ML20154L643

Insp Repts 50-327/88-35 & 50-328/88-35 on 880711-15 & 0822- 23.Violations Noted.Major Areas Inspected:Excessive post-trip Cooldown Condition & Effect on end-of-life Margin

ML20154L643
Person / Time
Site:	Sequoyah
Issue date:	09/12/1988
From:	Branch M, Jenison K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20154L630	List: ML20154L626 ML20154L643
References
50-327-88-35, 50-328-88-35, TAC-R00398, TAC-R00399, TAC-R398, TAC-R399, NUDOCS 8809260237
Download: ML20154L643 (18)

See also: IR 05000327/1988035

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t' UNITED STATES

jl _j NUCLEAR REGULATORY COMMISSION

o e REGION 11

' 101 MARIETTA ST., N.W.

...., ATLANTA. GEORGIA 30323

Reports Nos.: 50-327/88-35 and 50-328/88-35

Licensee: Tennessee Valley Authority

6N 38A Lookout Place

1101 Market Street

Chattanooga, TN 37402-2801

Docket Nos.: 50-327 and 50-328 License Nos.: OPR-77 and DPR-79

Facility Name: Sequoyah Units 1 and 2

Inspection Conducted: July 11-15 and August 22-23, 1988

Inspectors: N b

M. W. Branch, Team Leader

M Date

iL51gned'

l9er

Team Members: P. T. Burnett

E. F. Goodwin

Approved by: [ w 2 7 W I _ E Se odl2.1979

K/ M. Jenisorf,Kcting Chief Date Signdd

Pro 1ectsSectIon1

DivisionofTVAProjects

SUMMARY

Scope: announced inspection reviewed conditions surrounding

This special, identified Sequoyah Unit 2 excessive post trip cooldown

the recently

condition and its affect on end of life shutdown margin. The

inspection was conducted on site and at the corporate office to

independently assess the event and to evaluate the the licensee near

, term and long term corrective actions.

Results: The inspection determined that the licensee had failed on numerous

occasions to take effective corrective action necessary to maintain

the plant as described in the FSAR, which is the reference document

for many activities, including reload core design.

Specifically the excessive cooldown following a reactor trip had

beenidentifledin1982aspartofthestartuptestprogram.However,

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) roper and adequate corrective action was not taken at that time.

iad the licensee taken and documented effective corrective action, as

, required by 10 CFR Part 50 Appendix B, CriterionA' Vi, the current

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problem of reduced shutdown margin would not have occurred.

8809260237 880912

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Subsequent to the initial failure to take adequate corrective action,

the problem was not identified and pursued as a discrepancy during

any of the numerous post-trip reviews or in the reload licensing

interface process. The post-trip reviews did not adequately compare

the actual plant performance with the bases in the FSAR. Through the

reload licensing checklist, the licensee was specifically requested

to reconfirm for each reload cycle that the FSAR RCS temperature

value was still valid.

Additionally, the standard Westinghouse Generic Emergency Response

Guideline procedure ES-0.1 was modified during procedure development

by TVA without recognizing that the intent of the procedure was to

quickly establish stable RCS temperatures at or above the no-load

value to preserve shutdown margin. ,

The following Violations were identified:

Failure to take adequate corrective action when the excessive

cooldown discrepancy was first identified (Paragraph 4);

followed by subsequent failures to identify or take adequate

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corrective actions during the post trip review process

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(Paragraph 7.a), as well as the 10CFR 50.59 core reload analysis

(Paragraph 7.b), and the emergency procedure implementation

process (Paragraph 7.c).

The following Unresnived Item was identified:

Applicability of 10 CFR Part 21 requirements (Paragraph 8.b)

Completion of corrective actions to assure resolution of the

excessive cooldown encroachment on shutdown margin at end of life for

Unit 1 is a restart item.

Note: A list of the acronyms and abbreviations used in this report

is found in Paragraph 10.

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REPORT DETAILS

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1. Persons Contacted  !

Licensee Employees

• J. Bynum, Vice President, Nuclear Power Production  !

• G. Gault, Reactor Engineerin  :

• J. Lemons, Nuclear Engineer,g Supervisor

Nuclear fuels l

• J. Robertson, Manager, Nuclear fuels

• B. Schofield, Licensing Engineer  !

• S. Smith, Plant Manager

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Westinghouse Employees

Nancy Campbell WestinghousefuelsProjectMana l

Noel Pogorzelski, Commercial Nuclear Fuel Divis$eron Core Engineer

Other licensee employees contacted included engineers, operators, and

office personnel,

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NRC Personnel

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0 Fieno, Section Leader, Reactor Systems Branch NRR

H.Richings, Engineer,ReactorSystemsBranch,NkR

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• Attended exit interview i

2. InspectionObjectives

The Unit 2, Cycle 3 core is the first at Sequoyah to use low-leakage (

design, which has significant safety and economic benefits. However, one t

consequence of that design was that much of the excess shutdown margin (

present in the earlier cores was lost. Near the end of Cycle 3, the i

design shutdown margin values approached the minimum value required by TS.  ;

Ouring the series of reactor trips which followed the restart of Sequoyah  !

