05000245/LER-1996-046, :on 960627,failure to Perform Applicable 10CFR50 Appendix J Tests to Satisfy TSs Occurred. Caused by Lack of Support for Conservative Decision Making at Millstone Unit 1.Appendix J Program Has Been Completed

:on 960627,failure to Perform Applicable 10CFR50 Appendix J Tests to Satisfy TSs Occurred. Caused by Lack of Support for Conservative Decision Making at Millstone Unit 1.Appendix J Program Has Been Completed

ML20138Q239
Person / Time
Site:	Millstone
Issue date:	03/03/1997
From:	Robert Walpole NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20138Q205	List: 05000245/LER-1996-046, :on 960627,failure to Perform Applicable 10CFR50 Appendix J Tests to Satisfy TSs Occurred. Caused by Lack of Support for Conservative Decision Making at Millstone Unit 1.Appendix J Program Has Been Completed 05000245/LER-1996-046-03, Forwards LER 96-046-03,documenting an Event That Occurred at Millstone Nuclear Power Station,Unit 1 on 960627 Per 10CFR50.73(a)(2)(i).Util Commitments Made within Ltr, Listed
References
LER-96-046, LER-96-46, NUDOCS 9703060226
Download: ML20138Q239 (6)

LER-1996-046, on 960627,failure to Perform Applicable 10CFR50 Appendix J Tests to Satisfy TSs Occurred. Caused by Lack of Support for Conservative Decision Making at Millstone Unit 1.Appendix J Program Has Been Completed

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2451996046R00 - NRC Website

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROtfED BY OMB NO. 31bo-0104 (4-95)

EXPIRES 04/30/98 N[oN4TSoUco$tc'r"oIn7o"uYsT [oNs e NTEo "LSN n'a"'?o'L'u"l?ti "^!'amo T4!a" 42E!Uo '"e41 LICENSEE EVENT REPORT (LER) l'!W'u'? '=2""^lo*"t&"4an,s"s'"to'n'Nes"elc D#c'Mb.^~esria'l^n" sea"Jf%"1"o?"8eTe"*'-

u r (See reverse for required number of digits / Characters for eaCh block)

FOCluTY NAME m DOCKET NUMBER (2)

PAGE13)

Millstone Nuclear Power Station Unit 1 05000245 1 of 6 l

TITLE (4)

Failure to Perform Applicable 10CFR50 Appendix J Tests to Satisfy Technical Specifications 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 06 27 96 96 046 03 03 03 97 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)

MODE m N

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

X 60.73(a)(2)(i) 50.73(a)(2)(viii>

POWER LEVEL (10) 000 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)hi) 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)

,,....,..o.....

20.2203(a)(4) 50.73(aH2)(iv)

OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2Hv) specif y in Abstract below or in NRC Form 366A 20.2203(aH2)bv) 50.36(cH2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER ilnclude Area Codel 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 I

j SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR SUBMISSION f

YES NO l

(If yes, complete EXPECTED SUBMISSION DATE).

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

On June 27,1996, with the plant shut down and the reactor in the COLD SHUTDOWN condition, a self assessment program review determined that several containment penetrations did not have adequate local leak rate tests (LLRTs) performed pursuant to the requirements of 10CFR50 Appendix J. Not all testable connections were tested on ten 18" atmospheric control system valves, and there are no testable flanges on two 2" atmospheric control system l

valves. Additionally, a piping flange in the head spray system is not appropriately tested.

l The failure to perform individual Type B leakage tests in accordance with the requirements of 10CFR50 Appendix J l

results in the inability to adequately demonstrate primary containment integrity, which is required to be maintained by Millstone Unit 1 Technical Specification 3.7 A.3. This event is reportable pursuant to 10CFR50.73(a)(2)(i)(B) as a 4

condition prohibited by the plant's Technical Specifications.

Modifications to make the penetrations testable and subsequent testing will be completed prior to startup from the current refueling outage. There were no safety consequences as a result of this event, wo,- ennu wa <4 nsi 9703060226 970303 PDR ADOCK 05000245 S

PDR

'U.S. NUCLEAR REGULAVORY COMMISSION (4-95) 1 LICENSEE EVENT REPORT (LER) j TEXT CONTINUATION i

FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION j

Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 2 of 6 96 046 03 J

TEXT Uf more space is required, use additionalcopies of NRC Form 366A) (11) 1, l.

