05000336/LER-1997-005-02, :on 970204,inservice Test Instrumentation Did Not Meet Ansi/Asme Chapter XI Requirements.Caused by Inadequate Administrative Structure for IST Program. Procedure to Administer IST Program Was Implemented

:on 970204,inservice Test Instrumentation Did Not Meet Ansi/Asme Chapter XI Requirements.Caused by Inadequate Administrative Structure for IST Program. Procedure to Administer IST Program Was Implemented

ML20136F422
Person / Time
Site:	Millstone
Issue date:	02/04/1997
From:	Joshi R NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20136F418	List: 05000336/LER-1997-005, Forwards LER 97-005-00 Which Documents Event That Occurred at Mnps,Unit 2 on 970204.Commitments Made,Listed 05000336/LER-1997-005-02, :on 970204,inservice Test Instrumentation Did Not Meet Ansi/Asme Chapter XI Requirements.Caused by Inadequate Administrative Structure for IST Program. Procedure to Administer IST Program Was Implemented
References
LER-97-005-02, LER-97-5-2, NUDOCS 9703140102
Download: ML20136F422 (4)

LER-1997-005, on 970204,inservice Test Instrumentation Did Not Meet Ansi/Asme Chapter XI Requirements.Caused by Inadequate Administrative Structure for IST Program. Procedure to Administer IST Program Was Implemented

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3361997005R02 - NRC Website

text

.

NRC FORM 366 U.s. NUCLEAR REGIJLATORY CoMMisslON APPROVED BY OMB NO. 3150-0104 EXPIRES 04/30/98

',555^d3o,'~dYdd'o"N $;*~lN3[g'I!i%

LICENSEE EVENT REPORT (LER)

"oS4 OBZM %'"?!aTa'OMM^",??Me55PM' (See reverse for required number of M%^,",",5,gg%7,jJfc"rg no*,'!,Tg*og,2oy,tc oo',

o o

"^"^omesT As amer. w Aww.oron. oc am3 digits / characters for each block)

FCCILrif RAME (1)

DOCKET NUMBER (2)

PAGE13)

Millstone Nuclear Power Station Unit 2 05000336 1OF4 TITLE (4)

Inservice Test Instrumentation does not meet ANSI /ASME Chapter XI Requirements EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

SEQU AL RE S N MONTH DAY YEAR YEAR MONTH DAY YEAR NU R

02 04 97 97

-- 005 -

00 03 06 97 OPERATING THis REPORT is SUBMITTED PURSUANT To THr. REQUIREMENTS OF 10 CrR 5: (Check one or more) (11)

MODE (9)

N 20.2201(b) 20.2203(a)(2)(v)

X So.73(a)(2H0 So.75(a)(2Hviii)

POWER 20.2203(a)(1) 20.2203(aH3Hi)

So.73(aH2Hii)

So.73(aH2)(x)

LEVEL (10) 000 20.2203taH2>0) 20.2203(aH3Hio 50.73(a)(210ii) 73.71 20.2203(aH2)hi) 20.2203(all4)

So.73(aH2Hiv)

OTHER j

20.2203(aH2)Dii)

So.36(c)(1)

So.73(a)(2)(v)

Sescify in Abstract below or in NRC Form 366A 20.2203(a)(2)0v)

So.36(cH2)

So.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER pnclude Area Code)

R. G. Joshi, MP2 Nuclear Licensing (860) 440-2080 COMPLETE ONE LINE FOR EACH COMPONENT FAILisaE DESCRIBED IN THis REPORT (13)

"^'

^

CAUSE

SYSTEM COMPONENT MANUFACTURER

CAUSE

SYSTEM COMPONENT MANUFACTURER O PRDS g

MONTH DAY YEAR SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED YEs submission X No DATE (15)

(If yes, complete EXPECTED sUBMisslON DATE).

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single. spaced typewritten lines) (16)

On February 4,1997, with the plant defueled, a self-assessment of instrumentation used in support of safety related pump inservice testing identified three deficiencies where instrument calibration requirements imposed by Section XI of the Code of the American Society of Mechanical Engineers (ASME)," Rules for Inservice Inspection of Nuclear Power Plant Components" were not met. Two deficiencies related to instruments being used outside their allowable calibration range. One deficiency related to instruments not being included in the facility calibration program.

Technical Specification 4.0.5 requires inservice testing of safety related components to be performed in accordance with Section XI of the ASME Code. Therefore, this event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B).