Unit 2, it was identified that the post-trip RCS cooldown exceeded the i

design average coolant temperatures presented in the FSAR and used in the l

accident analyses. The affect of these excessive cooldowns was to ~

decrease the available reactor shutdown margin from the values assumed in  !

the cycle design and safety analysir,. l

The scope of this inspection was to evaluate information surrounding the l

discovery and proposed near and long term corrective actions associated  !

with the adequacy of shutdown margin for Units 1 and 2, subsequent to

reactor trips. Thefollowingobjectiveswereestablished:

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To summarize the chronology of the events leading up to the

identification of the problem, and the near term actions taken to

address the problem (Paragraph 3)

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To compare the plant response in the area of RCS cooldown following

reactor trips to the expected response documented in the FSAR and the

PLS documents (Paragraph 4)

To independently verify adequate shutdown margin for several of the

p(Paragraph 5)revious reactor trips associated with the Unit 2 Cycle 3 core

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To review with Westinghouse personnel the assumptions and bases of

the core reload analyses relative to RCS temperature, and to ensure

that the margins used in the main steam line break analysis are

preserved for continued plant operation (Paragraph 6)

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To evaluate the adequacy of the post-trip review process for

identifying post-trip plant performance which deviates from the FSAR

(Paragraph 7.a)

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To evaluate the TVA design controls and vendor interface associated

with the core reload licensing (Paragraph 7.b)

To evaluate the ade.quacy of the 10 CFR, Part 50.59 safety evaluations

for core reloads with respect to this problem (Paragraph 7.b)

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To determine the reportability of the discovery of the shutdown

margin problem under 10 CFR Parts 50.72 and 50.73 (Paragraph 8.a)

To review the fuel service and design analysis contract to ensure

that 10 CFR, Part 21 requirements are implemented (Paragraph 8.b)

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To review the proposed near term corrective actions (compensatory

measures), which involve modification to the standard Westinghouse

emergency procedure to instruct the operator to emergency borate if

RCS temperature fails below prescribed values and to review the

Westinghouse proposed modifications to the shutdown margin procedure

to ensure that

(Paragraphs they

6 and 9.a)are properly reflected in the TVA procedure

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To discuss with TVA long term corrective action to address plant

cooldown subsequent to reactor trips (Paragraph 9.b)

3. Chronology of Events

The following is a brief chronological summary of the events and

discussions which led to the discovery and the near term corrective

actions for the shutdown margin problem. The chronology is based on

information prepared by the licensee.

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5/19/88 - Unit 2 tripped from 72 percent power and cooled to a T l

of approximately 516?F. ave

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5/20/88 - The NRC resident ins 3ector questioned the excessive  !

cooldown associated wi;h the 5/19/88 reactor trip. l

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5/23/88 - Unit 2 tri ped from 70 percent power and cooled to a T f

of approxfmately512?F. The SI-38 Shutdown Margin ave  ;

calculated following this trip indS "d that adequate i

, shutdown margin was maintained dun RCS cooldown, [

with an excess margin of 10 ppi ... )ove the '

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requirement.  ;

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6/6/88 - Unit 2 tripped from 98 percut .'t , and cooled to

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T of approximately  :?F. . . . FAC resident

! iM$ectorquestionedthe ses .3% l

6/13/88 - The licensee initially id- A ,ue of shutdown

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margin reduction for post-tt,p M ,

) 6/14/88 - The ifcensee initiated CAQR %,. ..ch75 to document the {

j shutdown margin probleu, and discussed the core design with i

Westinghouse.

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6/17/88 - Westinghouse confirmed by analysis that SOM was not i

! violated in the 6/6/88 reactor trip.

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i Westinghouse transmitted the minimum allowable RCS I

i temperature (519?F) for maintaining the required SOM i

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following a reactor trip from 70 percent reactor power  !

(Reference 88TV*-G-0049). The specified temperature was t

, based on a reevaluation of the conservatisms used in the t

i original calculations. The 70 percent aower level was  :

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j based on plans to extend Cycle 3 operat:on into January

1989, i

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{ The Reactor Fuels and Analysis Branch issued boration l

j volumes required for the maintenance of SOM following i

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reactor trips with cooling below 520 degrees F and an l

f initialpowerlevelof70 percent (L32880617901).

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was  !

6/18/88 - ES-0.1, Emergency

revised to Procedurewith

ensure com)11ance for Reactor

technicalTrip Response, ions.

specificat l

! Revision 3 incorpora';ed a recommendstion that auxiliary  ;

i feedwater flow be limited to maintain RCS temperature above  !

l 520?F or a manual boration of the RCS be initiated.  ;

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j 6/19/88 - NRC discovers and informs plant management that PLS has

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language to the effect that plant cooldowns should be l

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restricted to preserve shutdown margin. TVA then notifies j

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NRC residents that ES-01 had been changed to compensate for

cooldown problems.

