05000423/LER-1996-046-01, :on 961108,time Response of Containment Fuel Drop Radiation Monitor Less Conservative Then Value Assumed within Fsar.Caused by Inadequate Original Design & Failure of Surveillance Test Program.Design Reviewed

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:on 961108,time Response of Containment Fuel Drop Radiation Monitor Less Conservative Then Value Assumed within Fsar.Caused by Inadequate Original Design & Failure of Surveillance Test Program.Design Reviewed
ML20132B854
Person / Time
Site: Millstone 
Issue date: 12/07/1996
From: Peschel J
NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20132B817 List:
References
LER-96-046-01, LER-96-46-1, NUDOCS 9612170420
Download: ML20132B854 (4)
v • d • e


LER-1996-046, on 961108,time Response of Containment Fuel Drop Radiation Monitor Less Conservative Then Value Assumed within Fsar.Caused by Inadequate Original Design & Failure of Surveillance Test Program.Design Reviewed
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(iii)
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
4231996046R01 - NRC Website
v • d • e

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NRC FORM 366 U.s. NUCLEAR REGULATORY COMMisslON APPROVED BY OMB NO. 3150-0104 (4-95)

EXPIRES 04/30'95

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S So RS EPO UD 5

n^c""9o'L'u"a"! "^W"a'?os!."A".l 2a?'o'nME i

UCENSEE EVENT REPORT (LER) 15!%^u? Tuct!!""^lsta"ota'E"PJ,s"^a^ 'a"!,~"!'s"&

r si'AT/hA,,"foEIT1'l#1"E"H'A%3"oi"8fRf'""

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(See reverse fOr required number of digits / characters for each block)

FACIUTY NAME (1)

DOCKET NUMBER 12)

PAGE (3)

Millstone Nuclear Power Station Unit 3 05000423 1 of 4 i

i TITLE 14)

Time Response Of Containment Fuel Drop Radiation Monitor Less Conservative Then Value Assumed Within FSAR I

l 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 11 08 96 96 046 00 12 07 96 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11) i McDE (9) 5 20.2201(b) 20.22o3(aH2)(v)

So.73(a)(2)(i)

So.73(a)(2)(viii)

POWER 20.2203(aH1) l LEVEL (10) 000 20.22o3(a)(3Hi)

So.73(a)(2)(ii)

So.73(a)(2)(x) 20.2203(a)(2Hi) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 i

20.2203(a)(2Hii) 20.22o3(aH4)

So.73(aH2Hiv)

OTHER

"'"N"""""""""""

20.2203(a)(2)(iii) 50.36(cH1)

X So.73(a)(2Hv) so.c., Yin Ao.ir.cinoio-j 20.22o3(a)(2)(iv)

So.36(c)(2) 50.73(a)(2)(vii) j LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER linclude Area Codel J.M. Peschel, MP3 Nuclear Licensing Manager (860)437-5840 i

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUF ACTURE8I REPORTABLE

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS TO NPRDS i

69 3

1<

J SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR SUBMISSloN g

YES No (If yes, complete EXPECTED SUBMisslON DATE).

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

On November 8,1996, at 1330, with the plant in Mode 5, it was determined that the containment purge isolation time r:sponse was greater than that specified in the Final Safety Analysis Report (FSAR). The condition was discovered during th3 performance of containment purge isolation time response testing as required by Technical Specifications and in 4

cccordance with the requirements contained within the Technical Requirements Manual. On November 8,1996 at 1413, a i

prompt report was made pursuant to 10CFR50.72(b)(2)(iii)(C and D) as a condition that alone could have prevented the fulfillment of a safety function of a system needed: to control the release of radioactivity; and, to mitigate the consequences of an accident.

ThD root causes for this event have been identified as: 1) Inadequate original design and; 2) Failure of the surveillance test program and initial startup test program to assure compliance with the FSAR.

3 Tha Containment Fuel Drop Radiation Monitor design will be reviewed. Based on this review the design of the monitor will be modified or a Technical Specification change request will be submitted to the NRC for approval. A design review of response time data for safety related non-Engineered Safety Features (ESF) actuations which support FSAR chapter 15 cn: lysis will be performed to ensure that the data is supported by surveillance procedures.

9612170420 961206 PDR ADOCK 05000423 g

PDR

  1. NRC FORM 366A
  • U.S. NUCLEAR REGULATORY Commission (4-95)

I LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISloN Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 2 of 4 96 046 00 TEXT (if more space is required, use additional copies of NRC Form 366A) (17)

J 1.

