 29 of 56

moderator temperature coefficient was 0.87 pcmfF, which meets the requirement of Core Operating Limits Report Section 2.1. In summary, the measured results were satisfactory.

NE-1187 SIC 16 Startup Physics Tests Report Page 30 of 56

e

• Table 5.1 SURRY UNIT 1 - CYCLE 16 STARTUP PHYSICS TESTS ISOTHERMAL TEMPERATURE COEFFICIENT

SUMMARY

ISOTHERMAL TEMPERATURE BANK TEMPERATURE BORON COEFFICIENT (PCMl°F)

PosmoN RANGE CONCENTRATION AVE DIFFER (STEPS) (°F) (oom) CID WU MEAS PRED (M-P)

I 544.3 D/209 to 1948 -0.83 -0.83 -0.83 -1.75 0.92 547.3 NE-1187 SlC16 Startup Physics Tests Report Page 31 of 56

>.*,-:t

SECTION6 POWER DISTRIBUTION MEASUREMENTS The core power distributions were measured using the moveable incore detector flux mapping *system. This system consists of five fission chamber detectors, which traverse fuel assembly instrumentation thimbles depicted in Figure 1.3. For each traverse, the detector voltage output is continuously monitored on a strip chart recorder, and scanned for 61 discrete axial points by the PRODAC P-250 process computer. Fun core, three-dimensional power distributions are determined from this data using the CECOR code4* CECOR couples the measured voltages with predetermined analytic signal-to-power conversions, pin-to-box factors, and average coupling coefficients in order to determine the power distribution for the whole core.

A list of the full-core flux maps taken during the startup test program and the measured values of the important power distribution parameters are given in Table 6.1. A comparison of these measured values with their Technical Specification limits is given in Table 6.2. Flux map 2 was taken at approximately 28% power to verify the radial power d1stribution (RPD) predictions at low power. Figure 6.1 shows the measured RPDs from this flux map. Flux maps 3 and 4 were taken near 70% and 100% power, respectively, with different control rod configurations. These flux maps were taken to check at-power design :predictions and to measure core power distributions at various operating conditions. The radial power distributions for these maps are given in Figures 6.2 and NE-1187 S 1C16 Startup Physics Tests Report Page 32 of 56

e **

6.3. These figures show that the average relative assembly power distribution measured/predicted percent difference was 1.4% or less for the three maps. The measured F-Q(Z) and F-DH(N) peaking factor values for all flux maps were within the limits of the Core Operating Limits Report (COLR) Sections 2.3 and 2.4, respectively.

All three flux maps were used to recalibrate the power range excore detectors.

In conclusion, the power distribution measurement results were considered to be acceptable with respect to the design tolerances, the accident analysis acceptance criteria, ..

and the COLR limits. It is therefore anticipated that the core will continue to operate as designed throughout Cycle 16.

NE-1187 S 1C 16 Startup Physics Tests Report Page 33 of 56

Table 6.1 SURRY UNIT 1 -CYCLE 16 STARTUP PHYSICS TESTS INCORE FLUX MAP

SUMMARY

Bum Bank F-Q(Z) Hot (1) F-DH(N) Hot Core F(Z) (2) Axial No.

Map Map up Pwr D Channel Factor Channel Factor Max Core Tilt Off of Description No. Date MWD/  % Steps Assy Axial F-Q(Z) Assy F-DH(N) Axial F(Z) Max Loe Set Thim MTU Point point (%) bles Less thn 30% Pwr 2 11/21/98 8 28 .171 Fil 30 2.142 Fil 1.523 26 1.302 1.0104 NE 0.715 44 Btwn 65% and 75% 3 11/26/98 35 69 193 Fil 30 1.929 Fl I 1.482 30 1.205 1.0095 NE -0.354 45 Grt than 95% Pwr 4 12/02/98 175 99 227 Fl I 32 1.852 Fl I 1.470 30 1.164 1.0056 NE -0.893 45 NOTES: Hot spot locations are specified by giving assembly locations (KG. H-8 is the center-of-core assembly) and core height (in the "Z" direction the core is divided into 61 axial points starting from the top of the core).

(1) F-Q(Z) includes a total uncertainty of 1.08.

(2) CORE TILT- defined as the average quadrant power tilt from CECOR.

(3) MAPS 2, 3, and 4 were used for power range detector calibrations.

NE-1187 S 1C 16 Startup Physics Tests Report Page 34 of 56

Table 6.2

• tj, J SURRY UNIT 1 - CYCLE 16 STARTUP PHYSICS TESTS COMPARISION OF MEASURED POWER DISTRIBUTION PARAMETERS

• WITH THEIR CORE OPERATING LIMITS Peak F-Q(Z) Hot F-Q(Z) Hot F-DH(N) Hot Map Channel Factor* Channel Factor** Channel Factor (At Node of Minimum Margin)

No. Meas. Limit Node Meas. Limit Node Margin Meas. Limit Margin

(%) (%)

2 2.142 4.628 30 2.130 4.582 26 53.5 1.523 1.898 19.8 3 1.929 3.373 31 1.896 3.297 21 42.5 1.482 1.707 13.2 4 1.852 1.153 32 1.852 2.341 30 20.9 1.470 1.565 6.1

• The Core Operating Limit for the heat flux hot channel factor, F-Q(Z), is a function of core height and power level. The value for F-Q(Z) listed above is the maximum value of F-Q(Z) in the core. The COLR limit listed above is evaluated at the plane of maximum F-Q(Z). . .

• • The value for F-Q(Z) listed above is the value at the plane of minimum margin, The **~ 1 minimum margin values listed above are the minimum percent difference between the measured values of F-Q(Z) and the COLR limit for each map.

The measured F-Q(Z) hot channel factors include 8% uncertainty as defined in Tech.

Spec 3.12.B.

NE-1187 S 1C 16 Startup Physics Tests Report Page 35 of 56

Figure 6.1 SURRY UNIT 1 - CYCLE 16 STARTUP PHYSICS TESTS ASSEMBLYWISE POWER DISTRIBUTION 28%POWER R p N M L K J H G F E D C B A Predicted Predicted Measured 0.247 0.23S 0.246 Measured 0.2SO 0.236 0.248 Pct Difference 0.992 0.406 0.600 Pct Difference 0.301 O.S63 1.073 0.824 1.069 0.S61 0.299 0.302 0.568 1.085 0.826 1.076 O.S67 0.304 0.488 0.928 1.129 0.197 0.6SI 1.0S2 1.560 2 0.271 1.()1)() 1.314 1.233 1.288 1.230 1.310 1.084 0.268 0.280 1.094 1.326 1.256 1.282 1.239 1.328 1.104 0.286 3.3S5 0.317 0.894 1.868 -0.4SO 0.747 1.391 1.851 6.512 3 0.268 0.630 1.325 1.272 1.325 1.245 1.322 1.270 1.320 0.631 0.271 0.275 0.634 1.313 1.277 1.338 1.255 1.340 1.292 1.338 0.633 0.270 2.445 0.724 -0.899 0.360 0.938 0.821 1.361 1.731 1.385 0.343 * -0.360 4 0.299 1.083 1.316 1.294 l.34S 1.211 l.31S 1.212 1.346 1.296 1.324 1.()91 0.302 0.307 1.124 1.337 1.312 1.353 1.217 1.333 1.237 1.382 l.311 1.300 1.084 0.308 2.422 3.732 I.S84 1.389 0.612 0.S40 l.41S 2.0S7 2.685 1.118 -1.791 -0.6S2 2.051 5 O.S61 1.310 1.267 1.340 1.161 1.206 I.ISi 1.214 1.166 1.344 1.272 1.317 0.S65 0.562 1.316 1.271 1.344 1.161 1.208 1.176 1.248 1.236 1.372 1.273 1.314 O.S67 i;,

0.073 0.422 0.264 0.29S -0.068 0.190 2.IS6 2.781 6.031 2.117 0.089 -0.239 0.3S2 .;,, 6 0.246 1.070 1.233 1.325 1.211 1.210 I.ISO 1.108 1.154 l.21S 1.213 1.326 1.238 1.080 0.249 0.248 _1.061 1.205 1.310 1.19S 1.196 1.128 I.Ill 1.163 1.247 1.231 1.329 1.231 1.087 0.249 0.S97 -0.875 -2.256 -1.124 -1.317 -1.183 -1.907 0.242 0.763 2.640 1.460 0.185 -0.533 0.619 0.306 7 0.236 0.829 1.293 1.247 1.317 1.155 1.113 l.04S 1.113 I.IS6 l.318 1.248 1.294 0.830 0.236 0.254 0.828 1.294 1.231 1.275 1.138 1.103 1.044 1.119 1.182 l.330 1.244 1.261 0.817 0.233 7.626 -0.099 0.029 -1.337 -3.192 -1.465 -0.829 -0.070 0.512 2.208 0.949 -0.341 -2.542 -I.S09 -1.004 8 0.249 1.080 1.238 1.326 1.213 l.21S I.ISS I.Ill 1.153 1.214 1.213 1.326 1.234 1.071 0.247 0.250 1.072 1.225 1.313 1.208 1.210 1.157 1.109 1.143 1.215 1.218 1.326 1.225 1.063 0.243 0.761 -0.687 -1.029 -0.981 -0.413 . -0.437 0.201 -0.190 -0.902 0.081 0.424 -0.003 -0.738 -0.767 -1.359 9 0.56S 1.317 1.272 1.344 1.167 1.216 1.155 1.214 1.166 1.344 1.269 1.312 0.S62 0.556 1.286 1.256 1.337 1.166 1.242 1.156 1.215 1.177 1.353 1.276 1.317 0.564