Description of Event

i On June 27,1996, with the plant shut down and the reactor in the COLD SHUTDOWN condition, a self assessment program review determined that several containment penetrations did not have adequate LLRTs 4

i l

performed pursuant to the requirements of 10CFR50 Appendix J.

Not all testable connections were tested on ten 18" atmospheric control system valves, and there are no testable flanges on two 2" atmospheric 3

control system valves. Additionally, a piping flange in the head spray system is not appropriately tested.

1 The failure to perform individual Type B leakage tests in accordance with the requirements of 10CFR50 Appendix J results in the inability to adequately demonstrate primary containment integrity, which is required i

to be maintained by Millstone Unit 1 Technical Specification 3.7.A.3. This event is reportable pursuant to I

l 10CFR50.73(a)(2)(i)(B) as a condition prohibited by the plant's Technical Specifications.

Modifications to make the valves testable and subsequent testing will be completed prior to startup from the current refueling outage. There were no safety consequences as a result of this event.

II. Cause of Event

j i

Based on the completion of the root cause evaluation and the Appendix J Self-Assessment, the cause of the i

event is modified as follows:

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The cause of this condition was due to significant deficiencies in the Appendix J Program resulting in a lack of adequate program implementation and a non-conservative interpretation of Appendix J requirements, in 5

3 that the "as-built" systems and equipment configuration were not identified as needing modifications when

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the 10CFR50, Appendix J criteria were implemented at Millstone Unit No.1. The three root causes are:

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2. There was a lack of support for conservative decision making at Millstone Unit No.1.

1.

Management commitment to the Appendix J program was weak.

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3. Appendix J training was not consistent with the Millstone Unit No.1 system configurations.

i l

l Ill. Analysis of Event Atmospheric Control Vacuum Breakers 1-AC-1 A-J: Penetration X-202A - H The self actuated check valves (vacuum breakers) of the drywell to torus penetrations form a part of the atmospheric control system and the valve components are part of the reactor building boundary. Five i

testable gasketed joints exist on each of the valves, namely the flanged body cover, and the shaft packing gland and the stuffing box located at both ends of the disc shaft. Each of these containment boundaries is j

testable and should receive a Type B LLRT in accordance with Appendix J.

However, only the double gasketed flanged body cover was appropriately tested per Appendix J. The shaft packing gland and stuffing i

box (at each shaft end) utilize pipe plugs to communicate between the inner and outer barrier at the lantern a

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gland and between the o-rings respectively. These boundaries have not been LLRT tested. When the pipe plugs are removed, a Type B test can be performed to verify the leak tightness of these potentialleak paths.

This configuration exists on all ten of the Atwood & Morrill Co.,18 inch Vacuum Breakers used in the Torus to drywell penetrabons.

1 i

d NBC FORM 366A (4 95) i

,U.s. NUCLEAR REGULATORY COMMisslON 14-95)

UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL AEVISION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 3 of 6 96 046 03 TEXT tif more space is required, use additional copies of NRC Form 366A) t17)

Since three of the five Type B tests are performed on static seals, it is unlikely that seal degradation would lead to a sufficient loss of leak tightness to result in a failed test. The review of the historical data of Type B test results of the body flange for all vacuum breakers has shown consistent leak tightness. The flange seals have shown continued acceptable leak tightness as indicated by LLRT results. Vacuum breakers are manually exercised quarterly to confirm operability.

Furthermore, two vacuum breakers have been overhauled each refueling outage since 1985. The overhaulincludes replacement of the packing and o-rings, and a complete inspection of all internal parts. The ILRT does challenge all the leakage barriers of the valves including the shaft seals, and no measurable leakage was observed at these locations.

Torus Exhaust Bvoass Valve 1-AC-9 & 1-AC-12 (X-25/202D)

The torus exhaust bypass valves are 2" air operated containment isolation valves located on a branch line from drywell vent penetration X-25/202D. Valves 1-AC-9 & 1-AC-12 exhibit standard fiber gasket material on both the upstream and down stream flange connections and do not exhibit testable flanges in either the upstream or downstream direction. The gaskets in the downstream direction (outside of containment) are tested during Type C testing of the system valves; however the inboard gaskets are only tested during the Type A ILRT of the overall containment.

Consecutive ILRTs have been performed during 1991 (RFO13),

1994 (RFO14) and is scheduled for current RF015. No measurable leakage has been observed at these locations. Since the previous ILRTs have been identified as being potentially invalid (LER 96-026-01),

resulting from the inability to perform Type C testing on several penetrations, any reference to historical ILRTs as a basis for acceptance has been eliminated where appropriate.

