ML20126H849

Submits Comments on 810310 Memo Re Physics Startup Test Program for Reload Pwrs.Concurs That Review of Concept Should Be Conducted

ML20126H849
Person / Time
Site:	Calvert Cliffs
Issue date:	04/07/1981
From:	Novak T Office of Nuclear Reactor Regulation
To:	Rubenstein L Office of Nuclear Reactor Regulation
References
NUDOCS 8104220155
Download: ML20126H849 (46)

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E' og\ . UNITED STATES E . - if , t NUCLEAR REGULATORY COMMISSION

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T';;.A'; CUM FOR: L. Rubenstein Assistant Director for S*l D ""APkl19b 3 C 2 l g,D 'arb Core and Containment Systems, DSI h./ %3 i

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y FF: T. Novak, Assistant Director for c5 [

Operating Reactors, DL vi  !

S'.3C ECT: PHYSICS STARTUP TEST PROGRAM FOR RELOAD PWRS

(:; e:uested that.I make coments on the Merch 10, 1981 Memorandub,from  !

, Cna:terton through D. Fieno to W. Johnston on the stated subject. I i a; se that much work has been Jone on physics startup tests for PWRs and  !

rat a revie,e of the concept should be made at this time. However, my l

c'usi:n is considerably different than that presented in Ms. Chatterton's  !

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- ;c:: pro; ram should receive attention at each of the following points: l

'. . se cf standards to require'the necessary startup tests to prove the eload core is loaded correctly and is in agreement with the calculated  ;

h.cs'cs ;arameters; j 2, e: ailed revie. of the licensee's physics startup test program procedures i

'r:1. ding the acceptance criteria; (

i I.  :.-site ecservaticn of the startup testing a's necessary to insure  ;

r:cedures are followed; and i l

. :ecie<! cf the results of the physics startup testing program, j

~~ x. i ease ::e chat the above points include the fcur items Ms. Chatter:cn  ;

e'1 eses all '!censees should submit information on for each reload (even  !

1. :r.:se ;erferred under 10 CFR 50.59). This r.ethcd seens very r'e;stative i

a es ecora far your staff when it is not necessary. Let me dis:uss l

: :elieve eacn of the above points should be handled, t

:' - ^ : Tne Technical $;ecifications (~S) "r all ::erating facilities  !

:: ;reserti; soecify the necessary physics startup tests. "f tne CF3 j s is net true, then a generi: issue exists tha re;uires

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i:: ' i:n in ;r: normal licensing manner; i.e. generic letter rec.esting .

: a ;as, staff review and issuance of license amendments.

"r- 1: The review of all o:erating preced res has, in the past, been the f 5:::rii:ility of IE. Enclosure 1 is the pertinent pages of the March 13,  !

• ili ks:ection re ort for the Calver: Cliffs units. Note that the Region I - l

1:e: 'rs;ec:or s;ecializing in procedure reviews spent around 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> j

: c'e ent tri;s to Calvert Cliffs) reviewing sections of the licensee's '

i' n: -'es' tin; P r: gram.  !

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81042201554  ;

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l Point 3: Enclosure 2 provides an index of IE inspection modules and some of tne inaividual modules inspectors perform for each reactor startup following a core reload. Note the detail of the IE procedures and the reliance en acceptance criteria. Also note the Enclosure 3, IE Procedure No. 72700 requirement for an inspector to observe at least five of the eight specific test to be performed.

Point 4: As indicated in the reference memorandum, unofficial (not required oy 15 or Regulations) reporting requirements have been put on the licensees.

Th s is an unacceptable practice that should be discontinued. TS for all operating plants require reporting of reactivity anomalies and errors dis-covered in the transient or accident analyses. In addition the record retention TS requires retention of records of reactor tests and experiments for at least five years. Thus, it would be a si~ ole matter for an inspector to review the data during or shortly after the performance of startup tests.

At this time when the staff workload is beyond our capabilities I recommenc that since nuch of the review suggested by the reference memorandum is controlled by TS and presently reviewed by IE inspectors, this entire review area by turned over to IE. I suggest this recommendation be presented at the next URR/IE interface meeting.

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Thomas M. No'vak, Assistant Director for Operating Reactors Division of Licensing

-nciosures: As stated s

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• EUCLOSURE 1  !

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> Eal-irore Gas and Electric Company AT-';: Mr. A. E. Lundvall, Jr.

" ice President, Supply

?. 3. Sex 1475 Ealtimore, Maryland 21203 3entlemen: .

Su b,'e::: Inspection Number 50-317/81-03 and 50-318/81-03

'nis refers to the routine safety inspection conducted by Mr. W. M. Treskeski of nis office on January 25-30, 1981 at Calver: Cliffs Nuclear Pcwer Plant Units 1 anc 2 Lusby, Maryland of activities authori:ed by NRC License N s. 0?R-53 and

?E-59 and Oc the discussions of cur findings held by Mr. W. M. Tresk ski with "r. L. 5. Russell of your staff at the conclusion of the inspection.

Areas examined durina this ins:ection are described in the Office of Inscection an: Enf:rce..ent Inspectien' Report which is enciesed with this letter. Witnin

nese areas, the ins:ection consisted of selective examinations of :recedures t r.: representa-ive records, interviews with personnel, and observaticns by the ns:e:::r, insac :n the results of this inspection, it apdears that ene of your ac-ivities Es cet ::nducted in full c: pliance with "RC requirements, as se- fer:n in the

':-ice f Viciatien, enciesed herewith as Appendix A. This itam cf nonc:::liance as bien :ste;:ri:ed into the levels described in the Federal ?.ecister Notice 45 R 55752) dated Cet:bar 7,1950. Y0u a.re recuirec to rts; nd to -his ie :er in: in :re;aring ycur response, you shculd fellcw the instructions in ;:per. dix

'n a:::rdance with Se:ti:n 2.790 of the "RC's "?.ules of Practice," Par- 2, Title

'C, C:ce Of Federai Pe;ulaticns, a copy of this le:.er and the enciesures will

s :. aced in -he NRC's :ublic Cocumen: Reen. If this reper contains any inf:rr.ati:n I

.a y:; (Or y ur c:ntra:: r) believe t: be er :rietary, it is necessary that

. :' . ake a ari-tan a::lication w thin 20 ' days to tnis : ice to withnoid sucn i

.': .1:icn fr:m public cisclosure. Any su n a; lica:icn us be ac::r:anied by tr affi:ay1: execu ed by the owner of :ne infor .ation, wnich icentifies the

u. er.: or ;ar: sou;nt to be withheld, and which c:ntains a statemen: cf rats:ns

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" :n 1: dresses wi-h specificity the i ams ahich will be censidered by the (

'.:--ission as listed in subparacra:n (.b) (c'j of Section 2.790. The informatien  !

;"t to be withheld shall be incor: crated as far as possible into a separate

1-: of the affidavit. If we do ne: hear fr = y:u in :nis recard within the l

e:ified peri d, the re:crt will be : laced in tne Public Occument Roc .

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U.S. NUCLEAR RE3ULATORY COMMISSION e%,,s,>i .;"-

CFFICE OF INSPECTION AND ENFORCEMENT V '."*..

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Region I . .

'h4 Report No. 50-317/81-03 50-318/81-03 Docket No. 50-317 50-318 License No. DPR-53 ,

OPR-69 Priority Category C Licensee: Saltinore Gas and Electric Con:any P.O. Box 1475 Baltimore :taryland 21203 Facility t!anc; Calvert Cliffs Units 1 and 2 ,

Inspection at: Lusby, Maryland Insceciicn c:nducted: January 25-30, 1981 Ins:ee: ors: .d ' ,' . - -/h - 3 //c/?/ i W. Tros(es(1, Reactor Ins:ector ca:e signec

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Section No. 1, RO&ilS Eranch i

Ins:setien !;=ary:

rs:e::icn :n January 25-20,1931 (00 bine: *ns:e::icn Eerer: ':s. 50-3*.7/51-03 ar: 5 -:.:. n- w , , ,

i Areas Ins:ected: R:utine, unanncunced inspection of li:ensee acticns en ;revicus I items; fuei nandling operations, and surveillance tasting related to refueling Technical 5:ecifications for Unit 2; s artup testing anc data recuction fer Uni: t 1; IE Cir:ulars;.and, adninistrative centrols. The ins:ection involved 22 ins:e::: -

hours onsite by a re;ien-based NRC ins;ector.