The cause of this event was an inadequate administrative structure for the IST program. There are no requirements to ensure ASME code requirements are met with regard to calibration, identification, and maintenance of quality related installed instruments under the current IST program. As a result, insufficient supervisory attention was applied to assure all ASME requirements were met.

Corrective action for this event includes the implementation of a procedure to administer and monitor an IST program which complies with the requirements of Section XI of the Code of the ASME, and the evaluation and correction of the identified instrumentation deficiencies.

9703140102 970306 PDR ADOCK 05000336 g

PDR

NHC FORM 366A U.s. NUCLEAR REGULATORY CoMMisslON (4 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FAClllTY NAME (1)

DOCKET LER NUMBER '6)

PAGE (3)

SEQUENTIAL REVISION YEAR NUMBER NUMBER 2OF4 i

Millstone Nuclear Power Station Unit 2 05000336 97 - 005 -

00 TEXT Ilt more space is required, use additional copies of NRC Form 366A) (17) l l.

Descriotion of Event On February 4,1997, a self-assessment of instrumentation used in support of safety related pump [P] inservice testing identified three deficiencies where instrument calibration requirements imposed by Section XI of the Code of the American Society of Mechanical Engineers (ASME)," Rules for Inservice Inspection of Nuclear Power Plant Components," were not met. Two deficiencies related to instruments being used outside their allowable calibration range. One deficiency related to irutruments not being included in the facility calibration program. At the time of discovery of this event, the unit was defueled.

Specific inservice testing instrumentation deficiencies include:

1.

Service Water [Bl] pump differential pressure is determined by subtracting the level in the intake bay (suction pressure) from the pump discharge pressure. The differential level instrument is calibrated over a range of 0 to 40 inches of water. The reference level differential during Service Water pump inservice testing is approximately 4 inches. ASME Section XI requires, in part, that the full range of each analog instrument shall not be greater than three times the reference value. The analog instrument used to measure differential level across the intake screen [SCN) did not meet this requirement.

2.

High Pressure Safety injection (HPSI) [BQ] pump inservice test (IST) flow is measured using a digital flowmeter calibrated over a range of 30 to 250 gpm. The reference flow for HPSI pump testing is 27 gpm. NUREG-1482 " Guidelines for Inservice Testing at Nuclear Power Plants," recommends the application of the rules of ANSl/ASME OM-6," Requirements for Pump Testing in Light Water Reactor Power Plants," for digitalinstrumentation. OM-6 requires digitalinstrument accuracy of +/- 2 percent of reading. This flow measurement was outside the calibrated range of the instruments.

3.

Low Pressure Safety injection (LPSI) [BP] pump IST differential pressure is determined by subtracting pump suction pressure, as determined from the plant process computer, from pump discharge pressure.

These two computer points were not included in the calibration program. ASME Section XI requires, in part, that a system of calibration records be used to identify each instrument and its date of calibration.

These instruments were last calibrated in 1990. No periodic calibration was specified.

Existing station administrative control procedures specifically exclude installed instrumentation from the requirements imposed on measuring and testing equipment (M&TE). There are no formal programmatic requirements for identification, calibration, and maintenance of quality related installed instrumentation used for inservice testing. As a result, insufficient supervisory attention was applied to assure all ASME requirements were met. These deficiencies were first identified in August 1995. Though the inservice test instrumentation was reviewed in August 1995, the review was not sufficiently detailed to identify these problems.

The program to meet the requirements of ASME Section XI, and the associated relief requests is delineated in the Millstone IST Program Manual. Pursuant to 10 CFR 50.55a(g)(6)(i), relief from ASME Code requirements and attemate testing method requests must be submitted to the NRC for approval. The IST Program Manual does not identify the above gauges as deviating from the ASME Code Section XI requirements, and a rellef request was not submitted to use the existing pressure gauges.

Technical Specification 4.0.5 specifies surveillance requirements for inservice inspection (ISI) and testing of ASME Code Class 1,2, and 3 components. This surveillance requires ISIS to be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. Failure to meet this surveillance results in the components / systems associated with the IST being inoperable and other associated LCOs applicable to these components not being met. A violation of the facility Technical Specification occurs if the instruments in question

~.~.

.U.s. NUCLEAR REGULATORY CoMMISslON (4-95) j LICENSEE EVENT REPORT (LER) i TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER i 6)

PAGE (3)

SEQUENTIAL REVISION AR NUMBER NUMBER 3OF4 Millstone Nuclear Power Station Unit 2 05000336 97

-- 005 --

00 TEXT (If more space os required, use additional copies of NRC Form 366A) (17) f are inoperable for a period longer than the applicable Technical Specification Action Statements allow.