6/21/88 - P0RS initiated PRO 2-88-178 for possible reporting of the

shutdown margin problem,

6/27/88 - Westinghouse transmitted allowable cooldown temperatures

following reactor trips from 80, 90, and 100 percent

reactorpower(88TV*-G-5556).  ;

6/29/88 - Westinghouse RF&A transmitted boration volumes required for

the maintenance of SOM from pretrip

90, and 100 percent (L32 880629 904) power levels of 70, 80,

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At the time of the inspection, Revision 4 to ES-0,1, providing guidance  !

for full power operation through end of cycle based on Westinghouse data l

(88TV*-G-0057), was in draft form. Emphasis was being placed on '

maintenance program activities to repair steam leaks and help mitigate the  :

excessive cooldown. Longer term options for mitigating post trip cooldown  ;

and addressing SOM requirements were being pursued.

4. Comparison of Actual Plant Cooldown Response to Oesign Response

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The inspectors evaluated the cooldown response of Unit 2 to several

reactor trips, and compared these responses to expected system design

responses contained in the FSAR. The post trip cooldown problem applied

to both units. However, for purposes of this inspection, the inspectors I

concentrated on specific data associated with Unit 2.

The Sequoyah F / states that accident analyses of the plant are based on

a no-load averar, RCS temperature of 547'F following a reactor trip.

Section 7.7, ent'tled Control Systems, and Section 15.1, entitled Accident

Analysis, Normal Operations and Operational Transients, both indicate that I

the control system are designed and groomed to maintain a post trip ,

no-lead 7 of 547 F. Specifically, Section 7.7.1 states, "The steam r

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dumpfeedNfercontrolsystemsaredesignedtopreventtheaveragecoolant

temperatore frem failing below the programmed nc-load temperature t

j following the trip to ensure adequate reactivity shutdown margin".

! The magnitude of th, excessive RCS cooldowns for Unit 2 Cycle 3 are

j demonstrated by the data below, which was provided by the licensee:

Trip # Date Burnup Power M Post Trip T ave

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48 12/29/84 45 15% Equilibrium 510.0'F

49 1/12/85 371 99.4% Transient 513.8'F i

i 50 1/14/85 385 32.4% Transient 530.0'F

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51 2/15/85 1575 99.4% Equilibrium 506.3'F

52 2/17/85 1582 30.2% Transient 534.5'F

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53 5/3/85 4401 99.5% Equilibrium 525.5'F

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Trip # Date Burnup Power Xe Post Trip T ave

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54 5/22/85 4706 99.5% Equilibrium 518.8?F

55,56 and 57 Non-Power Trips

58 5/19/88 8196 72% Transient 516.3?F

59 5/23/88 8227 70% Transient 512.4?F

60 6/6/88 8669 97.8% Equilibrium 527.1?F

61 6/8/88 8677 12.4% Transient 522.3'F

62 6/9/88 8677 19.7% Transient 511.9?F

As can be seen from the above data the alant typically cools to an average

temperature of approximately 520?F fol'owing a reactor trip. The standard

Westinghouse design for control systems indicate that the control systems

should be able to naintain RCS temperature at or near the no-load value of

547?F.

The inspector reviewed the Sequoyah design specifics to determine if the

control systems had a different control band from the standard

Westinghouse design. The PLS document provided to TVA by Westinghouse

during construction indicated that the T control system could control

temperature within a 4?F band. The PLS alW contained Precaution #7 which

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indicated that RCS temperature must be monitored and if T was not being

properly controlled following a reactor trip the operatoY'should reduce

feed water in order to preserve shutdown margin.

The licensee has maintained that the plant has always experienced the

magnitude of cooldowns typical of those discussed above. To establish a

starting point for when these excessive cooldowns began the inspector

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i requestedfollowing

response and was orovided

a trip a copy % reactor power (SU-9.4A).of

from 100 This test the initial st

was performed in May 1982. The inspector's review of this test fridicated

that a test deficiency (2-9.4A-1) was written against step 6.8.3 which

required that T a steady oct at or above no load T without manual

intervention on Y$edwater flow. The test deficiency Wdicated that the

control system could control no better than 12?F below the no-load value.

The test procedure was accepted by the PORC and annotated to the effect

that the deficiency was acceptable since there was no mandatory acceptance

criterion which it failed to meet. Additionally, the deficiency was later

reevaluated as still being acceptable based on the fact that modifications

to the main feed system to require a feed pump trip whenever a feed

isolation occurred were complete. However, no retests were performed. The

fact that the excessive cooldowns continued af te? the modifications were

complete, and that no actions were taken to address this discrepancy '

between actual plant response and the values in the FSAR, indicate that

the corrective action for the problem, 4hich was identified through

testing, was ineffective. Although 1' s identified that plant

operations were not as described in ths a no written safety evaluation

was performed as required by 10 CFR Part .69 (See paragraph 7.b). No

actions were initiated to update the FSAR as required by 10 CFR Part 50.71

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10 CFR, Part 50 Appendix "B" Criterion XVI, Corrective Action, requires

that measures be established to assure that conditions adverse to quality,

such as failures, malfunctions, deficiencies, deviations, defective

material and equipment, and nonconformances are promptly identified and

corrected. Contrary to this requirement, the evaluation and correction of

the test deficiency associated with the ability to control RCS temperature

following a reactor trip was improper and ineffective, in that, system

performance stated in the FSAR was not obtained resulting in the reduction

of safety margin associated with reactor shutdown margin. This is

identified as the first example of Violation 327,328/88-35-01.