?.esul ts : Of the five areas ins:ected, no iters of nenc:::liance were identified '-

f:ur cf tne areas, one i e: was fcur.d in :ne area (level 5, fai'.ure o f:liew

r::ecures, Cetaii 5;.

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. Reactivity Control Systems Through discussions with licensee rep.esentatives and review of control reem panels and controls, the inspector verified that the limiting c:nditions of operation for the refueling mode (mode 6) were met in that: ,

(1) High Pressure Safety Injection Pump No. 23 provided a flow path from the refueling water tank to tne Reactor Coolant System (TS 3.1.2 1).

(2) The refueling water tank provided a borated water source that met t volume, concentration and temperature requirements (TS 3.1.2.7).

No items of noncompliance were identified.

. Ir. verse 'Multiolication (1/M1 Ouring the fuel loading, the inspector reviewed the sections of Fuel Wandling Frecedure FW-6(Rev. 5) that dealt with neutron flux m: nit: ring.

tserva-icns of data ceing taken and 1/M ccm:utations being made were

nduc ed en 1/2S/S1 to verify procedure adheren:e. Incepencent 1/M calculations were also made by tne ins;ector as a check on the licensee's cal cul ations'. .No discrepancies were identified.

1 u: estin; - Unit 1 Ic::e Ie:-icns Of tr.e licensee's Startup Testir.; Fr: gram were reviewed :: verify nt: tre es s were perforce: in ac:crdar.ce wi n technically adepuate and I;;r:ved pr:cedures and Technical 5:ecification repuirements. Tes: data e e also reviewed to verify that the results meet acceptance. criteria,

f-dings n e #-s:sc: r reviewed P:s:-Startup Test Procedure (?STP)-2, Uni: 1, Cyc1t "

5, Ini-ial a::r:acn :: Criticality an: Lcw 70.ver Physics Testing, Rev. ..  !

t:a are ac:e::ance criteria s.ere c:m;ared fer:

.', Critical d:ren (ARD, 532 0F1 ,

'2; Iso:hermal Tem:erature C:sfficien:

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(3) CEA Group Worths (4) Critical boron (5320F; 5, 4, 3, 2,1, Full Inserted)

Each was within its defined specifications.

Technical Specification 3.1.1.1 requires that the shutdown margin shall be.

determined to be > 4.3% ao before exceeding 5% of rated . thermal power. The licensee successfully dem:nstrated this by meeting the above acceptance criteria parameters of procedure PSTP-2, that were presented in the following Baltimore Gas and Electric documents prepared by Combustion Engineering:

1. BG&E-9576-468, 10/17/80, "Calvert Cliffs Unit 1 Cycle 5 HZP

' Critical Baron Ccncentration".

2. EG&E-9676-452, "Calvert Cliffs Unit 1 Cycle 5 Licensee Submittal".

The inspect:r notes that Baltimore Gas and Electric is to submit a sumary report of plant startup and power escalation testing fc11cwing modifications of the plant due to the new core design. Fending NRC review cf this startup report, the inspector has no further questiens at this time.

5. IE Circulars IEC: 80-17, Fuel Pin Damage Due to Water Jet Frca 4 # 1e Pla e Corner, was issued July 23, ig50. This circular identiffed a . .el pin 4. lure mechanism that has acceared cnly in certain Westingneuse pWR's. Hewever, it has been distributec :: all ?WR's since there may be o-her plan; s:ecific cesigns of

ne 'as built' core :affle that could contribute to similar fuel pin failures.

F.ec:=enced actions included (1) determina:icn of core loca:icns :na: mign:

be subject to water je: icoingement upon fuel pins that could potentially te damagec by fretting, (2) examination of fuel pins -hat were discharged frem those loca:icns, or are new at those loca icns (during the next refueling outagel, and (3) take a:propriate acticns t: correct / prevent cccurrence cf this problem.

The inseec:Or discussed these :roblems with licensee recresentatives.

These re:resenta .ives stated that to cate, there has :een nc ::servec fuel I

in camage due : water in:inge ent. Selec ed fuel assemblies have been t discharged and ins;ected for :nis specific phenomena during past refuel ngs, with ne;ative resuits. The licensee further statec that tne fuel vencor, l Cembustien Engineering, had been ccntacted when the circular was issued.

The fuel vendor indicated to the licensee that fuel pin damage of tne kind acdressed by the circular had not cccurred at any of the C-E :lan s. When the inspector repuested cocumentaticn of the licensee - fuel vencer discussiens, the licensee's represent:tive stated that they wculd request a written ,

letter frem Combustion Engineering. Based on these discussions, Circular j No. 80-17 is closed.

.y ENCLOSURE 2

' Enclosure 1 to MC 2515

, Issue Date: 1/1/81 "h!PECTION .

E*CIDURE

. INSPECTION

'. iER TITLE FREOUENCY

!!701. Surveillance R Lil 02 Surveillance of Core Power Distribution Limits X*

51703 Calibration of the Local Power Range Monitoring System X**

s.704 APRM (Average Power Range Monitor) Calibration X" ,

31705 Incore/Excore Detector Calibration X f175E Core Thermal Power Evaluation X 31'37 Determination of Reactor Shutdown Margin X* i E1705 Isothermal Temperature Coefficient of Reactivity Measurement (PWR) X 51739 Power Coefficient of Reactivity (PWR) X i

f171C Control Rod Worth Measurement (PWR) X ,

5171~. Target Axial Flux Difference Calculation (W NS!S) _

X++++' i

- 5.c;1d also be ccepleted quarterly during operating cyc.ie  ;

X?:11owing initial fuel loading and all . subsequent refuelings as describec

i. Module 72700 i

--S.cald also be completed at mid point of operating cycle ,

---i .oald also be completed following detection of an inoperable control rod

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i.:;1d also be completed following power transients greater than 50% and s . art up following a unit trip -

" Eligible for reduced frequency ,

'515-El-3 1

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Surveillance of Core Power '

UIltribution Limits PFqcecure No.: 61702 Istue Date: 10/01/30 ,

SECTION I t INSPECTION OBJECTIVES  :

1. Verify that the plant is being operated within the 1;; censed power distribution limits. ,

2. Determine whether the means utilized to confirm operation within these limits have been submitted by the licensee to NRR fo;

• review.

3. Verify that changes or alterations to calculational methods are reviewed by the l'icensee for correctness.

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l Surveillance of Core Power ,

Districution Limits l Procedure No.: 61702  :

Issue Date: 10/61/60 (W-NSSS)

SECTION II-INSPECTICN REOUIREMENTS

% rolete the portions of this procedure pertaining to the NSSS for the facility

eing inspected.

.. WESTINGHOUSE NSSS

1. Determine from the licensee which data analysis code is used to process the information obtained by the movable incere instrumen-tation. Determine whether the analysis code has ,Leen submitted to and reviewed by NRR for approval.

2. From a characteristic flux map printout (preferra, sly >50% power),

verify the following:

a. That the control rod insertions, core power level and burnup at the time of the flux map were part of the irput to the code calculations.

b. That the predicted two-dimensional power distribution analytical data for all fuel sources and fluxes measurai in the thimble

. locations for each axial region in the core ,1re part of the input. -

3. Verify from the full flux map printout in item 2,1bove that all detectors independently traversed some reference i:alibration instru-ment tuce for that particular flux map. Examine the normali:ed detector data and verify that the relative set of measurec reaction rates (fission rates as seen by detector) for esca thimble location following normalization are printed out.

4. Examine the printout in item 2 above and review tre predicted versus measured reaction rates data and cotain an explan,1 tion from the

-responsible reactor engineer for any apparent ano,r,alies.

5. Hot Cnannel Factors (1 month sample)

a. Verify that the values of the applicable tecnnical specifica-

.icas hot channel factors calculated by the innalysis code and recorded in the reactor engineering logs (or equivalent) are within the prescribed limits,

b. Ascertain that the calculated values reflect applicable uncertainty and/or penalty factors,

c. Examine the printout edits fer the highest of each of the hot channel factors and verify that the licensee has accounted for ali coservec anomalies.

II-1

' Surveillance of Core Power .