Head Sorav System Flanae X-17 l

A piping system flange exists between the inboard containment isolation check valve 1-HS-5 and the outboard motor operated containment isolation valve 1 HS-4 to allow removal of the head spray piping when removing the reactor head cover. It was previously thought that the containment boundary which includes the flange is subjected to the test pressure during Type A ILRT of the overall containment. Subsequent l

review of the configuration identified that the flange is not independently tested as part of the ILRT. The existing test configuration does not provide for an appropriate Type B LLRT of this flange. An alternate test configuration will provide for an appropriate Type B LLRT of this flange.

The failure to adequately perform Appendix J testing on these penetrations calls into question the ability to demonstrate the operability of the primary containment during the past operating cycles. Since Millstone has failed to adequately demonstrate primary containment integrity, as required by Technical Specification 3.7.A.3, this event is reportable pursuant to 10CFR50.73(a)(2)(i)(B) as a condition prohibited by the plant's l

Technical Specifications.

l Annendix J Proaram Self-Assessment Results i

A self-essessment of the Appendix J program, including a review of all containment penetrations and associated drawings, LLRT procedures and the ILRT procedure was completed. The following summarizes l

the reportable issues identified; l

1.

Several system penetrations did not have their Type B or Type C minimum pathway penalty applied to the ILRT test result. The penetrations are identified as X-9A/B, X-10A, X-11B, X-14, X-15, X-16A/B, X-17, X-18, X-19, X 23, X-24, X-25, X-30f, X-34f, X-35A-E, X-36, X-37A-D, X-38A-D, X-39A/B, X-42, I

X-43, X-47, X-200B, X-202G, X-205 and X-211 A/B. The impact on the ILRTs performed since 1980 NRC FORM 360A 44-951

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• U.s. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 4 of 6

)

96 046 03 TEXT (If more space is required. use additional copies of NRC Form 366A) (17) can be completely assessed once the modifications are made to containment penetrations X 9A/B and the LLRT is performed. This activity has been addressed in LER 96-026.

2. Type C testing of the Main Steam isolation Valves (MSIV) is performed at a reduced pressure at 25 psig and added to the total of all Type B and Type C tests, and compared to the Technical Specification limit of 0.6 La. The measured leakage at 25 psig is not corrected to the equivalent leakage of 43 psig as identified in inspection Report 50-245/77-05, dated April 18,1977. Failure to correct equivalent leakage prohibits proper summation of Type B and Type C tests.

3. Millstone Unit No.1 Technical Specifications do not state a separate acceptance criteria for the air lock testing. Paragraph lil.D 2.(b).(iv) of Appendix J states "The acceptance criteria for air lock testing shall be stated in the Technical Specifications". The surveillance procedure identifies an administrative limit rather than an acceptance criteria.

4.

Area temperature survey, in accordance with ANSI N45.4-1972, is required in advance of ILRTs to verify i

containment temperature is adequately acquired with the specific test configuration. No documentation identified to indicate that a temperature survey has been done.

5. The LLRT performed on Main Steam drain valve,1-MS-5, is in the reverse direction and considered a non conservative test. No evidence or justification has been identified to verify the validity of the reverse I

direction test.

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6. The ILRT containment air pressure and verification flow connections on containment penetrations X-37A and X-38A were not tested after the ILRT for addition to the 1994 ILRT results, since the configuration does not allow Type C testing.

7. Containment penetration X-35E, TIP Purge, has only one containment isolation valve at the penetration where two barriers are required per Appendix A design criteria. The one containment isolation valve is Type C tested.

1 8.

For penetrations X-25 and X-202D, containment isolatb valves 1-AC-9 and 1-AC-12 have exemptions for reverse direction testing based on these valves being butterfly valves, however these valves are plug l

valves and the reverse direction testing is non-conservative.

l

9. For penetration X-30f and X-34f,1-RR-111 A/B did not have test connections to allow Type C test.

10.For penetration X-30f and X-34f, test configuration for 1-RR 25A/B did not allow for adequate draining i

of water for Type C test.