-1.615 -2.366 -1.247 *0.574 -0.094 2.133 0.039

• 0.076 0.934 0.678 0.498 0.355 0.328 IO 0.302 1.091 1.324 1.297 1.348 1.215 1.319 1.215 1.349 1.297 1.319 1.085 0.300 0.297 1.079 1.311 1.288 1.333 1.198 1.296 1.218 1.395 1.307 1.333 1.099 0.303

-1.370 -1.147 -0.959 -0.666 -1.096 -1.381 -1.727 0.188 3.367 0.706 I.OS9 1.266 0.887 11 0.271 0.631 1.321 1.271 1.325 1.248 1.329 1.275 1.328 0.631 0.269 0.282 0.626 1.302 1.241 1.266 1.218 1.316 1.273 1.330 0.646 0.288 3.931 *0.873 -1.441 -2.364 -4.455 -2.413 -0.980 -0.220 0.142 2.315 7.054 12 0.269 1.086 1.312 1.232 1.291 1.236 1.317 1.093 0.271 0.264 1.066 1.282 1.196 1.260 1.216 1.273 1.076 0.270

-1.694 -1.760 -2.317 -2.922 -2.401 -1.627 -3.392 -I.S02 -0.375 13 0.300 0.562 1.070 0.826 1.076 0.564 0.302 0.297 0.549 1.046 0.813 1.083 0.555 0.297

• 0.974 *2.226 -2.323 -1.599 0.685 -1.590 .1.599 14 0.247 0.236 0.248 0.241 0.232 0.248 STANDARD -2.312 -1.448 0.273 AVERAGE 15 DEVIATION PCT DIFFERENCE

= 1.268 = 1.3

SUMMARY

MAP NO: Sl-16-02 DATE: 11/21/98 POWER: 27 .68%

CONTROL ROD POSITIONS: F-Q(Z) = 2.142 QPTR:

• D BANK AT 171 STEPS F-DH(N) = 1.523 NW 1.0009 I NE 1.0104 F(Z) = 1.302 SW 0.9887 I SE 1.0001 BURNUP = 8.0 A.O. = 0.715%

NE-1187 S 1C 16 Startup Physics Tests Report Page 36 of 56

Figure 6.2 SURRY UNIT 1 - CYCLE 16 STARTUP PHYSICS TESTS ASSEMBLYWISE POWER DISTRIBUTION 69%POWER R p N M L K J H G F E D C B A Predicted Predicted Measured 0.265 0.256 0.264 Measured ¥,..":J 0.261 0.253 0.262 Pct Difference -1.251 -1.356 --0.596 Pct Difference 0.308 0.572 1.089 0.874 1.085 0.571 0.307 0.302 0.565 l.076 0.861 l.081 O.S78 0.311

-1.761 -1.362 -1.214 -1.S22 --0.400 1.371 l.423 2 ~--:~

0.274 1.067 l.281 1.223 l.280 1.220 1.2n l.062 0.271 0.281 l.047 1.264 l.212 1.246 l.21S l.289 l.078 0.284 2.543 -l.903 -1.342 --0.929 -2.632 --0.409 0.882 I.SIS 4.689 3 0.271 0.629 l.283 1.244 1.299 1.232 l.297 l.243 1.279 0.629 0.274 0.273 0.623 l.239 l.228 l.293 1.227 1.304 l.259 1.298 0.636 0.27S

""':.c..t.

0.838 --0.876 -3.443 -1.316 --0.489 --0.386 O.S36 1.344 1.471 1.163 0.481 4 0.307 1.061 1.276 1.270 1.331 1.213 1.307 1.214 1.332 1.271 1.282 1.068 0.308 0.310 l.084 1.286 l.288 1.332 l.220 1.320 1.233 1.365 1.293 1.287 l.071 0.306 l.291 2.181 0.761 1.380 0.073 0.566 1.049 l.S25 2.4SS l.666 0.330 0.298 --0.888 5 0.571 1.2n 1.240 1.327 1.217 1.238 1.178 1.246 1.221 1.330 l.244 1.283 0.574

,F * * *

• 0.570 1.278 1.236 1.313 1.211 1.240 1.206 1.272 l.28S l.361 1.256 1.289 0.578

--0.136 0.034 --0.389 -1.080 --0.504 0.145 2.431 2.069 5.272 2.359 0.953 0.473 0.642 6 0.264 1.086 1.223 l.299 1.213 l.243 1.190 I.ISO 1.194 1.247 l.215 1.299 1.227 1.095 0.266 0.264 1.080 1.206 1.285 1.194 1.227 1.168 1.148 1.181 1.271 1.233 1.307 l.227 1.110 0.269 0.269 --0.562 -1.429 -1.074 -1.567 -1.265 -1.865 --0.144 -1.047 1.972 l.521 0.640 0.000 l.352 1.136 7 f** ~

0.256 0.878 l.284 1.233 1.308 1.181 1.154 1.090 1.154 1.182 1.309 1.234 1.284 0.878 0.257 0.267 0.878 l.290 1.220 1.272 1.168 1.147 l.090 1.157 1.206 l.324 1.233 1.25S 0.882 0.257 4.007 --0.034 0.501 -l.039 -2.739 -1.127 --0.550 0.010 0.247 2.012 l.ll8 --0.024 -2.255 0.3n 0.232 8 0.266 1.095 l.227 1.299 1.215 l.247 1.195 1.152 1.193 l.245 1.214 1.300 1.224 l.087 0.264 0.266 l.08S 1.213 l.290 l.217 1.249 1.205 1.158 1.194 l.256 1.225 1.305 l.219 l.084 0.259 0.05S --0.856 -1.092 --0.726 0.237 0.219 0.849 O.S48 0.053 0.881 0.883 0.424 --0.393 --0.261 -1.673 9 0.574 l.283 l.244 1.330 l.221 l.247 1.181 l.245 1.220 l.329 1.242 1.278 OS71 0.562 1.241 1.230 1.332 l.232 1.286 1.196 l.256 l.237 l.344 l.255 1.286 0.569

-2.110 -3.226 -1.112 0.128 0.903 3.058 1.291 o.8n 1.319 l.143 1.071 0.570 --0.369 10 0.308 l.068 1.283 1.272 l.333 1.216 1.309 1.216 1.334 l.272 1.278 1.062 0.307 0.303 1.054 l.276 l.276 1.337 1.222 1.318 1.227 l.367 1.287 1.298 1.078 0.309

-1.639 -1.286 --0.541 0.347 0.262 0.521 0.695. 0.850 2.447 l.153 1.601 1.499 0.784 11 0.274 0.629 1.280 1.243 l.298 l.233 l.301 l.246 1.285 0.629 0.271  ;,~- l 0.286 0.628 l.275 l.233 l.276 1.218 1.294 1.244 l.292 0.651 0.288 4.529 --0.157 --0.420 --0.870 -1.723 -1.253 --0.587 -0.199 0.543 3.497 6.115 12 0.272 1.063 1.278 1.221 1.281 l.225 l.283 l.068 0.274 0.269 l.051 1.260 l.195 1.239 1.202 l.242 l.056 0.275

-1.005 -l.ll5 -1.450 -2.196 -3.331 -1.878 -3.190 -1.180 0.351 13 ,:,.~:

0.307 0.571 l.086 0.875 1.091 0.573 0.308

• 0.298 0.560 l.062 0.857 l.089 0.563 0.304

-2.920 -1.989 -2.213 -2.078 --0.139 -1.n1 -1.447 14 0.264 0.256 0.265 0.255 0.251 0.264 STANDARD -3.451 -2.101 -0.526 AVERAGE 15 DEVIATION PCT DIFFERENCE

= 1.061 = 1.3

SUMMARY

MAP NO: Sl-16-03 DA TE: 11/26/98 POWER: 68.61%

CONTROL ROD POSITIONS: F-Q(Z) = 1.929 QPTR:

D BANK AT 193 STEPS F-DH(N) = 1.482 NW 0.9931 I NE 1.0095 F(Z) = 1.205 SW 0.9945 ' SE 1.0029 BURNUP= 35 A.O. = -0.354 %

NE-1187 S 1C 16 Startup Physics Tests Report Page 37 of 56

\

Figure 6.3 SURRY UNIT 1 - CYCLE 16 STARTUP PHYSICS TESTS ASSEMBLYWISE POWER DISTRIBUTION 99%POWER R p N M L K J H G F E D C B A Predicted Predicted Measured 0.275 0.271 0.273 Measured 0.271 0.267 0.272 Pct Difference -1.201 -1.319 -0.678 Pct Difference 0.308 0.572 1.090 0.915 1.086 0.571 0.307 0.304 0.565 1.077 0.902 1.081 0.577 0.311

-1.493 -1.250 -1.152 -1.419 -0.487 U)!i5 1.218 2 0.273 1.041 1.251 l.2IO 1.272 1.208 1.248 1.036 0.270 0.284 1.025 1.236 1.199 1.241 1.202 1.256 1.050 0.291 4.250 -1.580 -1.240 -0.902 -2.462 -0.508 0.629 1.347 7.577 3 0.270 0.624 1.254 1.226 1.283 1.223 1.280 1.224 1.250 0.625 0.273 0.274 0.621 1.214 1.209 1.274 1.217 1.285 1.235 1.259 0.624 0.270 1.421 -0.507 -3.195 -1.332 -0.662 -0.472 0.351 0.902 0.730 -0.158 --0.981 4 0.307 1.035 1.247 1.257 1.330 1.220 1.305 1.222 1.331 1.258 1.253 1.041 0.309 0.312 1.064 1.258 1.271 1.327 1.219 1.317 1.239 1.357 1.264 1.222 1.028 0.312 1.708 2.740 0.813 1.152 -0.205 -0.096 0.919 1.396 1.889 0.449 -2.491 -1.283 1.063 5 0.571 1.248 1.222 1.327 1.280 1.268 1.200 1.276 1.284 1.329 1.225 1.252 0.574 0.571 1.248 1.214 1.304 1.271 1.270 1.233 1.307 1.344 1.350 1.220 1.243 0.572 ). .

--0.076 -0.006 -0.621 -1.703 -0.764 0.178 2.743 2.458 4.661 1.549 -0.421 -0.764 -0.332 6 0.273 1.087 1.210 1.282 1.220 1.272 1.223 1.182 1.226 1.276 1.222 1.282 1.213 1.095 *:,0.276 0.275 1.081 1.188 1.264 1.196 1.257 1.209 1.192 1.237 1.308 1.238 1.282 1.202 1.093 0.274 0.679 -0.555 -1.794 -1.368 -1.946 -1.214 -1.083 0.825 0.922 2.501 1.316 -0.035 -0.941 -0.126 -0.502 7 ,,,

0.271 0.918 1.276 1.224 1.307 1.204 1.185 1.126 1.186 1.205 1.307 1.225 1.276 0.919 0.271 0.287 0.921 1.282 1.263 1.191 1.135 1.202 1.241 1.324 1.222 1.242 0.902 0.266 1.209 1.185

'\!'*

5.869 0.246 0.493 -1.260 -3.336 -1.084 -0.049 0.769 1.374 3.013 I.JOO -0.251 -2.680 . -1.759 -1.763 8  ;.i,y 0.276 1.094 1.213 1.282 1.222 1.276 1.227 1.183 1.225 1.274 1.221 1.283 1.211 1.087 0.273 0.277 1.088 1.201 1.273 1.226 1.282 . 1.243 1.195 1.235 1.284 1.233 1.289 1.205 1.079 0.266 0.645 -0.609 -0.980 -0.671 0.363 0.437 1.270 0.952 0.790 0.713 0.996 0.480 -0.453 -0.709 -2.642 9 0.574 1.252 1.225 1.329 1.285 1.277 1.203 1.275 1.284 1.328 1.223 1.249 0.571 0.563 1.214 1.213 1.333 1.299 1.323 1.215 1.286 1.302 1.349 1.240 1.263 0.579

-1.886 -3.079 -0.939 0.306 1.135 3.630 0.984 0.864 1.401 1.561 1.415 1.146 1.288 10 0.309 1.041 1.253 1.259 1.332 1.223 1.308 1.223 1.333 1.259 1.249 1.036 0.307 0.305 1.032 1.249 1.264 1.336 1.224 . 1.289 1.226 1.366 1.287 1.278 1.062 0.313

-1.309 -0.909 -0.289 0.451 0.304 0.124 -1.420 0.254 2.479 2.234 2.350 2.504 2.030 11 0.273 0.625 1.251 1.225 1.281 1.224 1.284 1.227 1.255 0.625 0.270 0.290 0.627 1.249 1.216 1.257 1.202 1.273 1.225 1.267 0.649 0.297 6.485 0.321 -0.134 -0.722 -.1.884 -1.794 -0.882 -0.216 0.941 3.945 9.904 12 0.270 1.037 1.249 1.209 1.273 1.212 1.252 1.042 0.273 0.269 I.OJI 1.235 1.186 1.236 1.189 1.208 1.031 0.275

-0.4IO -0.566 -1.085 -1.908 -2.909 -1.892 -3.532 -1.115 0.763 13 0.307 0.571 1.087 0.915 1.091 ().573 0.309 0.305 0.564 1.()69 0.91JI l.()'JI 0.562 0.304

-0.602 -1.299 -1.627 -1.586 0.000 -1.811 -1.359 14 0.274 0.271 0.275 0.270 0.267 0.274 STANDARD -1.459 -1.410 -0.321 AVERAGE 15 DEVIATION PCT DIFFERENCE

= 1.354 = 1.4

SUMMARY

MAP NO: S!-16-04 DATE: 12/02/98 POWER: 98.8%

CONTROL ROD POSITIONS: F-Q(Z) = 1.852 QPTR:

D BANK AT 227 STEPS F-DH(N) = 1.470 NW 0.9934 I NE 1.0056

_______ I_______

F(Z) = 1.164 SW 0.9967 I SE 1.0044 BURNUP= 175 A.O. = -0.893%

NE-1187 S 1C 16 Startup Physics Tests Report Page 38 of 56

e

~

SECTION?

REFERENCES

.~* i

1. T.S. Psuik, "Surry Unit 1, Cycle 16 Design Report", Technical Report NE-1177, Revision 0, Virginia Power, November, 1998.

2. T. K. Ross, W. C. Beck, "Control Rod Reactivity Worth Determination By The Rod Swap Technique," VEP-FRD-36A, December, 1980.

3. Letter from W. L. Stewart (Virginia Power) to the U.S.N.R.C, "Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2: Modification of Startup Physics Test Program - Inspector Followup Item 280, 281/88-29-01 ", Serial No.89-541, December 8, 1989.

4. T. W. Schleicher, "The Virginia Power CECOR Code Package", Technical Report*

NE-831, Revision 4, Virginia Power, August, 1998.

5. Surry Unit 1 and 2 Technical Specifications, Sections 3.1.E.1, 3J2.B.l, 3.12.C.1, 4.10.A, and 5.3.A.6.b.

6. R. W. Twitchell, "Surry 1, Cycle 16 TOTE Calculations", PM-777, Revision 0, November, 1998.

~-***}

7. P. D. Banning, "Surry 1, Cycle 16 Flux Map Analysis", PM-781,.

Revision 0, and Addenda, November - December, 1998.

8. R.W. Twitchell, "Surry 1, Cycle 16 RSAC Calculations", PM-760 Revision 0, August 1998.

9. R.W. Twitchell, "Surry!, Cycle 16 Design Report Calculations",

PM-774, Revision 0, November 1998.

NE-1187 S 1C 16 Startup Physics Tests Report Page 39 of 56

APPENDIX STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEETS NE-1187 SIC 16 Startup Physics Tests Report Page 40 of 56

• * -:,. \) **.

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET I Test

Description:

Zero Power Testing Range Determination -.* .

Reference Proc No / Section: 1-NPT-RX-008 Sequence Step No:

II Bank Positions (Steps) RCS Temperature (°F): 547 Test Power Level(% F.P.): 0 Conditions SDA: 229 SDB: 229 CA: 229 Other (specify):

(Design) CB: 229 CC:

• CD:

• Below Nuclear Heating Ill Bank Positions (Steps) RCS Temperature (°F): !f'I 7. ::;-

Test Power Level(% F.P.): 0 Conditions SDA: 229 SOB: 229 CA: 229

• Other (specify):

(Actual) CB: 229 CC: 1,.1.,'1 CD: It? Below Nuclear Heating Datemme Test Performed:

11/ie,/<ts

. 0&40

'(-:_;

Reactivity Computer Initial Flux Background Reading os amps "'-.. ,

IV .A g,,f;v./(1/i.* (.*e .(c."'7'". !i'er '" I,. 1-1 1~q i< 10-q A,irl's" Test Results Flux Reading At 3JJ;ctfJ

-7amps Point Of Nuclear Heating h:10-~

-e .