Distrioution Limits ]

Procecure No.: 61702 i Issue Date: 10/01/80 (W-NSSS)

6. Axial Flux Differences

a. Ascertain from the operations log books (or equivalent) that the Axial Flux difference limits are being maintained within their applicable ranges. (1 month sample)

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b. Review a recent load change (>20%) and verify that the axial flux difference and the mechanics associated with the logging of the applicable penalty deviations are in accordance with the ,

requirements.

7, Ascertain from tre operations logs (or equivalent) that the Quadrant Power Tilt limits are being observed and no apparent anomalies exist. (1 month sample).

8. Identify primary. personnel responsible for the major steps in obtaining the results of the computer analysis coce calculations from the initial incore flux map data.

9. Assess the apparept technical competence of the site reactor engineering staff regarding the particular core analysis code being used at the facility, including capabilities and limitations of tha method. -

10. Examine the licersee's procedures for evaluating changes or altera -

tions to calculational methods. .

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Surveillance of Core Power Dist.ribution Limits Procecure No.: 61702 ,;

Issue Date: 10/01/80 (GE NSSS)

E. GENERAL ELECTIRC NSSS

1. Examine the data monitored in performance of a recent LPRM (Local Power Range Monitor) calibration and BASE distribution calculations, as well as the results of those calculations as ;;yped out by 00-1,

- "Whole Core LPRM Calibration and BASE Distributibns" on the on-demand typewriter. Investigate alarms, error and other in process messages that may be typed out during the course of the p; ogram.

2. For the printout in item 1, ascertain that the T:P (Traversing Incore Piobe) machine normalization factors were properly obtained for all machines by traversing each probe, one at a time, through the common calibration tube.

3. Verify for the item 1 00-1 printout, that TIP (T.aversing Incore Probe) data for all LPRM locations has been accepted by the computer, i

4. Verify from a recent P-1 the following:

a. Conformance with the Linear Heat Generation Rate (LHGR) limit,

b. If the CMPF (core maximum peaking factor) is above the design value Total Peaking Factor for that class of fuel ascertain that the APRM setpoints were adjusted (as required) by the appropriate amount specified by the Technical Specifications,

c. Following such an APRM gain adjustment, verify that the "APRM GAF" on the succeeding P-1 reflects such a change.

d. Exa ,ine a P-1 showing several BA'SE CRIT COCE entries and a high ,

CMPF. Examine a subsequent Pil once the Base Crit Codes have been cleared by running the necessary TIP traces and note the effect on the CMPF.

5. Examine the 00-6, " Thermal Data in a Specified Fuel Bundle," printout associated with the P-1 selected in item 4 and ascertain that the minimum critical cower ratio (MCPR) and the limiting Average Planar .

l Linear Heat Generation Rate (APLHGR) are within tneir prescribed limits.  ;

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6. Examine the adequacy of the licensee's plans for ascertaining cpera-tion within licensed limits uncer circumstancas where the process computer is unavailable.

7.- Verify, over a one month period, that each time the computer recovered from an outage, OD-15, " Computer Shutcown and Outage Recovery Monitor," was callec in.

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Surveillance of Core Power Distrioution Limits Procedure No.: 61702 Issue Date: 10/01/80 (GE-NSSS)

S. Verify by examining the records of the three most recent LPRM gain changes that an 03-1 or 00-2 was successfully run subsequent to the changes made.

9. Assess the apparent knowledgeability of the site reactor engineering staff regarding the particular core analysis code being used at the facility, including capabilities and limitations of the method.

10. Examine the licensee's procedures for evaluating changes or alterations to calculational methods.

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Surveillance of Core Power

. DTutribution Limits Frocedure ho.: 61702 Issue Date: 10/DTRU (CIH4SSS)

C. CO.BUSTION EN3INEERING NSSS >

-1. Determine from the licensee which data analysis code is used to l-process the information obtained by the incore -instrumentation.

2. From a characteristics flux map printout verify:

a. That the control rod insertions, core power level and burnup at the time of the flux maps were part of the input-to the code calculations.

b. That the predicted power distribution analytical data for all fuel sources and fluxes ~in the instrumented locations for each axial region in the core is part of the input.

3. a. For early-C-E , reactors employing four segment fixed detector strings and no moveable chambers, determino how the readings from the detector strings are intercalibrated.  !

b. For later C-E reactors employing five segment fixed cetector ,

strings coupled with a traveling detector system, ascertain that the procedure for intercalibration is being followed. ,

4 a. Ascertain from the core performance logs (or equivalent) that the axial shape index ir, being maintained within the allowable limits. (1 month sample) -

b. Ascertain that the various uncertainty fac;; ors and flux peaking augmentation factors have been included in the setting of the -

incore detector Local Pcwer Density alarms as required by the -

Technical Specifications. .

5. Hct Channel Factors (1 months sample)  ;

a. Verify that the values of the applicable Technical Specifica-tions hot channel factors calculated by the analysis code and recorded in tne reactor engineering logs (or equivalent) are within the licensed limits. j

b. Ascertain that tne calculate:: values reflect applicable i uncertainty and/or penalty factors pertinent to the license. 4

c. Examine the printout edits f;r the highest of each of the hot channel factors calculated from the item 2 flux map and verify -

that the licensee has accounted for all observed anomalies.

6. Ascertain from the operation logs (or equivalent;) that the azimuthal )

4 powc- tilt limits are being observed and that no apparent anomalies exist. (1 month sample) ]

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Surveillance of Core Power l

Distribution Limits l

Procedure No.: 61702 l Issue Date: 10/01/80 l (CE-NSSS) _

7. Identify and document the major steps and primary personnel responsibility in the overall process of obtaining the results of the computer analysis code calculations from the initial incore flux map data.

8. Assess the apparent knowledgeability of the site reactor engineering staff regarding the particular core analysis code being used at the facility, including capabilities and limitations of the method.

9. Examine the licensee's procedures for evaluating changes or alterations to calculational methods.

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Surveillance of Core Power Distrioutien Limits Procecure ho.: 61R2 Issue Date: 10/01/86 (EAW-NSSS)

., EAECOCK AND WILCOX NSSS

1. Determine from the licensee which data analysis code is used to process the information obtained by the incore instrumentation.

2. Obtain a printout of the applicable subroutine [see reference I.5.D.(4) of Section III] and verify:

a. That the control rod insertions, core power level and burnup at the time of the flux map were part of the input to the code calculations,

b. That the predicted power distribution analytical data for all fuel sources and fluxes in the instrumentec locations for each axial region in the core are part of the ir;put.

3. Verify that the incore detector calibration procedure is being followed.

4. Ascertain from tne core performance logs (or eq(ivalent) that the axial power imbalance is being maintained withir;. the licensed limits.

(1 mon.h sample)

5. Hot Channel Factors (1 month sample)

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a. Verify that the values of the applicable technical specifica-tions hot channel factors calculated by the analysis code and recordec in the reactor engineering logs (cr equivaient) are witnin the licensed limits.

b. Verify that the calculated values reflect applicable uncertainty and/or penalty factors pertinent to the license,

c. Examir.e the printout edits for the highest of each of the hot enannel factors and verify that the licensee has accounted for any apparent anomalies.

5. Verify from the operations logs (or equivalen.) that the Quacrant Power Tiit limits are being observed and that nc apprent anomalies exist. (1 month sample)

7. Identify and document the major steps and primary personnel responsibility in the overall process cf obtaining the results of the computer analysis code calculations from tht initial incore flux map data.

6. Assess the apparent knowledgeability of the site reactor engineering staff re;arding the particular core analysis cocle being used at the fa:ility, includi..; capabilities and limitations of the methoc.

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5urveillanceofCorePower i l

Distribution Limits Procedure No.: 61702 --

Issue Date: 10/01/80 (BAW-NSSS)

9. Examine the licent:ee's procedures for evaluating changes or l alterations to calculational methods.  ;

i Verify that the prrameters specified in the most current cycle reload report have' been implemented into the computer software and verified by a test case. .

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l Core Thermal Power Evaluaticn Ffocecure No.: 61706 Issue Date: 10/01/80

-SECTION I ,

INSPECTION 0 EJECTIVE G.-ify that the calculation of core thermal power is techqically correct and p: 'ar level instruments indicate reactor power within prescribed limits.