5. Independent Calculation of Shutdown Margin

NRC inspectors independently calculated shutdown margin for several of the

previous reactor trips on Unit 2 Cycle 3 to verify that adeouate margin

had been maintained.

The NRC requested the shutdown margin data for the five trips that

occurred since initial Unit 2 restart. These trips were identified by the

licensee as trip numbers 58 thru 62. The licensee was only able to

provided data for trips 58, 59 and 60 prior to the completion of the

inspection, as they indicated data for the other trips were in the

duplication process.

Using the licensee's procedure (SI-38}, the ins

shutdown margin calculations of the licensee. pectors verified the

The inspectors'

calculations were in general agreement with the licensee results. The

inspectors then calculated the shutdown margin for trip number 60, which

occurred on June 6,1988 from 98 percent power, using the worst case RCS

temperature (the lowest RCS temperature reached during the transient). The

results of this calculation indicated that, by using the worst case

temperature and the revision of !I-38 in effect at the time of the trfp,

the TS required shutdown margin of 1.6 % delta k/k was not maintained.

This calculation indicated that the boron concentration required to

3 reserve the TS SOM requirement would be 497.7 ppm, whereas the actual

)oron concentration was 442 ppm. However, as stated in the chronology

listed earlier, the licensee had Westinghouse perform a more refined and

precise calculation of the actual shutdown margin at the lowest RCS

temperature. The results of this calculation, transmitted 6/17/88

(88*-G-0049), indicated that the actual shutdown margin was greater than

that requirsd by TS.

The licensee was requested to calculate, using the lowest RCS temperatures

reached, the shutdown margins for trips 61 and 62. Both of these

calculations indicated that the TS shutdown margin values were met.

Thus, the inspectors confirmed that adequate shutdown margin had been

maintained throughout Unit 2 cycle 3 operation.

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It should be noted that the surveillance requirements for TS 3.1.1.2 do

not require an immediate determination of shutdown margin following a

reactor tri) and in no case does the TS require the lowest RCS temt'rature

be used. Fowever, the TS requires that the shutdown margin be pr *ved 1

at all times. Additionally, the Westinghouse developed emerge,s

procedure for actions required af ter a reactor trip recuires that a

shutdown margin calculation be performed as supplementa' action. The

licensee's position is that they perform the shutdown margin calculation

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for the actual conditions that exist at the time of performance and that

this method is consistent with other utilities that were polled.

6. Discussions of Shutdown Margin Design Assumptions with Westinghouse

The inspectors reviewed with Westinghouse personnel the assumptions and i

bases for the Sequoyah reload analyses ta determine whether adequate

shutdown marg'n exists on Unit 2 through the remainder of the current

cycle.

The Westinghouse design and analysis provided for the 1.6% delta k/k

shutdown margin required by TS to be preserved at end ,f life, when the

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core was most vulnerable to the steam-line-break accident. The FOR for

Sequoyah described plant behavior post-trip as a cooldown to the r.., load

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average coolant temperature of 547* F. The Westinghouse analysis

1 accounted for instrument errors by assuming the plant was operating 2

degrees F above full-load average coolant temperature immediately prior to

the trip and cooled to 2 degrees below no-load average temperature

following the trip, but the analysis did not provide any margin for any

, actual cooldown below 547*F.

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Once Westinghouse was informed that the actual post-trip cooldowns were

excessive, to as low as 510* F in one loop, they perfermed a bounding ,

calculation. That calculation confirmed that none of '.he cycle 3 trips  !

cnd cooldowns through June 6,1988 had reduced shutdowr margin below the

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limit. However, calculations for E0C conditions, whe i the moderator

temperature coefficient is most negative, showed thtt under some i

conditions trips from 100% RTP would lead tv insufficient shutdown margin  !

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if the cooldown was to 544 degrees F or less. Even trin from 100% RTP at }

nominal conditions, equilibeium xenon and 0 bank above 200 step withdrawn, r

would lead to reduction in shutdown margin if the RCS :ooled to less than  !

, 532' F. Since the cooldown for recent trips has bee; e .<>.t 520* F, it '

l was clear that additional action was necessary.

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! Past practice by Westinghouse has been to reduce calce ' eJ control rod l

! worth at any core condition by 10% to account for obser- differences t

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between predicted and measured control rod worths at L . However,

( Westinghouse has justified, per approved topical report WCAP 9217, a  !

i reduction of the factor of conservatism to 7%. That recalculation of

l control rod worth would provide some additional shutdown margin. However,

! FSAR Table 4.3.2-3 currently shows shutdown margin to be calculated using ,

10% calculated rod worth reduction. Discussions with NRR Reactor Systems

Branch personnel revealed that they had no technical reservations about

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reducing the rod worth adjustment to 7%. However, both they and the

inspectors objected to a further proposal by Westinghouse to reduce the

adjustment down to 5% for the current cycle of Unit 2. The basis for that

proposal was that the 30L measurement of total control rod worth was 2%

greater than calculated. However, that test confirms the adequacy of the

rod-worth-calculation model at BOL, it does not guarantee that the EOC

rod-worth calculation would also be an underestiniste.