Zero Power Testing Range to 1ox* 10 amps Reference Not Applicable V FSAR/Tech Spec Not Applicable Acceptance Criteria Reference Not Applicable

~ YES NO Design Tolerance is met** --

Acceptance Criteria is met** : -VYES NO VI

• At The Just Critical Position Comments ** Design Tolerance and Acceptance Criteria are met if ZPTR is below the/f int of Nuclear Heating and above background.

/ ,1 ,.JJ Prepared By: 1flA 1 ft/_,J,{I f 'V \

Reviewed By: C~ ~

NE-1187 SIC 16 Startup Physics Tests Report Page 41 of 56

SURRY . POWER STATION UNIT 1 CYCLE -

16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET I Test

Description:

Reactivity Computer Checkout . -

Reference Proc No I Section: 1-NPT-RX-008 Sequence Step No:

II Bank Positions (Steps) RCS Temperature (uF): 547 Test Power Level (% F.P.): O Conditions SDA: 229 SOB: 229 CA: 229 Other (specify):

{Design) CB: 229 CC:

• CD:

• Below Nuclear Heating Ill Bank Positions {Steps) RCS Temperature (°F): S'/7.o Test Power Level (% F.P.): 0 Conditions SDA: 229 SOB: 229 CA:. 229 Other (specify): -

(Actual) CB: 229 CC: ZZ.Cf CD: 117 Below Nuclear Heating

. Date/Time Test Performed:

- 11 I t'i '" i O 7: 2. ~ ,.

Measured Parameter Pc= Measured Reactivity using p-camputer (Description) Pt= Predicted Reactivity

. IV Test Results Measured Value Pc= -47. 0 I +.4q Pt= -47, 8 , -t-l{q,(.

%0= -J.7e/o 1 - /. Z. "lo Design Value %0= {(Pc - pJ/pJ X 100% S 4.0 %

Reference WCAP 7905, Rev. 1, Table 3.6 V FSAR/Tech Spec Not Applicable Acceptance Criteria Reference Not Applicable

./ .

Design Tolerance is* met VYES -- NO Acceptance Criteria is met : £/'YES NO VI

• At The Just Critical Position Comments The allowable range will be set based on the above results, as well as results from the benchmark test.

Allowable Range = - L/7, 0 +o +-'{Cf, o pew, Prepared By: ~ ...f. {2,.J. Reviewed B y ~ ~

NE-1187 S 1C 16 Startup Physics Tests Report Page42 of 56

• ~ ..

• e SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET I Test

Description:

Critical Boron Concentration -ARO '

Reference Pree No / Section: 1-NPT-RX-008 Sequence Step No:

II Bank Positions {Steps) RCS Temperature (°F): 547 Test Power Level (% F.P.): O Conditions SDA: 229 SDB: 229 CA:' 229 Other (specify):

(Design) CB: 229 CC: 229 CD: 229 Below Nuclear Heating Ill Bank Positions (Steps) RCS Temperature (°F): S'-/7."S Test Power Level (% F.P.): O Conditions SDA: 229 SOB: 229 CA: 229

• Other (specify):

(Actual) CB: 229 CC: 229 CD: 229 Below Nuclear Heating Datemme Test Performed:

(l I 1e;/91 O?:o~

Measured Parameter (Ca)MARO; Critical Boron Concentration - ARO (Description)

IV Test *;::,,,

Results Measured Value (Ca)MARo= i'i52 ppm (Design Conditions)

Design Value C 8 = 1939 +/- 50 ppm (Design Conditions)

Reference Technical Report NE-1177, Rev. 0 laC8 x Ca 1~ 1000 pcm 0

V FSARITech Spec Acceptance Criteria Reference Technical SpE:_cification 4.10.A Design Tolerance is met v YES -- NO Acceptance Criteria is met : VvEs NO VI Comments aCe = -7.05 pcm/ppm Ce°= l(Ce)MARo - C8I; Ce is design value Prepared By: ~ 2.P~ Reviewed B y : ~ / ~

NE-1187 S 1C 16 Startup Physics Tests Report Page 43 of 56

-~----.

SURRY. POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET I Test

Description:

HZP Boron Worth Coefficient Measurement ,

Reference Proc No/ Section: 1-NPT-RX-008 Sequence Step No:

II Bank Positions {Steps) RCS Temperature (°F): 547 Test Conditions 1----------------1 SDA:. 229 SOB: 229 CA:. 229 Power Level (% F.P.): O Other {specify):

(Design) CB: moving CC:

• 229 CD: 229 Below Nuclear Heating Ill Bank Positions (Steps) RCS Temperature (0 F): 5'0.:,

Test

---

1 Power Level {% F.P.): 0 Conditions SDA:. 229 SDB: 229 CA:. 229 Other (specify):

(Actual} CB:

• moving CC: 229 CD: 229 Below Nuclear Heating 1--.:...--"""'--t-----------------t Date11jme Test Performed:

/ I JI °I /~v J5; I ~

Measured Parameter aCa; Boron Worth Coefficient (Description)*

IV Test Results Measured Value aCa= -7. 3y pcm/ppm .*

Design Value aC 8 = -7.14 +/- 0.71 pcm/ppm  :*

(Design Conditions)

Reference Technical Report NE-1177, Rev. 0 V FSAR/Tech Spec Not Applicable Acceptance Criteria 1-----------+-----------------1 Reference Not Applicable Design Tolerance is met ~YES _NO Acceptance Criteria is met : ~ YES NO VI Comments Reviewed By: -::V ~

NE- l l S7 SIC 16 Startup Physics Tests Report Page 44 of 56

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET I Test Descn-ption: Isothermal Temperature Coefficient-ARO .

Reference Pree No I Section: 1-NPT-RX-008 Sequence Step No:

II Bank Positions (Steps) RCS Temperature (°F): 547 Test Power Level (% F.P.): O Conditions SDA: 229 SOB: 229 -CA: 229 Other (specify):

(Design) CB: 229 CC: 229 CD: 229 Below Nuclear Heating Ill Bank Positions (Steps) RCS Temperature (°F): 5"/./7. 3 Test Power Level (% F.P.): O Conditions SDA: 229 SOB: 229 CA: 229 Other (specify):

{Actual) CB: 229 CC: 229 CD: 2o9 Below Nuclear Heating.

Datemme Test Performed:

ti/1'1/c:tf 10;27 Measured Parameter (a'TISO)ARO,* Isothermal Temperature (Description) Coefficient - ARO IV Test Measured Value (a'TISO)ARO-_ -o. '83 pcml°F Results (Ce= / 'IJ/8 ppm)

Design Value

)ARo= - /. 7 5 +/-3.0 pcml°F 150 (Actual Conditions) (aT (Ce= l't'I~ ppm)

Design Value (Design Conditions) (a'TISO)ARO-_ -1.83 +/-3.0 pcml°F (Ce= 1939 ppm)

Reference Technical Report NE-1177, Rev. 0 150 V FSAR/COLR ar ~ 3.80

• pcm/ °F Acceptance aTooP = -1.70 pcml°F Criteria Reference COLR 2.1.1,T~chnical Report NE-1177, Rev. 0 Design Tolerance is met V YES --- NO Acceptance Criteria is met : ~YES - - NO VI Comments *Uncertainty on aTMoo= 0.5 pcml°F (

Reference:

memorandum from C.T. Snow to E.J. Lozito dated June 27, 1980.)

Prepared By: .,.~71 --ft~ '>?.f)Jl Reviewed B~~-~ -

NE-1187 S 1C 16 Startup Physics Tests Report Page 45 of 56

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET I Test

Description:

Control Bank B Worth Measurement, Rod Swap R~f. Bank Reference Proc No I Section:

• 1-NPT-RX-008 Sequence Step *No:

II Bank P~sitions (Steps) RCS Temperature (°F): 547 Test Power Level(% F.P.): 0 Conditions SOA: 229 SOB: 229 Ck 229 Other (specify):

(Design} CB: moving CC: 229 CD: 229 Below Nuclear Heating Ill Bank Positions (Steps) RCS Temperature (°F): ~'17- l Test Power Level (% F.P.): O Conditions SDk 229 SOB: 229 Ck 229 Other (specify):

(Actual) CB: moving CC: 229 CD: 229 Below Nuclear Heating Datemme Test Performed:

11 //-,Ir.,,, // :1'1 Measured Parameter I REF. Integral Worth Of Control Bank B, * *

• B '

(Description) All Other Rods Out

,it IV ')*

Test Measured Value I REF_

/ 3 7'; pcm 1\'

B -

Results Design Value (Design Conditions) 1tEF= 1371 +/- 137 pcm Reference Technical Report NE-11n, Rev. 0 And Engineering Transmittal NAF 98-0183, Rev. 0 If Design Tolerance is exceeded, SNSOC shall .