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Core Thermal Power Evaiuation Procedure No.: 61706 Issue Date: IC/01/s0 SECTION II INSPECTION REQ'J IRD'ENTS NOTE: Inspection requirements for BWRs and PWRs are provided in Part A and Part B of this section respectively.

A. SWR Inspect'on Recuirements

1. Review the licensee's core thermal power evaluation procedure for technical adequacy and review the results for a specific evaluation at >50% power.

a. Examine the " Core Performance Fvalmation" data sheet, or equivalent, and verify that correct units have been used for the various operating parameters used to ccm:ute core thermal power, and that the initial conditions required in the plcnt procedures are sc' equate and were met.

b. Where required, verify that physical properties obtained from figures and curves corresponding to specific reactor conditions have been accurately established, properly translated and recorded on the data forms.

c. Verify that test instruments utili~ zed meet tpe applicable accuracy and calibration specifications.

d. Verify the correctness of the licensee's equttion. Review the calculations and ascertain the correctness of the results.

e. Verify power level instrument settings.

2. Verify that the frequency of evaluations is as pr,iscribed by the facility's Technical Specifications. (1 month sa;nple)

3. Independently calculate a heat balance on the nuclear boiler using the licensee's procedure for manual calculations.

S. PWR Insoection Recuirements

1. Review the licensce's core thermal power evaluation procedure for technical adequacy and review the results for a specific evaluation at >50% power.

a. Examine the secondary heat balance data shee::, or equivalent and verify that correct units have been used for the various

! operating parameters used to compute core thermal power and l

that the initial conditions required in the plant procedure are stequate and were met.

II-1

l

- Core Thermal Power Evaluation Procecure No.: 61706

-Issue Date: 107DI7GO ,

b. Verify thitt physical properties obtained from figures and  !

curves co* responding to specific reactor conditions have been l accuratel,'t established and are properly translated and recorded on the da'.a forms.

c. Verify thtt'the configuration of the Steam Generator Blowdown -

System is established in the procedure, and during the data acquisition period, was as required by the plant's procedure.

d. Verify th.ht test instruments utilized meet the applicable accuracy ind calibration specifications,

e. Verify th! correctness of the licensee's equation. Review the calculati)ns and ascertain the correctness of the results.

f. Verify pover level instrument settings. .

2. Verify that the frequency of evaluations is as prescribed by the Technical Specifications. (1 month sample)

2. Irdependently talculate a secondary heat balance using the licensee's o-ocedure f or hanual calculations.

t e

)

i I:-2

0 . .

Core Thermal Power Evaluation

' Procedure No.: 61706 Issue Date: 10/01/80 SECTION I'II INSPECTION GUIDANCE A. GENERAL BWR GUIDANCE The thermal power of the reactor core is determined by a heat balance on the nuclear boiler using operating data. Under steady state condi-tions, the nuclear boiler heat output is obtained as the difference between the total heat removed from the boiler system and the total heat added in the flow streams returning to the boiler.

la. Operating data normally recorded for core performance evaluation are the Toilowing:

Reactor pressure (psig) 6 Feedsater flow (10 lbs/hr)

Control rod drive water flow (gpm)

Total steam flow (1bs/hr)

Bypass valve position (% open)

Control valve position (% open)

Feedwater temperature (*F)

Inlet temperature to recirculation pumps ( F)

Recirculation pump power (KJ) 6 Ja pump flew (10 ibs/hr)

Core delta-P (psi)

Cleanup system heat exchanger AT (*F) l Cleanup system flow (gpm)

• Condenser vacuum (in Hg) 6 Orive water flow (10 lbs/hr)

Reactor water level (inches. of water)

Gross electrical output (MWe)

III-1

Core Thermal Power Evaluation Procecure No.: 61706 Issue Date: 10/01/80 ,

• Net electrical output (L'e) h APRM (Average Power Range Monitor) readings (%)

t.. Physical properties of concern are the enthalpies of:

Feedwater Steam ,

Jet pumps (i.e., core inlet flow) l l

Cleanup system inlet / outlet

• Control red drive water

c. Accuracy requirements are normally found in the SAR or Bases of the TS. Witness calibration of process instrumentation if possible.

c. Core thermal power equals the difference between the total energy cut and total energy in.

Total energy OUT consists of the sum of the steam energy rate, the _.

cleanup system energy rate and the fixed losses energy rate. Total energy IN corsists of the sum of the feedwater energy rate, the retirculatior pump energy rate and the control rod drive flow energy rate. In synbol form, the equation is:' -

n

" core = (Q, 4 Q cu *Of1)~(OFU

• ORD + Opumps)

,:bere: O = steam energy rate s

Q = heat loss in cleanup system Qg:' miscellaneous fixed heat losses Q7g, :' feedwater energy rate Op ) :: rod drive cooling water energy rate 1

O,g-

= recirculation pum?ing power input to water i

' : I: Energy ratu is the pre:'u:t of e. ass flew rcte ti es the enthalpy o the flow strecm, (for exac;1e: lbs/h r 'I O !1bs

• STU/g7) .

s. 'verace Powee Rance Monitors are ad,'usted to agree with the results of the heat balance as required by :ne Technical Specifications (TS).

III-2 t

Core Thermal Power Evaluation Procedure No.: 61706 Issue Date: 10/01/60

3. Tne manual calculation should agree within t 5% of the computer calculated thermal power (MWe).

B. GENERAL PWR GUIDANCE Thermal power measu* ements are utilized in the checks and calibrations of the Nuclear Power fange Instrument Chcnnels. In the thermal calibration of the Nuclear Instrumentation System, the reactor power may be obtained either from the plant com;. uter's c&lorimetric progran or by a manual method of calculation. The latter is normally requi;*ed when the computer program is not working, or to double check results obtained from the computer.

l a. Typical initial conditions for a core thermal power determination are:

• The reactor is critical anc in power operation.

• At the desired power level, the plant has be!n operated for c sufficient length of time to show that stead;/ state opernting conditions have been attained.

• Feedwater flow, water levels, and all controllable temperatures and pressures shall remain, as nearly as pos sible, unchanged throughout the data acquisition period. Thi,s can be. accomplished by minimizing rod movement and changes in befon concentration.

• Steam generator blowdown may or may not be allowed by the particular procedure.

b. The physical properties normally obtained from the plant curve book are:

• The thermal expansion factor of the feedwater flow nozzle which is plotted versus the feedwater temper,ature. This parametar is a factor in the determination of the feedwater flow. l The feedwater density (equivalent to " specific weight") l Enthalpy of feedwater

• Enthalpy of steam l

• Reactor coolant pump power

c. The Steam Generator Blowdown System is designed to continuously )

J process steam generator blowdown flow that could contain radioactive contaminants in the event of a steam generator primc.ry to secondary leak. A consideration of the mass flow and enthalpy of this bicwdown flew is necessary in a thermal power evaluation if the system is in Cpiration.

III-3

I Core Thermal Power Evaluation Procecure No.: 61705 -

l Issue Date: 10/01/80 ,

c. Accuracy requirements ate normally found in the SAR or Bases of the

~3. Witness calibration of process instrumentation if possible, e .- ' heat balance across the secondary side of the steam generators is ,

the starting point to determine the core thermal power in a PWR.

This heat balance is modified by the following to obtain core thermal tower:

Letdown er'ergy loss Reactor Ccolant Pump energy input Fixed ener'gy losses (radiation) i

n symbolic f orm, the equation is:

O core = (Q s# OLD ^ OBD + 0FL)~(OFW ^ PO )

Where: Q = Steam energy rate s

Q = Letdown flow energy rate LD

• Q = Steam generator blowdown energy rate _

80 Q = Fixed heat losses (supplied by NSSS vendor FL or determined experimentally) 07,,,, = Feedwater energy rate C = Reactor coolant pump energy input p

• Not requit'ed for plants with once-through steam generators.

. Ocwer range fluclear instruments are adjusted to agree with the esults of t!:e heat balance as required by the Technical ,

5pecificatiotis (TS). (

3.  % r s,ual calculdtion should agree within 5% of the computer  !

n':L.: ned power @We), i i

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i III-4

Determination of Reactor Shutdown Marcin Procecure No.: 61707 Isgue Date: 10/01/80 SECTION I INSPECTION OBJECTIVES

1. To verify that the licensee is ensuring adequate shutdown margin throughout the operating cycle.

2. Verify that the calculation of the reactor shutdown cargin is technically correct and in accordance with the facility's Technical Specifications and procedures.