At the time of the inspection, Unit 2 was being operated at 70% RTP both

to extend core life and to reduce the cooldown encroachment on shutdown

margin. Worst case, rods at the insertion limit and transient xenon

conditions, calculations by Westinghouse showed that a post-trip

cooldown from 70% RTP to 519' F is acceptable. NRC Generic Letter 84-21

addressed the possible adverse effects of "LONG TERM LOW POWER OPERATION

OF PRESSURIZED WATER REACTORS." Westinghouse personnel stated that

specific guidance on that concern had been provided to TVA and that the

plant was currently operatinc in conformance with the guidance. The

essential element of the gu< dance is to naintain the axial power

distribution at reduced power consistent with that expected at full power.

That is accomplished by operating with control rods partially inserted at

reduced power.

7. Evaluation of Design Controls and Vendor Interface

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l Af ter the excessive cooldown was identified as a test deficiency in 1982,

j but was not adequately corrected, there were a number of additional

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opportunities to identify the problem. The inspectors reviewed activities

I related to design and safety review processes to detettaine why the

excessive cooldown and the resulting effect on shutdown margin were not

identified through other available mechanisms,

a. Post Trip Review Process

The inspectors questioned why the discrepancy between the actual RCS

temperature and the FSAR values following a reactor trip was not

identified during any of the post-trip reviews which had been

performed.

The inspectors reviewed the post trlp review procedure AI-18, File 18

in order to determine if there was a requirement to evaluate core

performance (i.e. , shutdown margin) for the lowest temperature

reached during the transient. This procedure does not specifically

require that actual core or olant performance be evaluated against

specific FSAR transient analysis requirements. It has only a simple

(yes/no) statement that 611 designated parameters were within

expected limits. Had the post trip procedure been more detailed and

required the evaluation of worst case shutdown margin or compared

actual post trip parameters with FSAR values, the fact that control

systems were incapable of satisfying FSAR requirements regarding

lowest T a

following a reactor trip and the calculated violation of

shutdewn drgin requirements would have been identified earlier. The

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licensee was requested to evaluate the procedure for calculating port-trip

shutdown margin and determine if required results can be achieved

with current procedure detail.

The surpose of the post trip review is to identify and correct

cond'tions that are not acceptable and affect plant expected response

to trip conditions. The failure of recent post trips reviews

performed subsequent to the May 19 and 23,1988 and June 6, 1988

reactor trips a well as post tri,; reviews performed prior to the

August 1985 shutdown to identify and correct the plant cooldown

condition is identifled as the second example of violation 327,

328/88-35-01 for ineffective corrective action .

j b. Vendor Interface and 50.59 Safety Evaluation for Reload Cycles

To determine why the cooldown discrepancy was not uncovered during

i the design review for the core reloads, the inspectors examined the

interface between the licensee and the vendor, and reviewed the

10 CFR 50.59 safety evaluation performed for Unit 2 Cycle 3 and Unit

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1 Cycle 4.

Although not totally within the scope of this inspection, the

i Westinghouse interface efforts supporting a core reload analysis were

briefly discussed with the Westinghouse Fuel Project Manager and a

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Westinghouse core engineer. The Westinghouse personnel presented the

following scope and timetable for the reload design interface

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process:

i FORMAL TVA/ WESTINGHOUSE COMMUNICATIONS

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RELOAD DESIGN PROCESS

MONTHS PRIOR TO STARTUP

A. INITIALIZATION PHASE

Reload Safety and Licensing Checklist 18

• Energy Requirements

Preliminary Loading Pattern 16

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Design Initialization Meeting 14

• Design Scheidule 13

8. CORE MANAGEMENT PHASE

• Loading Pattern Established 12

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HONTHS PRIOR TO STARTUP

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C. SAFETY ANALYSIS PHASE l

Reload Safety Analysis Checklist 6 l

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Reload Saf,ety Evaluation 4 ,

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D. OPERATIONS INFORMATION ,

Nuclear Design Report

and Operations Data 1 }

The reload safety and licensing chacklist was described by Westinghouse as I

the vehicle where plant specific performance parameters are transmitted l

between Westinghouse and TVA for the purpose of validation of data. The  !

following is an excerpt from the Sequoyah Unit 2 Cycle 3 Westin i

Reload Safety and Licensing Checklist, Revision 0, dated 7/21/83. ghouse

The performance characteristics of plant components or safety systems

assumed in prior safety analyses are important input to the safety e

analyses for the next cycle. Unless otherwise advised, the

performance characteristics found in the following documents will be i

assumed for the next cycle's licensing effort. The documents '

revisions must be consistent with the date of issuance of the i

completed checklist. j

(1) The original Plant Safety Analysis Report j

(2) Loss of Coolant Accident Submittals l

(3) Fuel densification submittalc (if any)

(4) Reload Safety Evaluation Reports

(5) Technical Specifications (approved or submitted)

(6) Any other special analyses such as Anticipated Transient Without

Trip (ATWT) analyses, analyses of the effect of a modified

system, etc. unless addressed in your response. Westinghouse

must assume that the only changes in core characteristics for the

reload are those found in the design of the reload core.