V FSAR/Tech Spec evaluate impact of test result on safety analysis.

Acceptance SNSOC may specify that additional testing Criteria be performed.

Reference VEP-FRD-36A Design Tolerance is met Acceptance Criteria is met :

~YES YES

-- NO NO VI Comments Prepared By: Reviewed By: Z ~

NE-1187 SIC 16 Startup Physics Tests Report Page 46 of 56

"'~

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET I Test

Description:

Critical Boron Concentration - B Bank In .

Reference Proc No / Section: 1-NPT-RX-008 Sequence Step No:

II Bank Positions (Steps) RCS Temperature (°F): 547 *.:!

Test Power Level (% F.P.}: 0 Conditions SDA: 229 SOB: 229 CA: . 229 Other (specify):

(Design) CB: 0 CC: 229 CD: 229 Below Nuclear Heating Ill Bank Positions {Steps) RCS Temperature {°F): S 't ]. ;-

Test Power Level (% F.P.): 0 Conditions SDA: 229 SOB: 229 CA: 229 Other (specify):

(Actual) CB: 0 CC: 229 CD: 229 Below Nuclear Heating Datem,:ne Test Performed:

• 11/icr/ql' I

1S:1S-Measured Parameter (Ce)Me;

• Critical Boron Concentration, (Description)* B Bank In IV Test -,~:-\

Results Measured Value (Ce)Me= / ?f... (. ppm (Design Conditions)

Design Value CB= 1747 + ~CB Prev +/- (10 + 137.1/laCBI) ppm (Design Conditions) CB=~ +/- -29 ppm nc.o ;u-Reference Technical Report NE-1177, Rev. 0 V FSAR/fech Spec Not Applicable.

Acceptance Criteria Reference Not Applicable Design Tolerance is met ~YES -- NO Acceptance Criteria is met : YES NO VI Comments aCB = -7.14 pcm/ppm

~CB Prev = (CB)MARO -1939 ppm Prepared By: M 7'51i/ Reviewed By: ~ ~

NE-1187 S 1C 16 Startup Physics Tests Report Page 47 of 56

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET I Test

Description:

Shutdown Bank A Worth Measurement, Rod Swap Reference Proc No / Section: 1-NPT-RX-008 Sequence Step No:

II Bank P9sitions *(Steps) RCS Temperature (°F): 547 Test Power Level (% F.P.): O Conditions SDA: moving SDB: 229 CA: . 229 Other (specify):

(Design) CB: moving CC: 229 CD: 229 Below Nuclear Heating Ill Bank Positions (Steps) RCS Temperature (°F): 5'-/ 7. 'Z...

Test Power Level (% F.P.}: O Conditions SDA: moving SDB: 229 CA: 229 Other (specify):

(Actual} CB: moving CC: 229 CD: 229 Below Nuclear Heating Datemme Test Performed:

uI,~ )q~ n :15' ,.

Measured Parameter ISARs; Integral Worth of Shutdown B~nk A, (Description) Rod.Swap IV Measured Value ISARS= /010 (Adjusted Measured Test Critical Reference Bank Position = I 'IC, steps)

,I Results Design Value

-'*'/

(Actual Conditions) I RS_ /OJ, L (Adjusted Measured SA -

Critical Reference Bank Position = /'{ c,_ .\ steps)

Design Value *'

(Design Conditions) ISARs= 1032 +/- 155 pcm (Critical Reference Bank Position= 146 steps)

Reference Engineering Transmittal NAF 98-0183, Rev. 0, VEP-FRD-36A FSAR/Tech Spec If Design Tolerance is exceeded, SNSOC shall V evaluate impact of test result on safety analysis.

Acceptance SNSOC may specify that additional testing Criteria be performed. -

Reference f\{EP-FRD-36A Design Tolerance is met  :~ YES -- NO Acceptance*Criteria is met : YES NO VI Comments Prepared By: LJ/ ),JL/) Reviewed By: ~~

NE- l l 8 7 S 1C 16 Startup Physics Tests Report Page 48 of 56

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET

~

I Test

Description:

Shutdown Bank B Worth Measurement, Rod Swap Reference Pree No / Section: 1-NPT-RX-008 Sequence Step No:

II Bank P.ositions (Steps) RCS Temperature (°F}: 547 Test Power Level (% F.P.): O Conditions SDA: 229 SOB: moving CA: 229 Other (specify):

(Design) CB: moving CC: 229 CD: 229 Below Nuclear Heating Ill Bank Positions (Steps) RCS Temperature (°F): 5'-/7. )"

Test Power Level (% F.P.): O Conditions SDA: 229 SOB: moving CA:. 229 Other (specify):

{Actual) CB:. moving CC: 229 CD: 229 Below Nuclear Heating Datemtne Test Performed: ,.-...

1,/10,/,y /7.'00 Measured Parameter lsBRs; Integral Worth of Shutdown Bank B, (Description) Rod Swap

.-~-:

IV Measured Value I RS_ // 2c; (Adjusted Measured SB -

Test Critical Reference Bank Position = I~ i . steps}

Results Design Value (Actual Conditions) I RS_ {/ l. 5 (Adjusted Measured SB -

Critical Reference Bank Position = . / l 7 steps} . i Design Value (Design Conditions) lsBRS= 1129 +/- 169 pcm (Critical Reference Bank Position= 164 steps}

Reference Engineering Transmittal NAF 98-0183, Rev. 0, VEP-FRD-36A FSAR/Tech Spec If Design Tolerance is exceeded, SNSOC shall V evaluate impact of test result on safety analysis.

Acceptance SNSOC may specify that additional testing Criteria be performed.-

Reference ~EP-FRD-36A Design Tolerance is met Acceptance* Criteria is met :

. '< YES YES

-- NO NO VI Comments Prepared By: /LIJ 1tliJ Reviewed By: / ./ . ~

NE-1187 SIC 16 Startup Physics Tests Report Page 49 of 56

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RES ULTS AND EVALUATION SHEET I Test

Description:

Control Bank A Worth Measurement, Rod Swap Reference Proc No / Section: 1-NPT-RX-008 Sequence Step No:

II Bank Positions (Steps) RCS Temperature (°F): 547 Test Power Level(% F.P.): 0 Conditions SOA: 229 SOB: 229 CA: moving Other (specify):

(Design} CB: moving CC: 229 CD: 229 Below Nuclear Heating Ill Bank Positions {Steps} RCS Temperature (°F): 5'i1.Z.

Test Power Level(% F.P.): O Conditions SDA: 229 SOB: 229 CA: moving Other (specify):

(Actual} CB: moving CC: . 229 CO: 229 Below Nuclear Heating Dat~ime Test Performed:

11 i,/e;11 1,: sa Measured Parameter* I,.RS: Integral Worth of Con.troi Bank A, (Description} * . Rod Swap IV Measured Value I A

RS_

'2 ~l. (Adjusted Measured Test Critical Reference Bank Position = $' 1 steps)

Results Design Value (Actual Conditions) I A

RS_

'L 7 o (Adjusted Measured Critical Reference Bank Position = S7

• steps)

Design Value * .).

{Design Conditions} 1/s= 279 +/- 100 pcm (Critical Reference Bank Position = 52 steps)

Reference Engineering Transmittal NAF 98-0183, Rev. 0, VEP-FRD-3SA FSAR/Tech Spec If Design Tolerance is exceeded, SNSOC shall V evaluate impact of test result on safety analysis.

Acceptance SNSOC may specify that additional testing Criteria be performed. -

Reference \(EP-FRD-36A Design Tolerance is met Acceptancef Criteria *is met :

~YES YES

-- NO NO VI Comments Prepared By: {ti ]ji:// Reviewed By: 7/ ~

NE-1187 S 1C 16 Startup Physics Tests Report Page 50 of 56

{

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RES ULTS AND EVALUATION SHEET I Test

Description:

Control Bank C Worth Measurement, Rod Swap Reference Proc No / Section: 1-NPT-RX-008 Sequence Step No:

II Bank Positions (Steps) RCS Temperature (°F): 547 Test Power Level (% F.P.): O Conditions SDA: 229 SOB: 229 CA: 229 Other (specify):

(Design) CB: moving CC: moving CD: 229 Below Nuclear Heating Ill Bank Positions {Steps) RCS Temperature (°F): *SLfl.J Test Power Level(% F.P.): O Conditions SDA: 229 SOB: 229 CA: 229 Other (specify):

(Actual) CB: moving CC: moving CD: 229 Below Nuclear Heating Datemr:ne Test Performed:

,tf,,l,v- Jt,:J:i Measured Parameter leRS; Integral Worth of Control Bank C, (Description) Rod Swap IV Measured Value I RS_ 717- (Adjusted Measured C -

Test Critical Reference Bank Position = I O S"" steps)

Results Design Value I RS_

(Actual Conditions) C - 'i7 (Adjusted *Measured Critical Reference Bank Position = / o5" steps)

Design Value (Design Conditions) leRs= 725 +/- 109 pcm (Critical Reference Bank Position =98 steps)

Reference Engineering Transmittal NAF 98-0183, Rev. 0, VEP-FR0.36A FSAR/Tech Spec If Design Tolerance is exceeqed, SNSOC shall V evaluate impact of test result on safety analysis.