3. Verify that the SHUTDOWN MARGIN determination has betn performed at the frequency requ+: red by the plant Technical Specificatlons.

9 4

1 I-1

Determination of Reactg- l

- Sr.utdown Marain i Procecure No.: 61707 Is. sue Date: 10/01760 SECTION II INSPECTION REOUIREMENTS 4

!.37E: Inspection requirements for the Pi!R and BWR ar;i provided in part A and part 8 of this section respectively.

A. PWR Insoection Recuirements

1. Review the licensee's shutdown margin determin,ition procedure for technical adequacy.

2. For a specific shutdown margin determination, /erify that the most recent critical conditions prior to the shutdown have been accurately recorded.

3. For the above selected shutdown margin determination, verify that the Core Reactivity chan;e frota the most ;recent critical due to the following factors has been properly obtained:

a. Reactivity change due to Boron.

b. Reactivity change due to Full Length Cont;,ol Rod Banks worth changes as a result of position, bo; ration, etc.

c. Reactivity change due to Shutdown Er - Ro,:!s .

d. Reactivity change due to Part length Rods (if in use; either way, verify that their operational status is reflected in the shatdown margin allowance). .

e. Reactivity char.ge due to Temperature.

f. Reactivity change due to Power.

i

g. Reactivity change due to Xenon.

h. Reactivity change due to Samatium and other fission products. )

{

i. Reactivity change due to fuel burnup and burnable poison i depletion.

4. s Examine the total shutdown margin calculation and verify that conditions and actions prescribed by the Technical Specifications ,

are met. {

5. Verify .nat Snutdown Margin calculations have been performed at j the frequency specified in the plant's Technical Specifications.

II-1 1

r. .

^ '

Determination of Reacter

~

Shutdown Marcin Procedure No.: 6UO7 Issue Date: 10/F17s6

6. . Ascertain that changes made in boron contentration as a ceasequence of the shutdown margin calculation results are properly verified by chemical analysis. (Sample size dependent en frequency of shutdown margin determinations. 'See guidance for inspection requirement A.5.).

7. Ascertain that changes in shutdown mergin due to rod misalignment '

have been addressed as required by the Technical Specifications.

B. GE-NSSS Inspection Recuirements *

1. Review the licensee's shutdown margin procedure for technical adequitcy.

2. Examine the shutdown margin determination made at the beginning of the current operating cycle and verify that results are in agreement with Technical Specification requirener.ts.

3.' Examine the :alculations made to determine the amount of control rod withdrawal required to correspond to the specified shutdown margin.

4. Verify that licensee has reviewed all data supplied by the fuel vendor which is utilized in the Shutdown Margin determination.

~

5. Verify that a shutdoen margin determination took place after any recent incidence of a control rod's inability to insert. Ensure that conside*ation was given to the effects-of temperature, Xenon, Samarium and other fission products, burnep and poison depletion on reactivity ai approcriate.

6. Examine the license 2's analysis of a condition where the shutdown margin could not be met and evaluate the adequacy of corrective actions that were taken.

l l

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Determination of Reactor Snutcos.n Marain Proceaure tio. : 61707 Issue Date: 10/01/e0 SECTION III INSPECTION GUI3ANCE GENERALPWRCUIDAq:

1. Minimum shutdown margin as specified in the Technical Specifications is required fcr the power operating condition, the hot standby shutdown condition and the cold shutdown condition. In all analyses involving reactor trip, the single highest worth Rod Control Assembly is postulated to remain un-tripped in its full out position.

Two independent reactivity control systems are provided, namely control rods and soluble boron in the coolant. The control rod system can co.:pensate for the reactivity effects of the fuel and water temperature changes accompanying power level chan, ries ever the range from full lood to no-load. In acdition, the cnntrol i od system provides the rainimur.i shutcown margin under Condition I, (normal operation and operational transients), events and is capable of making the core subtritical rapidly enough no prevent exceeding acceptable fuel damage limits, assuming that the highest worth control rod is stuck cut upon trip.

The boron system can compensate for All reactivity changes due to xenon burnout and buildup, temperature changes from hot shutdown to cold shutdown, fuel burnup, poison cepletion, and fission product changes and will maintain the reactor in the cold shutdown condition.

Thus, backup and emergency shutdown provisions are provided by a mechanical and a chemical shim cc.rjt rol system.

2. Conditions such as boron concentration, full length rod position, part length rod position, reactor average tenporature, power level, xenon and samarium concentrations burnup and poison depletions, are parameters contributing to the overail core reactivity and conse-quantly to the determination of total reactivity change from critical conditions to shutdown. As such, a knowledge of these parameters for the most recent critical concitions preceding a shutdown is essential.

3. A calculation of the Total Reactivity Change requires analysis of I each of the contributing factors listed in iten 3 of the inspection requirements.

The basic computation performed to determine the reactivity change associated with each parameter is to multiply the reactivity l coefficient of eacn parameter times the parameter's change in going from the cost recent critical condition to the shutdown conditions.

Reactivity coefficients (or reactivity values) for the various II:-1

6 e.', g- .

Determination of Reactor Shutdown Maroin Procedure No.: 61707 -

Issue Date: 10/01/80 -

parameters inyt.,1ved can normally be obtained from curves found in the plant's technical data book. Rod worths and reactivity coefficients will vary with burnup and boron concentrations.

It should be noted that under shutdown conditions, in calculating '

the reactivity associated with each of the various parameters,

~

negative reactivities imply positive shutdown margin and positive reactivities imply negative shutdown margins. .;

~

4. If the available shutdown margin resulting from the total reactivity [

change is insufficient to meet the Technical Specification, an additional amount of negative reactivity in the form of boration of the reactor coolant system must be added. A calculation of the boron concentration needed to meet the required shutdown margin involves determining the difference between the required shutdown reargin and the available shutdown cargin. Then, the difference h2 tween this result and the differential boron worth at the specific pre-boration conditions is equivalent to the required boration.

Finally, the minimum boron concentration to satisfy the required shutdown margin would be equivalent to the addition of the actual boron concentration at the specific conditions plus the calculated boration.

Some procedures require that calculations be checked by someone other than the one initially obtaining the results. If so, this procedural requisite should be confirmed.

5. Actual shutdown cargia calculations are required by Technical Specifications for cont'itions such as:

a. Prior to in'tial operation above 5*; Rated Thermal Power after each fuel loading,

b. After detection of an inoperable control rod.

Consequently, it is possible tnat only one actual calculation was performed for sperating cycles lacking conditions of inoperaole control rods. The samble si::e to satisfy this inscettien recuirement will the e-fore depend on the plant specific operating history. It should be noted that current Stardard echnical Specifications receire that overall core reactivity balances be compared to predicted values at least on:e per 31 Effective Full Pcwer Days.  :

B. GE!!ERAL BWR GUIDANCE:

The purpose of the shutdown nargin test is to demonstrate that the reactor can be maintai'ed subcritical by the margin scecified in the Technical .

Specifications with the highest worth rod withdrawn and the core in its most reactive condition. Normally the core will be most reactive when r Moron-free with the roderatc* at celd (20 0) conditiers.

l t

l I!!-2 l

Determination of Reactor

. Snt.tacwn Marcin Prcceaure tio. : 61707

'l Issue-Date:-10/ul/60 The shutdown margin requirement influences reactor dssign and operation, i Following are some of the direct and indirect effects: t a) Ensures tne reactor can be made sweritical from all operating i conditions.

.b) Ensures that postulated reactivity transients are controllaui within. acceptable limits.

~

c) Permits rod withdrawal for maintenance during shutdown. t c) Limits reactivity of reload fuel.

e) Requires careful planning of fuel design and loading arrangement.

1. The procedures for shutdown margin tests will generally fall into

, one of three Droad categories: 1 i

a. Two Rod m2thod - The rod cricul:tec to hevo the highest w th is fully withdrawn. Either a face adjacent, or diagonally .

c.djacent rod is withdrawn to the position calculated to equal the specified shutdown margin. A variation of this method is to continue withdrawing the second rod and perhaps a third rod -

until the reactor is critical.

b. In-sequence Critical - The rod calculated to have the highest '

worth is fully withdrawn. The reactor is then taken critical using a regular rod withdrawal sequence. ,

c. Five Rod Critical - The rod calculated to have the highest  !

worth is fully withdrawn. Tag four surrounding diagonal rods ,

are withdrawn as a Lank until the reactor is critical. A variation of this method involves a symmet;ic group of rods  !

surrounding the hignest worth rod being sequentially withcrawn  !

until the reactor is critical.

