05000245/LER-1997-010-01, :on 970214,determined LLRT Pressure Being Used May Be Less than Accident Pressure.Caused by Weakness in Mgt Commitment to App J Program.Llrts Modified

:on 970214,determined LLRT Pressure Being Used May Be Less than Accident Pressure.Caused by Weakness in Mgt Commitment to App J Program.Llrts Modified

ML20136J592
Person / Time
Site:	Millstone
Issue date:	03/17/1997
From:	Robert Walpole NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20136J578	List: 05000245/LER-1997-010, Forwards LER 97-010-00,documenting Event Occurred at Unit 1 on 970214.Commitments Made within Ltr Submitted as Listed 05000245/LER-1997-010-01, :on 970214,determined LLRT Pressure Being Used May Be Less than Accident Pressure.Caused by Weakness in Mgt Commitment to App J Program.Llrts Modified
References
LER-97-010-01, LER-97-10-1, NUDOCS 9703200148
Download: ML20136J592 (3)

LER-1997-010, on 970214,determined LLRT Pressure Being Used May Be Less than Accident Pressure.Caused by Weakness in Mgt Commitment to App J Program.Llrts Modified

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2451997010R01 - NRC Website

text

i NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB No. 3150-0104 (4-95)

EXPlREs 04/30/98 coLIEC'r N REQUE ST So o M S E O 1ED L o

rio L'#Mo^%%3"! "^ Pane '"'on"s" Jef8S/o ^N!R LICENSEE EVENT REPORT (LER) 15"^U ? 'ac5?"*A'a ^r&"'?A"As"s*#%'N 'a^s" e

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(Sea reverse for required number of digits / characters for each block) l Ft.CitrrY NAMI (1)

DOCKET NUMBER (2)

PAGE (3)

Millstone Nuclear Power Station Unit 1 05000245 1 of 3 TITLE I4)

LLRT Test Pressure Less Than Accident Pressure EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACluTY NAME DOCKET NUMBER NUMBER

" " ~^"'

02 14 97 97 010 00 03 17 97

~ OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)

ODE m N

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

X 50.73(a)(2)(i) 50.73(a)(2i(viii)

POWER 20.2203(a)(1)

LEVEL (10) 000 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x) 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv)

OTHER 50.36(c)(1) 50.73(a)(2)(v) specify in Abstract below

)

20.2203(a)(2)(iii) or in NRc Form 366A 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(al(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER tinclude Area Code)

Robert W. Walpole, MP1 Nuclear Licensing Manager (860)440-2191 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

cAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE To NPRDs To NPRDs CUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR S BWSSION

[

YES NO (if yes, complete EXPECTED SUBMISSION DATE).

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

On February 14,1997, with the plant in COLD SHUTDOWN, a review determined that the Local Leak Rate Test (LLRT) pr:ssure being used may be less than accident pressure. Design Basis Accident Pressure is 43 psig per Updated Final Safety Analysis Report, Section 6.2.1.1.1. Test pressure during the LLRT is also 43 psig. Use of LLRT test pressure equal to Design Basis Accident pressure fails to take into consideration pressure gauge inaccuracies and line losses due to flow through the test rig and flow meters. The test gauge was located at the inlet of the flow meters and as much as 200 feet of small diameter tubing was used to connect to the penetration. The failure to test at accident pressure invalidates past l

LLRTs and results in the inability to adequately demonstrate containment integrity. Containtnent integrity is required by Millstone Unit No.1 Technical Specifications, Section 3.7.A.3, thus this event is reportable as a condition prohibited by the l

plant's Technical Specifications. The cause of this event is weaknesses in the management commitment to the Appendix J l

Program resulting in an inadequate technical review of the implementing procedures.

Th:re were no safety consequences as a result of this event. Procedural changes and changes to the test rig configuration era being undertaken to increase the LLRT pressure thus assuring a full Design Basis Accident Pressure is applied to the volume under test. A review of impacted test results will be undertaken and tem results conservatively corrected.

i I

l 9703200148 970317 PDR ADOCK 05000245 S

PDR

C NRC FOhM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKliT NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 2 of 3 97 010 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) 1.

Description of Event

On February 14,1997, with the plant in COLD SHUTDOWN, during a review of a procedure as part of the current 10CFR50.54(f) review, it was determined that the LLRT test pressure being used may be less than accident pressure. Design Basis Accident pressure is 43 psig per Updated Final Safety Analysis Report Section 6.2.1.1.1.