Acceptance SNSOC may specify that additional testing Criteria be performed. -

Reference \lEP-FRD-3SA Design Tolerance is met Acceptance Criteria is met :

~YES YES

-- NO NO VI Comments Prepared By: /&I .2::iiJ Reviewed By: ~~

NE-1187 SICI6S tartup Physics Tests Report Page 51 of 56

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP . PHYSICS TEST RESULTS AND EVALUATION SHEET .

I Test

Description:

Control Bank D Worth Measurement, Rod Swap ' *-

Reference Proc No / Section: 1-NPT-RX-008 Sequence Step No:

II Bank Positions (Steps) RCS Temperature (°F): 547 Test Power Level(% F.P.): O Conditions SDA: 229 SDB: 229

• CA:. 229 Other (specify):

(Design) CB: moving CC: 229 CD: moving Below Nuclear Heating 0

Ill Bank Positions (Steps) RCS Temperature ( F): SC/7. .s' Test Power Level {% F.P.}: O Conditions SDA: 229 SOB: 229 CA: 229 Other (specify):

(Actual) CB: moving CC: 229 CD: moving Below Nuclear Heating Datemme Test Performed:

11/1~/t'ir 15:~o Measured Parameter 10 Rs; Integral Worth of Control Bank D,

{Description) Rod Swap ,*,-~

~ * /0 ?0 /'ttA IV Measured Value

• l 0 RS= ff}'=/(... fu/'I (Adjurted Measured Test Critical Reference Bank Position = ~ t e p s )

Results Design Value Jo?l ~

(Actual Conditions) 10 RS= ~ (Adjusted Measured Critical Reference Bank Position ==-fil~teps)

Design Value

• • ~'

(Design Conditions) 10 Rs= 1076 +/- 161 pcm (Critical Reference Bank Position= 154 steps)

Reference Engineering Transmittal NAF 98-0183, Rev. 0, VEP-FRD-36A FSAR/Tech Spec If Design Tolerance is exceeded, SNSOC shall V evaluate impact of test result on safety analysis.

Acceptance SNSOC may specify that additional testing Criteria be performed. -

Reference VEP-FRD-36A

"- YES Design Tolerance is met

~ -YES

-- NO Acceptance Criteria* is met : NO VI Comments Prepared By: /W .J5i:/;{' Reviewed B y / / ~

NE- l 187 SIC 16 Startup Physics Tests Report Page 52 of 56

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET I Test

Description:

Total Rod Worth, Rod Swap Reference Proc No / Section: 1-NPT-RX-008 Sequence Step No:

II Bank Po_sitions (Steps) RCS Temperature (°F): 547 Test Power Level(% F.P.): O Conditions

• SDA: moving SOB: moving CA: moving Other (specify):

(Design) CB: moving CC: moving CD: moving Below Nuclear Heating Ill Bank Positions (Steps) RCS Temperature (°F): S't7..J Test. Power Level (% F.P.): O Conditions SDA: moving SOB: moving CA:. moving Other (specify):

(Actual) CB: moving CC: moving CD: moving Below Nuclear Heating Datemme Test Performed:

,, I,,'""

. ////l/

Measured Parameter IToea1; Integral Worth of All Banks, (Description) Rod Swap IV Measured Value 1TOlal= ss~ I pcm Test Results Design Value

{Actual Conditions) ITOlal= 5 ~ 2,(. pcm Design Value (Design Conditions) !Total= 5611 +/- 561 pcm Reference Engineering Transmittal NAF 98-0183, Rev. 0, VEP-FRD-36A FSAR/Tech Spec If Design Tolerance is exceeded, SNSOC shall V evaluate impact of test result on safety analysis.

Acceptance Additional testing must be performed.

Criteria -

Reference ~EP-FRD-36A Design Tolerance is met : ~ YES -- NO

• • Acceptance-Criteria is met : YES NO VI Comments Prepared By: dlJ} ]:;tlzk Reviewed By:

NE-1187 S 1C 16 Startup Physics Tests Report Page 53 of 56

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET I Test

Description:

M/D Flux Map- Banks D,C at Insertion Limits Reference Proc No /Section: 1-NPT-RX-008 ,002 Sequence Step No:

II Bank Positions (Steps) RCS Temperature (uF): TREF +/- 1 Test Power Level (% F.P.): ~ 30 Conditions SDA: 229 SDB: 229 CA: 229 Other (specify):

(Design) CB: 229 CC:

• CD:

• Must have ~ 38 thimbles**

Ill Bank Positions (Steps) RCS Temperature (°F): NQM '2. 7.? %

Test. Power Level (% F.P.): '27. 7?0 Conditions SDA: 229 SOB: 229 CA: 229 Other (specify);

(Actual) CB: 229 CC: 1.:ZCf CD: 1, I Datemme Test Performed:

II /i,J Cf 8 /!,~ IZ.

Maximum Relative Nuclear Enthalpy Total Heat .Maximum Measured Assembly Rise Hot Flux Hat Positive lncore Parameter Power%D1FF Channel Factor Channel Quadrant IV (Description) (M-P)/P FAH(N) Factor F0 (Z) Power Tilt Test Measured G.o P.-&.-.,.

• q Results. Value 7,& Pi.LI 'l /. ?23 '2.. /4 2 /. 0/04 Design Value +/-10% for Pi c!:0.9 (Design +/-15% for Pi<0.9 NIA N/A s1.0206 Conditions) (Pi = assy power)

Reference WCAP-7905, Rev.1 None None WCAP-7905,Rev.1

I NE-1177, Rev. 0 NE-1177,Rev. 0 V FSAR/COLR None FAH(N)

S1 .56(1 +0.3(1-P)) Fo(Z):S4.64"K(Z) None Acceptance Criteria Reference None COLR2.4 COLR2.3 None

/

Design Tolerance is met *.*.* ./ YES

-- NO Acceptance Criteria is met : 7YES NO VI

• As required Comments Al} I) /J/J Prepared By: Vff/;1)/fjl ... Reviewed By:<?~ / ) ~

I , - *--*-*-- -

NE-1187 S1CI6 S V "l.

.. . tarttif Physics Tests Report P?ge 54 of 56

SURRY POWER STATION UNIT 1 CYCLE 16-STARTUP PHYSICS TEST RES ULTS AND EVALUATJON SHEET I Test

Description:

M/0 Flux Map - At Power .

• Reference Proc No/ Section: 1.:NPT-RX-008 ,002 Sequence Step No:

II Bank Positions (Steps) RCS Temperature r'F): TREF +/- 1

. Test Power Le~el (% F.P.): 65 s P s 75 Conditions SDA: 229 SDB: 229 Ck. 229 Other (specify):

(Design) CB: 229 CC: 229 CD:

• Must have ~ 38 thimbles**

Ill Bank Positions (Steps) RCS Temperature (°F): µor,, G,,r,y7.s Test Power Level (% F.P.): bf'. t. 1 90 Conditions SDA: 229 SOB: 229 Ck. 229 Other (specify):

(Actual) CB: 229 CC: 229 CD: /1.l Datemme Test Performed:

11/u./~11 O.l :oo Maximum Relative Nuclear Enthalpy Total Heat Maximum Measured Assembly Rise Hot Flux Hot Positive lncore Parameter Power%D1FF Channel Factor Channel Quadrant IV (Description) (M-P)/P FAH(N) Factor F0 (Z) Power Tilt Test Measured s. ~ P.* ~.' J .Lfr1-,.. /. i'l./j /. oo,~

Results Value <. *' P.* < . CJ Design Value +/-10% for P1 ~0.9 (Design +/-15% for P1<0.9 NIA N/A ~ 1.0204 Conditions) (P1= assy power)

Reference WCAP-7905, Rev.1. None None WCAP-7905,Rev.1 NE-1177, Rev. 0 NE-1177,Rev. 0 V FSAR/COLR None FAH(N)~t.56(1 +0.3(1-P)) F0 (Z).:$2.32/P"K(Z) None Acceptance Criteria Reference None COLR2.4 COLR2.3 None*

• Design Tolerance is met *,*-,

-~ YES

. -- NO Acceptance Criteria is met : ~ YES NO VI

• As required .*1 Comments ** Must have at least 16 thimbles for quarter core maps for multi-point calibrations Prepared By: {lJ) ~ ** *

• Reviewed By: 4.-,..Q;__~ "=--_.