The above methods are not intenced to be all inclusive and there may '

be some other variations.

2. Snutdown n.argin must be demonstrated at the begintiing of cycie (E0C). If due to burncble poisons the core reactivity exhibits an t increase with exposure (cefinec as the R-value), an additional  !

increment of shutdown margin equal to this incraase must be  ;

demonstrated at the 500.  !

3. Cold shutdown margin calculations will involve the following:

a. Location of the highest worth control rod. .

b. . 'The maximum increase in core reactivity wi;n exposure (F,-v a' ua ) .  !.

III-3  !

-e ,; y - ,_..,.~e,...m - ,, . , , - . - . . ., . . . ~ . --....4

. Determination of Reactor Shutdown Margin l Procedure tio.: 61707 )

Issue Date:-10/DI75U

c. Predicted control rod (s) position at critical or when the specified shutdown margin has been inserted. 4 i

. If rod worth curves supplied by the fuel vendor have been adjusted by the licensee, the reasonifor and validity of this adjustment  ;

should be determined.

4. -The ' data provided by the fuel vendor normally includes:

a. - New core loading pattern. . .

b. Location of highest worth control rod.

c. Rod-worth curves,

d. Increase in core reactivity with exposure (D-value).

5. Per~ Technical Specification surveillance requirerent. .

5. Under circumstances where the shutdown margin cannot be met, several items'can be checked for anomalous conditions; for example:

a. Rod drifting.

5. Fuel assembly positions.

c. Water temperature. -

d. Water cheminitry (boron carbide tubes may have ruptured),

a. Manufac;uring (make sure manufacturing record matches actual fuel).

In any event, with the shutdown margin less than the license limit, ,

soecific action on the p&rt of the licensee is prescribed by the  :

facility's Techn ' cal Specitications.  ;

[

b 1

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III-t

1 Isothermal Temperature Coefficient I of Reactivity Measurement (hlR) l I

Procedure ho.: 61708 l Issue Date: 10/01/80 SECTION I INSPECTION OBJECTIVE

.a-t'y t..at the measurement of the Isothermal Temperature Coefficient is te:re.i:aily correct and consistent with Technical Specification requirements.

e I-1

~ . . . . - . . - - - - ~ - - . . . -.

E Isothermal Toccerature Coefficient i of Reactivity Measurement (PWR) l Procedure No.: 61708 '

Issue Date: 10/01/80 SECTION II ,

INSPECTION REOUIREMENTS i 1. Examine the adequacy of the licensee's procedure for measuring the

-Isothermal Temperature Coefficient of Reactivity and review the results for the most recent measurement,

a. Verify that the prerequisites for the measurement as delineated in the procedure were met.

b. , Verify that during the measurement, precautions as may be indicated in the procedure were observed, f i

c. Verify that plant conditions during actusi measurement correspond to those plant conditions assumed in obtaining the analytical predictions, against which the actual mearurer..ents are compared.

d. Verify that the values obtained for'the Isothermai Temperature Coefficient have been correctly determined and are within the upper and lower limits used in the FSAR accident analysis and Technical Specifications.

e. Verify that the licensee has properly accounted for any obs'erved discrepancies between actual measurements and ana'lytical predictions.

2. Verify that the frequency of measurement of the Isethermal Temperature Coefficient is as prescribed by Technical Specificatfor.s.

P g

I i

I 11-1 r

__. . - . . . . _ _ _ . , _ - . ,,,.m, . . _ _ , . , ,,.,_,.s___ - -.. . , ,c. _ m

1

- o Isothermal Temoerature Coefficient of Rescrivity Measurement (n.R)

Procedure No.: 6170s Issue Date: 10/01/50 SECTION III INSPECTION GUIDANCE

• 1:nral ne kinetic characteristics of the reactor core determine the response of the core to changing plant conditions or to operator adjustments made during cr .a1 operation, as well as the core response during abnor, rial or accidental Ort.sients. These kinetic characteristics are quantified in reactivity c oef ficients. The reactivity coefficients reflect the changes in the neutron

-/ ,iciiication due to varying plant conditions such as power, moderator or

'uti te.peratures, or less significantly (in PWRs) due to a change in pressure

/c#c conditions. Reactivity coefficients change during the life of the

, e and consequently ranges of coefficients are employed in transient analysis determine the response of the plant throughout life.

.i .sothermal Temperature Coefficient of Renctivity represents the conbined if's: on core reactivity of wo discrece cc;.gonents, ncmely, the fuel e :e ature (Doppler) coefficient and the Moderator Temperature (Density)

sffi:ient. The Fuel Temperature (Doppler) Coefficient is defined as the

.1 ge in reactivity per degree change in effective fuel temperature ano is

c'.ari,y a measure of the Doppler broadening of U-238 and Pu-240 resonance 1:1:rotion peaks.

. ir:rease in fuel temperature increases the effective resonance absorption

r:3s sections of the fuel and produces a corresponding recuction in reactivity.

~

ri Fo:.erator Temperature (Density) coefficient is defined as the change in et:-ivity per degree change in the moderator temperature. A decrease in

s ator censity means less moderation which results in a negative mocarator

sf'icient. The soluble boron used in the r# actor as a means of reactivity

-t :1 also has an effect on the moderator censity coefficient, since the 3: *u:le boron poison density (boron atoms per unit volume; not ppm) as well as

..i . iter density is decreased when the coolant temperaturn rises. A decrease

.e soluble poison density (boron atoms per unit volume) introduces a

li.ive component in the moderator coefficient. Thus, if the concentration

' s:: cole poison (in ppm) is large enough, yielding a high poison density

: atoms per unit volume), the net value of the coeffii:ient may be positive,

. tre:y ;.:tentially necessitating a boron concentraticn reduction frcm the C'  ::.5-out coron end point to ensure a negative moderator temperature

sf'icient.

.. Experimentally, measurement of the Isothermal Temperature Coefficient

ensists of calculating the slope of a plot of core r,aactivity versus Tha slope of these T,yc for various control rod bank configuraticas.

trres rapresents the Isothermal T(.mperature Coeffici. ant for the specific

pntrol rod configuration. .

III-1

l i

. e Isothermal Temperature Ceefficient of Reactivity Measurement (PWR)

Procedure No.: 61708 Issue Date: 10/01/80 ,

I i

a. Typical prerectisites for this measurement are the following:

' Maintenance of Reactor Coolant System pressure and temperature within established values.

t Achievement. of reactor criticality and specific power level with a given control rod configuration.

Control of coolant temperature via secondary steam bypass to the cond'enser or steam dump to the atmosphere. ' '

Reactor Coclant Pumps are in operation.

The Pressurizer boron concentration is within the allowable limits compared to the Reactor Coolant System boron concentration.

Neutron flux and coolant temperature signals have been adequately connected for monitoring and recording,

b. Maintaining the plant operating status as specified in the Technical Specifications is of primary concern. Also, Reactor Coolant boration or dilution sFould be avoided during the performance of the test.

. c. Validation of factors such as the Moderator Temperature Coefficient calculations is obtained ~by comparison with plant measurements a.t hot zero power. It is important to clarify whether the Doppler Coefficient '

contribution has been subtracted from the Isothermal Temperature Coefficient.in order to make these comparisons meaningful. Comparisons between predicted and treasured parameters should always correscond to ,

aiv. Imn plant. cen i Lius 'a.g. , noracli:3d to consistent control rod

'a n': configuratior.: a.. baron concentration).

d. In conjunction with 1.c., verify that the slopes calculated from the plots of reactivity versus temperature have been properly determined and the resulting values are within the bounds of the accident analyses in

the FSM, (see Inspection Requirement,1.c.)

e. Any apparent discrepancies between predictions and actual measurements sheuld be sctisfactorily cccounted for by the licensee.