Test pressure during the LLRT is also 43 psig. Use of LLRT test pressure equal to Design Basis Accident pressure fails to take into consideration pressure gauge inaccuracies and line losses due to flow through the test rig and flow meters. The test gauge was located at the inlet of the flow meters and as much as 200 feet of small diameter tubing was used to connect to the penetration. The failure to test at accident pressure invalidates past LLRTs and results in the inability to adequately demonstrate containment integrity.

11.

Cause of Event

The cause of this event is weaknesses in the management commitment to the Appendix J Program resulting in an inadequate technical review of the implementing procedures. The LLRT test pressure was maintained at an accident pressure of 43 psig at the test rig and not at the test volume. No allowance was provided for instrument accuracy, pressure drop acnss the rotameters, or the pressure drop along the test tubing.

Ill. Analysis of Event The failure to test at accident pressure invandates past LLRTs and results in the inability to adequately demonstrate containment integrity. Containment integrity is required by Millstone Unit No.1 Technical Specifications, Section 3.7.A.3.

Thus this event is reportable pursuant to 10CFR50.73(a)(2)(i)(B) as a condition prohibited by the plant's Technical Specifications.

A review of the LLRT test rig configuration and procedures has determined inaccuracies in reported LLRT results have been introduced by the test rig configuration, equipment and surveillance procedures. Specific problems noted:

1. The LLRT surveillance procedures require adjusting the pressure regulator to obtain 43 psig (25 psig for main steam isolation valves (MSIVs)). The pressure is read at the test rig which is not representative of the volume undergoing the test.

2. The pressure gauge used for testing has an accuracy of +/- 0.2 psi. Accuracy was not considered in I

setting the test pressure.

3. The pressure gauge was located on the inlet of the rotameters on the LLRT test rig. No allowance was made for the effects of pressure drop through the rotameter and test line length to the penetration under test.

4. The rotameter temperature correction formula used in the LLRT procedures is inaccurate for the rotameters in the LLRT test rig. The formula is accurate for sharp edge rotameter floats whereas the LLRT test rig uses ball floats.

These items failed to ensure the penetration volume under test was actually at accident pressure during the test.I l

}

J

i A

9 hNRC FC*AM 36CA U.S. NUCLEAR REGULATORY COMMISSION f

14-9 51 UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION

{

FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER 16)

PAGE (3)

YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 3 of 3 l

97 010 00 1

i TEXT (if more space is required, use additional copies of NRC Form 366A) (11)

There were no actual safety consequences as a result of this event. However, the failure to test at accident pressure invalidates past LLRTs and resuits in the inability to adequately demonstrate containment integrity.

Containment integnty is not required under the current plant condition of being in cold shutdown and reactor defueled.

j t

I IV. Q. orrective Action l

Procedural changes and changes to the test rig configuration are being undertaken to increase the LLRT pressure thus assuring a full Design Basis Accident pressure is applied to the volume under test. A review l

l of impacted test results will be undertaken and test results conservatively corrected. The corrected results l

will be used to verify all acceptance criteria remain satisfied prior to startup from the current refueling outage. Specifically:

l Northeast Nuclear Energy Company (NNECO) has modified LLRT test rigs to relocate Heise pressure e

gauge from rotameter inlet to rotameter outlet.

NNECO will perform analyses and evaluations to determine necessary procedure and process changes to ensure full pressum LLRTs are conducted with at least 43 psig (25 psig for MSIVs) at the penetration.

i-l NNECO will modify the LLRT series of procedures to incorporate changes needed to ensure at least 43 psig (25 psig for MSIVs) at the penetration for full pressure LLRTs.

NNECO will correct and recalculate the LLRT results used in the running total for the current iefuel outage using the Franklin Research Center methodology.

i The Appendix J Program is being reviewed as part of the on-going 10CFR50.54(f) review effort.

e As committed to in LER 96-046-03, NNECO will develop and implement a Containment Leak Rate e

Testing Program Administration Manual to provide a clear delineation of Appendix J program responsibilities, in-plant equipment design and configuration information, training requirements, references, test equipment operation, setup and use.

I V.

Additional Information

Similar Events i

LER 96-046-03

• Failure to Perform Applicable 10CFR Appendix J Tests to Satisfy Technical Specifications" Manufacturer Data I

Not Applicable a

I 9$RC FORM 366A (4-95)

I