NE-1187 S1C16-Startup Physic~ Te~ts Repor_t Page 55 of 56

SURRY POWER STATION UNIT 1 CYCLE 16 STARTUP PHYSICS TEST RESULTS AND EVALUATION SHEET I Test

Description:

MID Flux Map -At Power Reference Prac No / Section: 1-NPT-RX-008 ,002 Sequence Step No:

II Bank Positions (Steps) RCS Temperature (uF): TREF +/- 1 *-

Test Power Level(% F.P.): 95 ~ P ~ 100 Conditions SDA: 229 SDB: 229 CA: 229 Other (specify):

(Design) CB: 229 CC: 229 CD:

• Must have ~ 38 thimbles..

Ill Bank Positions (Steps) RCS Temperature (°F): ,:Jr)M~-

Test Power Level(% F.P.): ti£;.st, Concfitions SDA: 229 SOB: 229 CA: 229 Other (specify):

(Actual) CB: 229 CC: 229 CD: -Z,."2--1 Date/Time Test Performed:

\,-I 1,/qt; ,m~

Maximum Relative Nuclear Enthalpy Total Heat Maximum Measured Assembly Rise Hot Flux Hot F)osiwe lncore Parameter Power%D1FF Channel Factor Channel Quadrant IV (Description) (M-P)/P F~H(N) Factor F0 (2) Power Tilt Test Measured 'i."'"7\. A,z.f':';"01 -

Results Value 9."'l f-0( P, "u*°t l.'-(1.? l .toSZ- i.co,~

Design Value +/-10% for P1~.9 -

(Design +/-15% for P,<0.9 NIA NIA :s 1.0204 I Conditions) cP1"'" assy power)

Reference WCAP-7905, Rev. 1 None None \NCAP-7905,Rev.1 i

NE-1177. Rev. O NE-1177,Rev. 0 V FSAR/COLR None F6H(N~1.56(1"'(J.3(1-P)} Fa(Zb2.32JP"K(Z) None Acceptance Criteria Reference None COLR2.4 COLR 2.3 None Design Tolerance is met Acceotance Criteria is met :

L'lES 2vEs - NO NO VI

• As required Comments .. Must have at least 16 thimbles for quarter core maps for multi-point calibrations I fl i u1()0,,rr,,, ?IY1, Prepared By:

• - v- -

i Reviewed By:

NE-1187 S 1C 16 Startup Physics Tests Report Page 56 of 56

VIRGINIA ELECTRIC AND PowER COMPANY RICHMOND, VIRGINIA 23261 e

February 12, 1999 United States Nuclear Regulatory Commission Serial No.99-029 Attention: Document Control Desk NL&OS/GDM RO Washington, D.C. 20555 Docket Nos. 50-280, 281 50-339 License Nos. DPR-32, 37 NPF-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 NORTH ANNA POWER STATION UNIT 2 ASME SECTION XI RELIEF REQUESTS North Anna Power Station Unit 2 is presently in the second ten year inservice inspection interval, and examinations are conducted to the requirements of the 1986 Edition of ASME Section XI. Surry Power Station Units 1 and 2 are presently in the third ten year inservice inspection interval, and examinations are conducted to the requirements of the 1989 Edition of ASME Section XI. . Pursuant to: 10 CFR 50.55a(a)(3)(ii), relief is requested from certain requirements of the ASME Section XI Code associated with Code required examinations.

The Code requirements for the Code Editions referenced. above require system hydrostatic testing and associated VT-2 visual examination. of all Class 1 pressure retaining piping and valves. However, small diameter (~ 1 inch), Class 1, reactor coolant system (RCS) pressure boundary vent .and drain, sample, and instrumentation connections are equipped with valves that provide for double isolation of the reactor coolant system (RCS) pressure boundary. These valves are maintained closed during normal operation and the piping outboard of the first isolation valve is, therefore, not normally pressurized. Therefore, relief is requested from performing. the hydrostatic testing and associated VT-2 visual examination for these small diameter lines because imposition of Code requirements would cause a burden that would not be compensated by an increase in quality and safety. The basis for the relief is provided in the attached relief requests.

Similar relief has been previously granted to the Edwin I. Hatch Nuclear Plant, Units 1 and 2, in the Safety Evaluation for relief request RR-17 provided in the NRG.letter from

(--- -ii~sr£~~\

G PDR .

~~ N)'-it'l 1,

Mr. H. N. Berkow of the NRC to Mr. H. L. Sumner, Jr. of the .Southern Nuclear Operating Company, Inc., dated September 3, 1998.

Relief request SPT-17 for North Anna Unit 2 is provided in Attachment 1. Relief request no. 13 for Surry Unit 1 and no. 7 for Surry Unit 2 are provided in Attachments 2 and 3, respectively. The relief requests have been approved by the applicable Station Nuclear Safety and Operating Committee.

If you have any questions concerning these requests, please contact us.

Very truly yours,

~~°(}

L. N. Hartz Vice President - Nuclear Engineering and Services Attachments cc: U. S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30323 Mr. R. A. Musser NRG Senior Resident Inspector Surry Power Station Mr. M. J. Morgan NRC Senior Resident Inspector North Anna Power Station Mr. R. Smith Authorized Nuclear Inspector Surry Power Station Mr. M. Grace Authorized Nuclear Inspector North Anna Power Station

e ATTACHMENT 1 ASME SECTION XI RELIEF REQUEST NO. SPT-17

• NORTH ANNA POWER STATION UNIT 2

• North Anna Power Station Unit 2 e

Second 10 Year Interval Request for Relief Number SPT-17 I. IDENTIFICATION OF COMPONENTS Approximately 30, small diameter (~ 1 inch), Class t, reactor coolant system (RCS) pressure boundary vent and drain, sample, and instrumentation connections.

II. IMPRACTICABLE CODE REQUIREMENTS Section XI, 1986 Edition, Examination Category B-P, Items 815.51 and 815.71 require system hydrostatic testing and associated VT-2 visual examination of all Class 1 pressure retaining piping and valves.

Ill. ISi BASIS FOR RELIEF REQUEST These piping segments are equipped with valves, or valve and flange, that provide for double isolation of the reactor coolant system (RCS) pressure boundary. These components are generally maintained closed during normal operation and the piping outboard of the first isolation valve is, therefore, not normally pressurized. The proposed alternative provides an acceptable level of safety and quality based on the following:

1. ASME Section XI Code, paragraph IWA-4400, provides the requirements for hydrostatic pressure testing of piping and components after repairs by welding to the pressure boundary. IWA-4700(b)(5) excludes component connections, piping, and associated valves that are 1 inch nominal pipe size and smaller from the hydrostatic test. Visual examination of these < 1 inch diameter RCS vent/drain/sampling connections once each 10-year interval is unwarranted considering that a repair weld on the same connections is exempted by the ASME XI Code.

2. The non-isolable portion of the RCS vent and drain connections will be pressurized and visually examined as required. Only the isolable portion of these small diameter vent and drain connections will not be pressurized.

3. All piping connections are typically socket-welded and the welds received a surface examination after installation. The piping and valves are nominally heavy wall (schedule 160 pipe and 1500# valve bodies). The vents, drains, and sample lines are not subject to high stresses or cyclic loads, and the design ratings are significantly greater than RCS operating or design pressure.

The Technical Specifications (TS) require RCS leakage monitoring (TS 4.4.6.2.1) during normal operation. Should any of the TS limits be exceeded, then appropriate corrective actions, which may include shutting the plant down, are required to identify the source of the leakage and restore the RCS boundary integrity.

During the 1998 North Anna Unit 1 refueling outage similar piping segments were pressurized by removing a flange and connecting a test rig. A majority of these piping

• segments are located in close proximity to the RCS* main loop piping thus requiring personnel entry into high radiation areas within the containment. The dose associated with this testing was 1.5 man-Rem .

. IV. ALTERNATE PROVISIONS As an alternative to the Code required hydrostatic test of the subject Class 1 reactor coolant system pressure boundary connections the following is proposed:

1. The RCS vent, drain, instrumentation, and sample connections will be visually examined for leakage, and any evidence of past leakage, with the isolation valves in the normally closed position each refueling outage during the ASME XI Class 1 System Leakage Test (IWB-5221 ).

2. The RCS vent, drain, instrumentation, and sample connections will also be visually examined with the isolation valves in the normally closed position during the 10-year ISi pressure test (IWB-5222 and Code Case N-498-1). This examination will be performed with the RCS at nominal operating pressure and at near operating temperature after satisfying the required 4-hour hold time .

. In addition, during modes 1 through 4 the RCS will be monitored for leakage at the following frequency pursuant to TS requirements:

1. Every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, during steady state operation, the reactor coolant system leak rate will be monitored to assure the limit of one gallon per minute unidentified leakage is maintained.