2. Deternination of the *.:cderator Te5crcture Coefficient is typically rcouf red by the current TecMice' 57ecifications prior to initial op?rt:f on above EC of rated thornal T. tar, after e:ch fuel locdine. ,

In cedition, because of the pete:.cici for a positive te:.perature t coefficient near the end-of-cere life, within 7 EFP0s ef ter reaching a rated thernal pcuer equilibrium boron concentration of 300 ppm, a mec.surement is normally required.

III-2 l' ,

9

-Power Coefficient _of Reactivity (PUT0 Proccoure lio: 61'/09 Issue Date: 10/01/80 .

SECTION I INSPECTION OBJECTIVE

'.;rify that the accsurement of the Pow 2r Coefficient of Rea.:tivity is technically '

c:rrc:t and consistent with Technical Specification (TS) requirements.

p

• e ig I-l l

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ll

, - - ,-.-r. m m- ,. ~ -, M =.m...

< a Power Coefficient of Reactivity (PWR)

Procecure fio.: 61709 Issue Date: 10/DTK5 SECTION II INSPECTION REOUIREMENTS

1. Examine the adequacy of the licensee's procedure for measuring the Power Coefficient of Reactivity and the Power Defect, and review the results for the most recent measurements,

a. Verify that the prerequisites and initial conditf ons for the measurements as delineated in the procedure were met.

b. Verify that during the measurements, precautions as indicated in the procedure were observed.

c. Verify that plant conditions durir.g actaal measurement correspond to those required by the procecure anc assumed in the analytical predictions.

d. If these condi t. ions c.ra different irc a procedurai re:;uir:nonts, verify that any relaxations were approved by the responsible personnel and that 75 limitations were observed its appropriate.

e. Verify that the values obtained for the power coefficient and power defect are within the acceptance criteria. Verify the correctness of tne calculations. .

f. If the difference between the measured and predie:ted values exceeds the acceptance criteria, verify that the licensea has accounted for the discrepancy. Verify adequacy of licensee's .sctions.

2. Verify that Technical Specification limits were met daring the test.

II-1

Power Coefficient of Reactivity (PWR)

Proceoure No.: 61709 Issue Date: 10/01/80 SECTION III INSPECTION GUIDANCE

~ E;, Ei!;.L

Jring power level changes where the effects of xenon can be adequately

, E: counted for, measurements are made of reactor power and the associated ta:tivity changes. From these results, the power coefficient of reactivity t.c power defect are determined. The total power coefficient is essentially le result of the combined effect of moderator temperature and fuel temperature

3r.;es as the core power level changes. It is expressed in terms of reactivity

.a..ge per percent power change. These ceasurements are performed curing

, or escalations at preselected levels or plateaus (such as 30%, 50%, 75%, and

• . %% power.)

~.. One method for calculating the differential power coefficient of

reactivity and the int

gr61 power doisct is by folloaing Lurbine load demands with the control bank, throughout the r,snge of the

,,rogrammed load changes. The main turbine is in auto,natic control

and load-changes are initiated at the turbine panel.in the control i room. The reactor is in manual control with Tava maintained coincident with Tref by the manual insertion or withdrawal 5f tne control bank.

Thermal power measurements should be performed befor,e and after load changes. From the collected dita and subsequent analysis for xenon,

-he power coefficient and integral power defect can pe determined as ,

functions of reactor power. I

! a. Typical prerequisites and initial conditions for tnese measurements are as follows: -

,i j -

operational alignment of the neutron monitoring system has been

! satisfactorily comple.ed. l j

tne reactivity " m oter is installed and ope,ational, j i

the detail re6Gur power history (power versus time) must be 4

available for the approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> priar to the start l of the measurement in order to be able to construct the xenon 4

reactivity history over the duration of the measurement.

l a reactor thermal power measurement was performed prior to load changes.

j actual rod bank configuration is as required by procedure.

l III-1

w, ~. . .- . ~ . . . _ - -.- . -. . - . . . . .-

l l.: -

l Power Coe'fficient'of Reactivity (PWR) l Procedure No.: 61709 l

, Issue Date: 10/5T/T5 control of subsystems affecting overall plant transient response are i left in automatic (e.g. , pressurizer level, steam generator level  ;

and steam dump).  :

b. Typical precautions for this measurement are: i

. i procedural restrictions on the magnitudes 'and rates of power level '

changes shot:1d be observed (e.g., typical values are on the order  !

of.1.0% per minute). .  ?

primary system makeup during any power or load change should be avoi.ded.  :

c. The power coef ficient changes with core burnup, reflecting the combined effect of moderator and fuel temperature coefficients. As a result,  !

the value of the coefficient (experimer,tal or analytical) will depend  !

on whether the transient of interest is examined at the beginning or .

end of core li fe.  :

i

c. Typically, deviations from these procedures require as a minimum, concurrence of the lead test engineer and the shift supervisor. '

e. A typical Powe Coefficient Calculation sheet would contain the

. following parapeters: I

. initial and final core thermal power average power level, F -

initial and final xenon reactivity ,

Reactor Coo" ant System boron concent ation .

e x ti t y nvo:vid in coltrol red bar.h position changes  :

1 To perform the desired calculations, these parameters would combine  ;

as follows.

i. Average Power, F = Initial Power + AP j 2'- ,

where AP = Fir.a1 Power - Initial Power ii.

Power Defect = - (Iaprods

• APxenon) i i

l

)

III-2 l i

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f :" .

Power t,cel t ic ient. c deactivity (Ph'R)

Proceaure rio.: 61709 .

Issue Date: 10/01/80 where: Iap rods = the summation of all reactivity changes associated with changing rod bank configurations from one position to another.

Ap

• O = final xencn reactivity minus initial xenon reactivity, iii. Power Coefficient * = Power Defect '

[4P J absolute value of power change. Actual laP{ = power change, AP, may be + or - depending on whether final power is > cr < than initial power. Hoever,[aP}isalways+,

• The value obtained is unique to the specific average power, F and RCS boron concentration at the time the measurements were made.

Since induced chengas in ?ouce ~ wel t.cre achimd by -- "

changes in control bank positions, it should be noted that by plotting the change in xenon corrected reactivity (caused by rod bank position change) as a function of time, together with plotting the change in power level as a fur,ction of time, one can determine the power coefficient from ratios of these two plots for corresponding time intervals, f.e.

Power Coefficient = ao/at aP/at The numerator, ap/at, would be obtained frcn1 a reactivity computer trace for a given at interval., and the denominator AP/at is obtained by determining, through calorimetric calcuiations, how tne core power varies, as a function of time, during the corresponding time interval, at, selected on the reactivity computer trace. (Iceally, points on the reactivity versus time trace used for the numerator should be beyond the initial prompt responne portion of the curve.)

Discrepancies between predictions and actual me.tsurements should be satisfactorily accountec for by the licensee.

I. ~he test procedure snould clearly identify relevant T5 relaxation (s) if any, and highlight those requirements that are pertinent to the txpected plant conf,igurations (e.g., limitations on not channel

' actors and allowable power districutions shou'a ba observed at all times).

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Control Rod Vorth Measurements (PWR)

Procedure No.: 61710 Issue Date: 10/LT55 SECTION II INSPECTION REOUIREMENTS NOTE: The specific inspection requirements that follow are applicable for test conditions involving both boron addition and boron ailution measurements of control rod worth. Differences, if any, are indi-cated in Section III of this procedure under the corresponding guidance for the specific line item requirement of this section.

1. Examine the adequacy of the licensee's procedure for measuring the differential and integral control rod worths during boron addition and/or dilution, and revies the results for the most recent measurements.

a. Verify that the prerequisites and ir,itial ccnditions for the measurements as delineated in the procedure were met,

b. Verify that during the measurements, precautions as ;ndicated in the procedure, were observed,

c. Verify that plant conditions during actual measurement correspond

. to those plant conditions required by the procedure und assumed in the analytical predictions. .

d. If these conditions are different from procedural requirements, verify that any relaxations were approved by the responsible personnel anc that IS limitations were observed as appropriate,

e. Verify that the values obtained for the control red vorths are within the acceptance criteria. Verify correctness of the calculations.

f. Verify that the Reactor Coolant System and pressurizer were sampled for boron concentration as required to determine boron worth daring control bank movement. (Typical sampling frequencies are at 15 minute intervals).

g. Verify that tne licensee has properly. accounted for tny discrepancies between actual measurements and expected results.