Section II-C Thermal Hydraulics

(a) Change in operating pressare - none

(b) Change in operation temperature - none

T in --- 547 degrees F

T ave --- 547 degrees F to 578 degrees F

Delta T --- 60.3 degrees F

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Thus, the Westinghouse reload licensing checklist specifically states that

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the FSAR values, including no-load TA will be assumed in the reload

design analysis unless Westinghouse $, notified otherwise by TVA.

Therefore, the licensee had been formally made aware of the continuing use

of the originally assumed no-load T design value for the reload

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calculations.

, The inspector evaluated the adequacy of the TVA 10 CFR Part 50.59

eva'uations for the Unit 2 Cycle 3 and Unit 1 Cycle 4 reloads to attempt

l to determine why the expected effects of excessive plant cooldowns were

not addressed in these safety evaluations. USQO 84-34 for Unit 2, dated

a 9/7/d4, and USQD 85-20 for Unit 1, d&ted 11/1/85, contained only 2 pages

i

.

and consisted of nothing more than cover sheets with signature blocks for

the Westinghouse Reload Safety Evaluation.

j 10 CFR 50.59 allows the holder of a license to make changes in the

l facility as described in the safety analysis report without prior

commission approval unless the proposed change involves a change in the

,

technical specifications incorporated in the licensee or an unreviewed

!

safety question. A proposed change shall be tieemed to invo' 'e an

-

unreviewed safety question; 1) if the probability of occurrence or the

consequences of an accident or malfunction of equipment importaat to

'

safety previously evaluated in the SAR be increased; or 2) if the

possibility for an accident or malfunction of a different type than any

, evaluated previoudy in the SAR may be created; or 3) if the margin of

j safety as defined in the basis for any technical specification is reduced,

i

In performing the 10 CFR 50.59 safety evaluations for the reload cores,

I the accuracy of the post-trip cooldown values presented in the FSAR were

assumed to be correct and were not questioned, based on the assumption

i

that tite FSAR had been kept up to date.

! The initial failure to take adequate corrective action when the excessive

cooldown was initially identified including the failure to comply with

10 CFR 50.59 at that time was addressed in paragraph 4 as violation

327,328/88-35-01. As a result of the initial failure, both Units 1 and 2

!

had been operated since licensing outside the system design described in

i the FSAR. Specifically, FSAR Section 7.7.1 required that the steam dump

l

and feed water control system be designed to prevent the average coolant

2

temperature from falling below the program no-load temperature following a

reactor trip to ensure adequate reactivity shutdown margin is preserved.

The excessive cooldown constituted a change to the operation of the

facility as described in the FSAR and should have been supported by a

written safety evaluation.

i

l

The Westinghouse analysis, based on the criteria citablished by the Reload

Safety and Licensing checklist discussed above, used as a basis a post

trigT value of two degrees less than the no-load value of 547 F (i.c.

545 F). ave Using this post trip temperature of 545'F resulted in a

calculated EOC shutdown margin of 1.61 % delta K/k compared to the TS

required value of 1.6% delta k/k for Unit 2 and 1.64 % delta k/k compared

_ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _

o .

. .

.

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12

to the TS value of 1.6 % delta k/k for Unit 1. Although the amount of

excess shutdown margin available at E00 tended to be lessened by the low

leakage design, this reduction in itself did not constitute an unreviewed

safety question per 10 CFR 50.59 as long as the required 1.6 % delta k/k

was maintained,

i Review of corresponrience from Westinghouse to TVA reaarding this issue

included a June 27, 1988 letter (88TV*-G-0056) in which Westinghouse

stated that cooldown temperatures as low as 520'F might result in future

, loading pattern restrictions, which would reduce the low-leakage

I capability, with loss of its attendant advantages. The amount of shutdown

,

margin reduction associated with a post-trip temperature change from 545'F

!

to 520 F would reduce the EOC shutdown margin value by approximately 0.7%

i

delta

allowed k/kTS

down

value toofapproximately/k.0.9

1.6% delta k  % delta k/k,recommended

Westinghouse which is below the the

limiting

cooldown as discussed in paragraph 7c.