l 11.For the following containment penetrations, the existing procedures do not allow adequate draining and/or venting of the penetration piping thus has resulted in an invalid Type C test of the valve. Any modifications or adjustments to these valves may inhibit our ability to assess the impact of the improper l

test configuration, thus these issues are included as reportable:

i I._

h yU.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3) 4 4

1 YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 5 of 6 96 046 03 TEXT fit more space is required, use additional copies of NRC Form 366A) (17) i i

X-211 A Post Accident Sampling 1-PAS-24,1-PAS-25 l

X-14 RWCU 1-CU-2A e

X-16A/B Core Spray 1 -CS-5A/B X-42 Standby Liquid Control 1 SL-7 3

l X-47 Reactor Recirculation 1 -R R-37 i

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IV. Corrective Action

j f

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The Appendix J program review has been completed and additional discrepancies which affect the LLRTs i

and ILRTs have been included in this LER supplement.

l As a result of the root cause investigation of this event that has been completed, the following are revised

corrective actions

Northeast Nuclear Energy Company (NNECO) will complete by May 31,1997 (prior to perforrning the next ILRT), the program administrative control document, " Containment Leak Rate Testing Program Administration," which will identify the specific roles and responsibilities of implementing and maintaining the program. This document will also list the primary containment isolation barriers with associated basis for each penetration.

NNECO will determine Appendix J Program staffing requirements and obtain program management approval by May 31,1997.

NNECO will develop an Appendix J action plan which will identify the importance of conservative decision making regarding implementation of Appendix J requirements from all perspectives including design, operation and engineering. NNECO will incorporate this plan into an administrative procedure, " Containment Leak Rate Testing Program Administration," by May 31,1997.

NNECO will determine Appendix J programmatic enhancement training requirements and include in administrative procedure, " Containment Leak Rate Testing Program Administration," by May 31, 1997.

NNECO willimplement this training by September 30,1997.

Modifications will be made to the procedure for performing the local leak rate test on the vacuum breakers to include testing of the shaft packing and stuffing box.

LLRTs will be performed before the overall containment Type A ILRT, and prior to startup for operating cycle 16.

Modifications will be made to the containment side bolting flange of valves 1-AC-9 & 1-AC-12, and changes will be made to the procedure to perform the LLRT on the bolting flange between valves 1-HS-4 and 1-HS-5 during refueling outage 15. LLRTs will be performed before the overall containment Type A ILRT, and prior to startup for operating cycle 16.

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

l NRC FoFtM 366A (4-95)

l

• NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION 1

(4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 6 of 6 2

96 046 03

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TEXT (If more space is required, use additional copies of NRC Form 366A) (17) j Additional Corrective Actions From Appendix J Proaram Self-assessment i

The following corrective actioris, to be completed prior to startup for operating Cycle 16, are the result of reportable items identified during the Appendix J program review.

All LLRT surveillance procedures, which require revision to provide a valid LLRT, will be revised prior to

{

performance of the as-left LLRT prior to startup for operating Cycle 16.

)

Evaluate the need to modify Millstone Unit No.1 Technical Specifications to include air lock leakage acceptance criteria.

l Revise ILRT procedure to include correct containment penetration valve lineup for each containment i

penetration and include containment temperature survey in accordance with ANSI N45-4 -1972 prior to I

performance of ILRT for RFO 15.

Modify containment penetrations X-30f and X-34f to allow valid Type C test of isolation valves 1-RR-111 A/B and 1-RR-25A/B.

Modify surveillance procedures for the following penetrations to allow valid Type C test:

I X-211 A Post Accident Sampling 1 -PAS-24,1 -PAS-2 5 X-14 RWCU 1.C U-2 A 1

X-16A/B Core Spray 1-CS-5 A/B X-42 Standby Liquid Control 1 -S L-7 X-47 Reactor Recirculation 1-RR-37 j

Revise ILRT procedure to provide LLRT for containment air pressure and verification flow connections on e

i containment penetrations prior to performance of ILRT prior to startup for operating Cycle 16, i

l Revise MSIV LLRT procedure to correct leakage to accident pressure of 43 psig.

l Modify containment penetration X-8 to allow valid LLRT on 1-MS-5 (committed to in LER 96-026-01 e

Commitment No. B15675-1).

Modify containment penetration X-35E by installing an additional containment isolation valve and associated test connections. Revise LLRT procedure to test the new containment isolation valve and implement testing.

1 Modify penetrations X 25 and X-202D to allow valid Type C test.

e 1

4 l

V.

Additional Information

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Similar Events

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l LER 96-026-01

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Manufacturer Data i

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None.

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