2. Verify that Technical Specification limits were observed during the measurements.

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1 Control Rod Worth Measurements (PWR) .

. Procecure No.: 61710 Issue Date: 10/01/80 e SECTION III INSPECTICN GUIDANCE GL EU.

T i st::ivity worth of each bank is typical?y measured with the reactor c-i .'c11 at hot zero power. Rods may be " diluted into or borated out of" the c: s w-ile their worth is recorded by a reactivity computer. The computer w- " s:h a the "INiiOUR" equation using a power range nuclear instrumentation

s. 3:sr :Pannel as input. Control rod worth measurements can also be performed w' n:u a reactivity computer by correlating the reactivity ast;ociated with a c in;s i,. boron concentration between two rod configurations and the reactivity i is id or witncrawn according to the difference in rod configurations between t i to: 5;Etes. In the case of dilution, primary grade water is injected into tt es: cr Coolant System and the reactivity insertion caused by boron dilution i ::. :ensated for by insertion of the controlling cank until the react:r is a;1i - :ritical. The reactor coolant tecoraturo end prcsscre are r.nintrin *.

c: s a : .nrougnout the test. The procedure is typically performed witn each c: : : tank as the controlling bank, thus obtaining an integra,1 reactivity w: -* 'or eacn of the control banks.

I - e :sse of boren addition, with the reactor critical at hot. zero power, b: a e: 5.ater is injected into the Reactor Coolant System and t.he negative rn:-i.it/ caused by the boron injection is compensated for by withdrawal of t6 :o :roliing bank. As for the dilution case, reactor coolar,.t temperature a: : us;te are maintained constant throughout the test. The typical range o' a 1:.ea:,le values for RCS temperature is 542-549"F, maintaining the actual ti :tra u e witnin :1 F of the selected value. For RCS pressure, the typical vz'.e s 2235 : 25 psig. Typically, integral rea:tivity worths are obtained f: 6a:, cank separately (operation without normai overlap) anc for the control bi <: . i i:ing normal oank withdrawal sequences (with overlap).

1. L ,E-izentally, rod reactivity worth curves are obtained by plotting some 5.::r:;riate form of the output of the reactivity computer versus red bank

.* ' ; .t . For differential worth curves, differential bank worth, Ap/Ah, is ; o-ted versus bank height, h. For integral rod worth curves, the e;ral bank worth, ap, is tplotted versus bank height, h. The procedure

': nast :aiculations should provide for periccic recording of parameters ,

!.:P as reactor thermal power, reactor coolant system ter;erature and  ;

--istere and coron concentration, and pressurizer boron concentration. ,

s' eration of these parameters is essential in the determination of  !

t -:rths. I1 E. Ty?ical prerequisites and initial conditions for these measurements

,are as folicws:

Operational alignment of the neutron monitoring system has been sa-isfactorily completed.

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Control Rod Worth Measurements (PWR)

Procedure No.: 61710 Issue Date: 10/01/80

• The reactivity computer, if required, is installed and operational.

Chemistry support is available to sample the reactor coolant system and pressurizer for boron concentration as required.

• The reactor is critit.a1 and stable, at the preestablished temperature and at zero power, with the neutron flux in the range established for zero power physics tests. ,

• The rod bank 1 are in their required configuration, and the rod controi switch is in its predetermined position as indicated by procedure (e.g., use of the MANUAL mode to move rods during the separate bank, no overlap measurement of rod bank worths should invalidate the results).

b. The following represent typical precautions to be observed during these measurements:

• A limitation on the maximum start-up rate allowed during the test.

Close adherence to the prescribed values of temperature and

- pressure of the reactor coolant system throughout the test.'

(Refer to last paragraph of GENERAL GUIDANCE section for ,

typical values of RCS temperature and pressure). ,

• Adherence to the neutron flux and reactor power limits established for zero power physics tests.

• Separate, ne overlac bank movement for the portion of measurement yielding independent bank reactivity worths.

Clear awareress and understanding of special TS requirements during the test (e.g. , Group Height and Rod Insertion Limits) and positive adherence to unrelaxed requirements such as hot channel factor and thermal power limitctions.

c. The t mal rod bank configuration for each of these measurements shoulc corresponc! to tM ana'ytical configurations used for the predic ud reactivity '.corth of the rod banks. For independent bank torth measurec 0";.s, it is importent that the bcnk everlao mode of p aration not be esed if the r:sults are to be velid.

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d. Typically, changes and/or deviations from the test procedure require, as a minimum, agreement of the head test engineer and the shift supervisor,

e. The calculations involved in rod worth measurements deal primarily wi',5 two parameters, namely, reactivity and rod position. By main-taining a record of the reactivities calculatea by the reactivity II'-2 .

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- ENCLOSURE 3 .

Startup Testing - Refueling

• Procedure No.: 72700 Issue Date: 1/1/g1 f

SECTION I INSPECTION OBJECTIVE ,

i.

1. Verify that testing is conducted in accordance with approved procedures.

2.- Verify that facility is being-operated in conformar,ce with NRC requirements and licensee procedures.

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Control Rod Worth Measurements (PWR)

Proceoure No.: 61710 Issue Date: 10/01/80 computer as a function of initial and final rod positions, plots of oifferential bank worth (op/Ah) versus bank position (h) and integral bank worth (IAp) versus oank position (h) may be obttined. These results are then compared to the acceptance criteria which reflect the analytical predictions.

f. Eoron concentration analysis of the recctor coolant system and pressurizer is necessary in order to determine RCS boron worth during control bank movement. Boron samples should tie marked with Results of the the time they were taken and the sample points.

analyses shou'Id bt logged.

g. Apparent discrepancies between predictions and actual measurements should be satisfactorily accounted for by the licensee.

2. 7 e test procedure should clearly highlight those TS requirements that E e pertinent to the expected plant configurations.

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Proceoure no.. 72700 Issue Date: 1/1/81 SECTION II INSPECTION REQUIREMENTS

1. Observe at least three of the following tests for BWR's or five for PWR's and verify that tney were performed in accordance with echnically adequate and approved procedures and Technical -

Specification requirements. Verify by record rev: ew tnat the remainder of the tests were conducted,

a. Boiling Water Reactors (1) . Control Rod Drive Scram Time Tests (2) SI response to Rod Movement and any rea:tivity coefficients measured (3) Core Power Distriburion Limits (Procedure 61702)

(4) Calibration of Local Power Range Monitors (Procedure 617C3)

(5) APRM Calibration (Procedure 61704)

(5) Core Thermal Power Evaluation (Proceoure 61706)

(7) Determination of Reactor Shutdown Margin (Procecure 61707)

d. Pressurized Water Reactor Prior to Criticality (1) Roc drive and rod position indication checks (2) Reactor themocouple/RTD Cross Calioration

c. PWR's After Criticality (1) Core Power Distribution Uimits (Procedyre 61702)

(2) Incore/Excore Calibration (Procedure 6:.705)

(2) Core The al Power Evaluation (Procedure 61706) '

(4) Determin: tion of Reactor Shutdown Marg;n (Procecure 61707)

(5) Isothermal Temperature Coefficient (Procedure 61705)

(6) Power Coefficient of Reactivity Measurecent (Procecure 51709 )

(7) Control Rod Worth Measurement (Procecure 61710)

(5) Tar; t Axial Flux Differen:e Calculation, W-N555 (Procedure 61711) 2.

Review the test data for all tests icentified in Item 1 ano verify tne resu'.ts meet acceptance criteria anc tnat all ceficiencies are resolved in a timely manner, l

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i Startuo Testing - Refueling Procedure No.: 72700 Issue Date: 1/1/81_ l l

SECTION III INSPECTION GUIDANCE General The time required to complete inspection effort associated with the referenced procedures f9r Items 1 and 2 will be recorded on the 766 Form with the referenced procedure number identified as the module number inspected. Inspection items which do not have a referenced procedure will also be recorded on the 766 Form with Procedure 72700 identified as the module number inspected.

1. The licensees master outage check list normally iqentifies the startup tests to be accomplished in connjection with the r efueling outage. The verification should incluce a determination tnat test procedures a.'e available for eacn test and that any changes thertto since the previous test have been reviewed and approved by Licensee lanagement.

2. Within two refuelings, all tests shall be'witnessgd.

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