2

The adequacy of the reload cycle 10 CFR 50.59 evaluations, which did not

1 identify as an unreviewed safet ouestion the decrease in E0C shutdown

margin below the value allowed TS is identified as the third example of

j

violation 327,328/88-35-01 for i effective corrective action,

i The inspector reviewed the following Nuclear Fuels Procedures to determine

j if, in general, adequate design and interface controls exist:

'

NFP 7.0, Control of Reload Core Design and Analysis

• NFP 7.1, Organization and Interface for Reload Design and Analysis

'

NFP 7.2, Reload Design Document Control

The inspector concluded, based on the procedures reviewed, that in

j general, acequate controls did exist to obtain a proper core reload

analysis. However, the procedures reviewed were issued in 1987 rather

than 1983 when the analysis was performed. The manager of Nuclear

Fuels indicated that similar procedures did exist at the time the core

J

reload analysis was performed. Additionally, it should be noted that even

the current procedures will be subsequently modified to reflect the

-

February 1988, organizational change that made the Nuclear Fuels Division

a part of DNE.

c. Emergency Procedure Review

Since part of the TVA corrective action was to modify the standard

l

Westhghouse owner's group emergency operation procedure to

compensate for the excessive cooldowns, the inspector conducted a

l review of the procedure and the Westinghouse guidance. Specifically,

! procedure ES-0.1 Recctor Trip Response was reviewed. The Westinghouse

i guideline contained language to the effect that RCS and secondary

1 plant stabilization at no-load conditions was part of the procedures

I major action goals. In fact, the logic tree for step 1 of the

l procedure shows actions required for temperature decreasing below the

'

no-load values as stop dumping steam followed by controlling AFW flow

!

I

i _ _ _ _ __ .__ - _ _ _

i

' '

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13

to maintain SG 1evel at the bottom of the level band and to isolate

the main steam line if necessary.

The TVA implementing procedure issued October 4,1984, did not

specify the Westinghouse course of action to preserve temperature at

or above the no-load value. The TVA procedure stated that if Tave is

decreasing in an uncontrolled manner, then verify steam dumps and

secondar

valves. y PORV closed does

This method followed

notby closing

appear to the MSIVsthe

preserve andno-load

their bypass

temperature and consaquantly the reactor shutdown margin. The

inspector reviewed the TVA stap deviation for this procedure issued

subsequent to procedure implementation and determined that the TVA

basis for this deviation was that the AFW system design includes

automatic level control valves and therefor manual control of AFW is

not necessary. This deviation does not appear to address the issue at

hand to preserve shutdown margin possibly at the expense of reducing

AFW flow to the SGs,

If, at the time of implementation of the generic guides in

Oc' Wer 1984, TVA had questioned the pui aose of the steps in the

W . inghouse procedure the excessive coulcown/ shutdown margin problem

1

may have been properly resolved at that time. This failure to

identify and correct a nonconforming condition is identified as the

fourth exam

tive action.ple of violation 327,328/88-35-01 for ineffective correc-

8. Reportability

a. 10 CFR 50.72 and 50.73

As indicated above, the NRC considers that the reduction in EOC

shutdown margin associated with the excessive slant cooldown

constituted an unreviewed safety cuestion and cou d have resulted in

the plant being in a condition tha", was outside the design basis. In

fact, the licensee's near term corrective action was to limit reactor

power to 70 percent and to change the standard Westinghouse post-trip

emergency procedure as a compensatory measure to ensure that the

plant could be operated within the design basis.

The CAQR (SQP880375) dated 6/14/88 indicated that the discovered

condition was not reportable. The copy of PRO (2-88-178) dated

6/21/88 provided to the inspectors did not have a reportability

determination made at the time of the inspection. However the

licensee did provide the written report , LER 328-88-030 withinthe

required 30 day period.

b. 10 CFR Part 21

Due to the potu ttal generic implication of the above shutdown margin

problem, the inspector reviewed the Fuels and Analysis service

contract (68p-84-TI) between TVA and Westinghouse to determine if the

requirements of 10 CFR 21 regarding vendor re:ponsibility as to

__ _ ___- - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ ______ _____________ _ _ _ _ _ _ .

'

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,

'

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14

notification were applicable. The inspector determined that the

original contract dated in 1966 was issued prior to the January 6,

1978, date specified in 10 CFR 21. However, this contract has been

amended several times since Part 21 first became applicable. None of

the contract amendments contained language that the requirements of

10 CFR Part 21 apply. The contract did however, contain language to

s

'

the effect that all NRC rules and regulations both current and

future apply. The inspector requested that the licensee evaluate the

current contract and determine if it should be amended to ,

i specifically state that Part 21 applies. This item is identified as

1 unresolved item 327,328/88-35-02 pending further NRC review with the

licensee and the NRC vendor branch.

!

] 9. Review of Corrective Measures

) a. Near Term Compensatory Heasures l

]  !

,

As previously described, the licensee's near term corrective actions i

included operating at a reduced power level of 70% RTP with power t

distribution guidance provided by Westinghouse.  !

l

In addition, current and proposed plant procedures require post-trip

'

emergency boration as a compensatory action to restore shutdown

margin rapidly if the cooldern is beyond power and burnup dependent

limits. The limits and required boration were obtained from

Westinghouse, but before they were accepted and implemented they were ,

subjected to independent review and an>>1ysis by the TVA PWR Core

Design Section of the Reactor Fuel and Analysis Department. The  !

inspectors' review of the records confirmed that TVA used independent

core performance calculations to confirm that the vendor calculations .

gave results equivalent or conservative with respect to theirs. The i

TVA methods have not been described in a topical report approved by

NRR, but were deemed acceptable for quality control purposes.