2. Every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> the containment atmosphere particulate radioactivity will be monitored.

The proposed alternative stated above will ensure that the overall level of plant quality and safety will not be compromised.

V.

IMPLEMENTATION SCHEDULE This alternative to Code requirements will be implemented upon receiving NRC approval for the remainder of the second ten-year inspection interval.

By a letter dated September 3, 1998 the NRR approved a similar relief request for the Edwin I. Hatch Plant, Units 1 and 2.

• e ATTACHMENT 2 ASME SECTION XI RELIEF REQUEST NO. 13 SURRY POWER STATION UNIT 1

• Surry Power Station Unit 1 Third Year Interval Request for Relief Number 13 I. IDENTIFICATION OF COMPONENTS Approximately 30, small diameter (=s; 1 inch), Class 1, reactor coolant system (RCS) pressure boundary vent and drain, sample, and instrumentation connections.

II. IMPRACTICABLE CODE REQUIREMENTS Section XI, 1989 Edition, Examination Category 8-P, Items 815.51 and 815.71 require system hydrostatic testing and associated VT-2 visual examination of all Class 1 pressure retaining piping and valves.

Ill. ISi BASIS FOR RELIEF REQUEST These piping segments are equipped with valves, or valve and flange, that provide for double isolation of the reactor coolant system (RCS) pressure boundary. These components are generally maintained closed during normal operation and the piping outboard of the first isolation valve is, therefore, not normally pressurized. The proposed alternative provides an acceptable level of safety and quality based on the

following:

1. ASME Section XI Code, paragraph IWA-4400, provides the requirements for hydrostatic pressure testing of piping and components after repairs by welding to the pressure boundary. IWA-4700(b)(5) excludes component connections, piping, and associated valves that are 1 inch nominal pipe size and smaller from the hydrostatic test. Visual examination of these < 1 inch diameter RCS vent/drain/sampling connections once each 10-year interval is unwarranted considering that a repair weld on the same connections is exempted by the ASME XI Code.

2. The non-isolable portion of the RCS vent and drain connections will be pressurized and visually examined as required. Only the isolable portion of these small diameter vent and drain connections will not be pressurized.

3. All piping connections are typically socket-welded and the welds received a surface examination after installation. The piping and valves are nominally heavy wall (schedule 160 pipe and 1500# valve bodies). The vents, drains, and sample lines are not subject to high stresses or cyclic loads, and the design ratings are significantly greater than RCS operating or design pressure.

The Technical Specifications (TS) require RCS leakage monitoring (TS Table 4.1-2A, Item No. 10) during normal operation. Should any of the TS limits be exceeded, then appropriate corrective actions, which may include shutting the plant down, are required to identify the source of the leakage and restore the RCS boundary integrity.

The required pressure testing was recently performed during their 1998 .refueling at North Anna Unit 1. Similar piping segments were pressurized by removing a flange and connecting a test rig. A majority of these piping segments are located in close proximity to the RCS main loop piping thus requiring personnel entry into high radiation areas within the containment. The dose associated with this testing was 1.5 man-Rem.

Conditions at Surry would yield comparable exposure results, if the testing were performed.

IV. ALTERNATE PROVISIONS As an alternative to the Code required hydrostatic test of the subject Class 1 reactor coolant system pressure boundary connections, the following is proposed:

1. The RCS vent, drain, instrumentation, and sample connections will be visually examined for leakage, and any evidence of past leakage, with the isolation valves in the normally closed position each refueling outage during the ASME XI Class 1 System Leakage Test (IWB-5221).

2. The RCS vent, drain, instrumentation, and sample connections will also be visually examined with the isolation valves in the normally closed position during the 10-year ISi pressure test (IWB-5222 and Code Case N-498-1). This examination will be performed with the RCS at nominal operating pressure and at near operating temperature after satisfying the required 4-hour hold time.

In addition the RCS will be monitored for leakage at the following frequency pursuant to TS requirements:

1. The reactor coolant system leak rate will be monitored daily to assure the limit of one gallon per minute unidentified leakage is maintained.

Additionally, TS 3.1.C.1 states the following:

"Detected or suspected leakage from the Reactor Coolant System shall be investigated and evaluated. At least two means shall be available to detect reactor coolant system leakage. One of these means must depend on the detection of radionuclides in the containment."

The proposed alternative stated above will ensure that the overall level of plant quality and safety will not be compromised.

V.

IMPLEMENTATION SCHEDULE This alternative to Code requirements will be followed upon receiving NRG approval for the remainder of the third ten-year inspection interval.

By a letter dated September 3, 1998 the NRR approved a similar relief request for Edwin I. Hatch Plant, Units 1 and 2.

ATTACHMENT 3 ASME SECTION XI RELIEF REQUEST NO. 7 SURRY POWER STATION UNIT 2

• Surry Power Station Unit 2 Third Year Interval Request for Relief Number 7 I. IDENTIFICATION OF COMPONENTS Approximately 30, small diameter (:s; 1 inch), Class 1, reactor coolant system (RCS) pressure boundary vent and drain, sample, and instrumentation connections.

II. IMPRACTICABLE CODE REQUIREMENTS Section XI, 1989 Edition, Examination Category 8-P, Items B15.51 and B15.71 require system hydrostatic testing and associated VT-2 visual examination of all Class 1 pressure retaining piping and valves.

Ill. ISi BASIS FOR RELIEF REQUEST These piping segments are equipped with valves, or valve and flange, that provide for double isolation of the reactor coolant system (RCS) pressure boundary. These components are generally maintained closed during normal operation and the piping outboard of the first isolation valve is, therefore, not normally pressurized. The proposed alternative provides an acceptable level of safety and quality based on the following:

1. ASME Section XI Code, paragraph IWA-4400, provides the requirements for hydrostatic pressure testing of piping and components after repairs by welding to the pressure boundary. IWA-4700(b)(5) excludes component connections, piping, and associated valves that are 1 inch nominal pipe size and smaller from the hydrostatic test. Visual examination of these ~ 1

• inch diameter .RCS vent/drain/sampling connections once each 10-year interval is unwarranted considering that a repair weld on the same connections is exempted by the ASME XI Code.

2. The non-isolable portion of the RCS vent and drain connections will be pressurized and visually examined as required. Only the isolable portion of these small diameter vent and drain connections will not be pressurized.

3. All piping connections are typically socket-welded and the welds received a surface examination after installation. The piping and valves are nominally heavy wall (schedule 160 pipe and 1500# valve bodies). The vents, drains, and sample lines are not subject to high stresses or cyclic loads, and the design ratings are significantly greater than RCS operating or design pressure ..

The Technical Specifications (TS) require RCS leakage monitoring (TS Table 4.1-2A Item No. 10) during normal operation. Should any of the TS limits be exceeded, then appropriate corrective actions, which may include shutting the plant down, are required to identify the source of the leakage and restore the RCS boundary integrity.

The required pressure testing was recently performed during their 1998 refueling at North Anna Unit 1. Similar piping segments were pressurized by removing a flange and connecting a test rig. A majority of these piping segments are located in close proximity to the RCS main loop piping thus requiring personnel entry into high radiation areas within the containment. The dose associated with this testing was 1.5 man-Rem.

Conditions at Surry would yield comparable exposure results, if the testing were performed.

IV. ALTERNATE PROVISIONS As an alternative to the Code required hydrostatic test of the subject Class 1 reactor coolant system pressure boundary connections the following is proposed:

1. The RCS vent, drain, instrumentation, and sample connections will be visually examined for leakage, and any evidence of past leakage, with the isolation valves in the normally closed position each refueling outage during the ASME XI Class 1 System Leakage Test (IWB-5221).

2. The RCS vent, drain, instrumentation, and sample connections will also be visually examined with the isolation valves in the normally closed position during the 10-year ISi pressure test (IWB-5222 and Code Case N--498-1). This examination will be performed* with the RCS at nominal operating pressure and at near operating temperature after satisfying the required 4-hour hold time.

In addition the RCS will be monitored for leakage at the following frequency pursuant to TS requirements:

1. The reactor coolant system leak rate will be monitored daily to assure the limit of one gallon per minute unidentified leakage is maintained.

Additionally, TS 3.1.C.1 states the following:

"Detected or suspected leakage from the Reactor Coolant System shall be investigated and evaluated. At least two means shall be available to detect reactor coolant system leakage. One of these means must depend on the detection of radionuclides in the containment."

The proposed alternative stated above will ensure that the overall level of plant quality and safety will not be compromised.

V.

IMPLEMENTATION SCHEDULE This alternative to Code requirements will be followed upon receiving NRG approval for the remainder of the third ten-year inspection interval.

By a letter dated September 3, 1998 the NRR approved a similar relief request for Edwin I. Hatch Plant, Units 1