Consequently, these conditions are being reported in accordance with 10 CFR 50.73(a)(2)(i)(B), any operation or condition prohibited by the plant's Technical Specifications.

i l

11.

Cause of Event

l i

The cause of this event was an inadequate administrative structure for the IST program. There are no requirements to ensure ASME code requirements are met with regard to calibration, identification, and l

j maintenance of quality related installed instruments under the current IST program. As a result, insufficient supervisory attention was applied to assure all ASME requirements were met.

1 j

Failure to provide for periodic calibration indicates further weakness within program to program interfaces.

Information necessary to comply with the ISI ASME Section XI requirements was not provided. This resulted in

{

inappropriate instrumentation being selected for IST measurements.

i lli. Analysis of Event inservice Testing in accordance with ASME Section XI ensures the operational readiness of pumps and valves i

which are required for shutting down the reactor to a safe shutdown condition, maintaining the safe shutdown

]

~

condition, or mitigating the consequences of an accident, j

The Technical Specification Surveillance requirements were not met for this instrumentation. However, 1.

Service Water pump operability was not affected since the possible error in measurement of intake screen differential level is less than 1/10 the allowable error of the differential pressure measurement.

1 i

Thus, the overall accuracy of the measurement used to verify Service Water pump operability was not significantly affected.

1 3

2.

The use of the digital instrument below the calibrated range is not significant since the measured flow is j

within 10 percent of the lowest calibrated flow (30 gpm). The vendor's specification for this flow 1

instrument indicates an expected linearity of.003 ft/sec (equivalent to approximately 0.1 gpm or 0.4 l

percent accuracy). The instrument error at 30 gpm was less than i percent of the reading. Even i

considering a maximum linearity error, it is likely that the error at 27 gpm was within the ASME requirements. Further, NUREG-1482 permits routine testing of these pumps without flow measurement if i

necessary. Thus, the operability of the HPSI pumps was not jeopardized.

l

3. LPSI pump operability was not affected since, though not part of the calibration program, the LPSI pump suction pressure computer points remained in calibration from 1990 to the present and the accuracy of the measurements used to verify LPSI pump differential pressure was not affected.

Based on the above, this event is not considered to be safety significant.

IV. Corrective Action

1 l

As a result of this event, the following actions have been, or will be, performed.

i 1.

A procedure will be implemented to administer and monitor the IST program prior to entry into MODE 6 from the current outage. This procedure will delineate the framework and requirements for adequate supervisory review of program implementation.

4 1

MRC FORM 366A 44-95) 4

. NRC FORM 366A U.S. NUCLEAR REGULATORY COMMisslON (4 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION 4

FACILITY NAME (1)

DOCKET LER NUMBER 16)

PAGE (3) l SEQUENTIAL REVislON YEAR NUMBER NUMBER 4OF4 j

l Millstone Nuclear Power Station Unit 2 05000336 j

97 - 005 -

00 l

k TEXT lit more space is required, use additional copies of NRC Form 366A) (11) i 2.

The identified instrumentation deficiencies shall be evaluated and corrected prior to entry into MODE 6

)

from the current outage.

Additional corrective action relating to the identification, calibration, and maintenance of installed instrumentation I

are being considered. If necessary, a supplement to this LER will be submitted to document these corrective action.

I V.

Additional Information

]

]

Similar Events

l Similar events involving ASME Section XI Testing deficiencies include:

LER 96-030: On August 20,1996 a review of the IST Program procedure identified that the pump and valve testing system design bases for the HPSI system pump discharge check valves was incorrect and the required ASME Section XI Testing for the HPSI System pump discharge check valves was not being performed. The 3

associated review for this event identified that the safety function of all check valves listed in the Pump and

]

Valve Testing Bases Document should be reviewed to verify that their bases descriptions are correct, and the corresponding surveillance procedures should be reviewed and revised, if appropriate, to ensure that safety functions are adequately tested.

]

LER 94 039: On October 31,1994, it was identified that some components were potentially not included in the IST program. The associated review for this event identified one valve that needed to be added to the program and seventeen valves that required testing beyond that specified in the program.

Energy Industry identification System (Ells) codes are identified in the text as [XX].

l 1

s k

i j

I l

3 i

i d

l