Finally, the TVA calculations were reviewed by independent reviewers

and the differences from Westinghouse results rationalized by a

reviewer familiar with Westinghouse methodology. The TVA staf f

generated curves and tables of required boration as a function of ,

burnup, power level, and cooldown using a computer program written in  :

house. That program is well-documented internally, and has been l

accepted by peer review.  :

The inssectors concluded the TVA review of both Westinghouse and

internal calculations was satisfactory in both conduct and i

documentation. The inspectors did express one concern with the l

'

procedures that have or will result from these activities. The

procedures will specify the volume of 21,000 ppm boric acid to be  !

injected. At other facilities the boric acid and primary water flow i

integrators have not shown acceptable accuracy for this purpose. The  !

calibration and reliability of the boric acid integrator was not  !

established during this inspection. The inspectors expressed this  :

I

!

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. 15

concern to the TVA staff during the inspection and to management at

the exit interview on July 14, 1988.

The inspectors discussed the licensee's calculation activities and

proposed compensatory actions and procedures with members of the NRR

Reactor Systems Branch. The NRR staff had no criticism of either the

calculations or compensatory action for Unit 2 as described to them

by the inspectors. The NRR staff did express reservations about

accepting similar compensatory action for Unit 1, which is faced with

the same problem at EOC, but has yet to restart after being refueled

.

during the current outage. That reservation was forwarded to plant

management at the exit interview. Management stated they did not

intend to restart Unit I until the basic problen of excessive

'

r

cool-down to an unanalyzed temperature had been resolved. Management

further stated they intended to ecmplete their determination of the

best method to limit post-trip cooldoen within 30 days,

b. Long Term Corrective Actions '

'

As part of this inspection the inspectors planned to discuss details

of TVA's plans to minimize RCS cooldowns following reactor trips.

The licensee indicated that they are currently invrGtigating several

i

methods to attempt to control cooldowns. They include a chance in

steam dump setting from T

possible modifications to D$ control to Steam

auto-level controls Pressure contro',

associated with and

the

AFW system. Details and schedule were not discussed.

10. List of Acronyms and Abbreviations

AI -

Administrative Instruction

AFW -

Auxiliary Feedwater

ATWT -

Anticipated Transient without Trip ,

CAQR

-

Condition Adverse to Quality Report ,

CFR -

Code of Federal Regulations '

BOC -

Beginning of Cycle >

BOL -

Beginning of Life

DNE -

Division of Nuclear Energy

EOC -

End of Cycle

ES -

Emergency Procedure

FSAR -

Final Safety Analysis Report '

LER -

Licensee E,ent Report

MSIVs - Main Steam Isolation Valves

NFP -

Nuclear Fuels Procedure

NRC -

Nuclear Regulatory Commission

NRR -

Nuclear Reactor Regulation '

PLS -

Precautions Limitations and Setpoint Document

PORC -

Plant Operations Review Committee  !

P0RS -

Plant Operations Review Staff >

PORV -

Power Operated Relief Valve

PPM -

Parts Per Million

t

- _ _ _ _ _ _ . _ _ _ _

_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

.

.-

. 16

PWR -

Pressurized Water Reactor

RCS -

Reactor Coolant System

RF&A -

Reactor Fuels and Analysis

RTP -

Rated Thermal Power

SAR -

Safety Antlysis Report

SDM -

Shutdown Margin

SG -

Steam Generator

SI -

Surveillance Instruction

SQN

-

Sequoyah

SU -

Start Up Test Procedure

T

AVE

-

Average Reactor Coolant Temperature

TS -

Technical Specifications

TVA -

Tennessee Vallay Authority

URI -

NRC Unresolved item

USQ

-

Unreviewed Safety Question

USQD

-

Unreviewed Safety Question Determination

XE Xenon

'

11. Exit Interview

The inspection scope and findings were summarized on July 14, 1988, and

again on August 3,1988, with those persons indicated in paragra3h 1

above. The inspectors described the areas inspected and discussec in

detail the inspection findings listed below. During the course o' the

inspection the' inspectors were provided numerous documents which the

licensee considered as proprietary. However, no proprietary material is ,

contained in this report. Dissenting comments were not received from the  !

licensee during the July 14, 1984 exit. However, during the August 23.

1988 reexit the licensee did comment that their position was that the

shutdown margin 3roblem was licensee identified and was not prompted oy

tne NRC questioning of the excessive cooldown discussed in this report.

Item Numbsr Description and Reference

327,328/88-35-01 Violation: Failure to take adequate corrective

action when the excessive cooldown discrepancy

was first identified (Paragraph 4); followed by ,

subsequent failures to identify or take adequate

corrective during the post trip review process

(Faragraph 7.a), as well as the 10 CFR 50.59 core

reload analysis (Paragraph 7.b) and the emergenc

procedure implementation process (Paragraph 7.c)y .

327,328/88-35-02 Unresoived Item: Determine the applicabilit

10 CFR Part 21 :equirements (Paragraph 8,b) y of

i

.

. _ . _