Descriotion of Event On November 8,1996, at 1330, with the plant in Mode 5, it was determined that the Containment Purge Isolation time i

risponse was greater than that specified in the Final Safety Analysis Report (FSAR). The condition was discovered during the performance of Containment purge isolation time response testing as required by Technical Specifications and in accordance with the requirements contained within the Technical Requirements Manual (TRM). On November 8,1996 at 1413, a prompt report was made pursuant to 10CFR50.72(b)(2)(iii)(C) and 10CFR50.72(b)(2)(iii)(D) as a condition that alone could beva p;evented the fulfillment of a safety function of a system needed: to control the r-lease of radioactivity; and, to mitigate the consequences of an accident.

l The Containment Fuel 3 rop Radiation Monitors (RMS*RlY41 and RMS*RlY42) were being response time tested to m:et the Technical Sp ecification Surveillance requirements established in License Amendment 129. This testing ditermined that the rautation monitor could not meet the response time requirements of the TRM. This condition is being reported pursuant to 10CFR50.73(a)(2)(v)(C & D) as a condition that alone could have prevented the fulfillment of a safety function needed to control the release of radioactive material or mitigate the consequences of an accident.

]

The Containment Purge System had previously been declared inoperable due to an overdue surveillance and the Containment Purge Isolation valves (3HVU*CTV32A/B and 3HVU*CTV33A/B) had been closed. No fuel movement was being performed.

II.

Cause of Event

The causes for this event have been identified as:

1.

Inadequate original design. The design process goveming the initial design relied on vendor calculation and inputs which were not adequately verified by plant design engineering or by onsite testing.

2.

Failure of the surveillance test program and initial startup test program to assure compliance with the FSAR.

The process to assure the complete logic circuit was adequately response time tested and therefore met the response time requirements contained within the FSAR was inadequate.

Ill. Analysis of Event The FSAR in Section 15.7.4.2 states, that transit time for the purge air from the inlet of the containment purge duct to th3 isolation valves (3HVU*CTV32A/B and 3HVU*CTV33A/B) is 5.0 seconds based upon 2 seconds for the detector to rispond and 3 seconds for valve closure.

A calculation was performed in 1985 to verify that the containment purge isolation valves will close before contaminated air would reach the isolation valves after receiving a signal from the radiation monitor (s). This calculation was revised in 1993. Based on this revised calculation, contaminated purge air will not reach the isolation valves for 5.89 seconds. The 5 seconds specified in the FSAR was therefore bounded by analyzed response time in that the isolation valves would close prior to radioactive air reaching the valves. However neither the FSAR, nor the unit's Technical Requirements Manual were updated to reflect the revised response time.

)

The containment fuel drop radiation monitor utilizes a microcomputer with two electronic filters, one of which has a viriable response times based upon the magnitude of the input signat. The variable radiation monitor response time 9

gRgFORM 366A*

U.s. NUCLEAR REGULATORY Commission 4

UCENS3E EVENT REPORT (LER)

TEAT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REvislON j

Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 3 of 4 4.

96 046 00 i

TEXT (If more space is required. use additional copies of NRC Form 366A) (11) c::used the containment fuel drop radiation monitors to fail the surveillance test. As a result, the containment purge i

isolation has never been capable of performing its specified response time functions.

)

j There were no adverse safety consequences as a icstt of this condition in that the unit did not experience a fuel hrndling accident requiring closure of the Containment Purge Isolation Valves. However, based on the above analysis, the containment fuel drop radiation monitors are incapable of meeting their response time requirements.

Therefore, the radiological consequences of a fuel handling accident in containment, as documented in FSAR chapter 1

15.7.4.2.2, could be more severe then previously analyzed.

I

IV. Corrective Action

The following corrective actions will be implemented:

1.

The Containment Fuel Drop Radiation Monitor design will be reviewed. Based on this review the design of the monitor will be modified or a Technical Specification change request will be submitted to the NRC for approval by May 1,1997 2.

As part of the ongoing FSAR design review, the response time data for safety related non-Engineered j

Safety Features (ESF) actuations which support FSAR chapter 15 analysis will be reviewed to verify that the data is supported by surveillance procedures. Applicable surveillance procedures will be updated as l

necessary prior to entry into mode 4.

i 1

3 3.

As part of the ongoing FSAR review, applicable sections of the FSAR will be revised to accurately reflect the isolation time of the Containment Purge isolation valves ((3HVU*CTV32A/B and 3HVU*CTV33A/B) l by January 31,1997.

4.

Coordinated with the ongoing FSAR review, Section 3.2.2 of the Technical Requirements Manual (TRM) l will be revised to accurately reflect the response time of the Containment Purge Exhaust and Supply Valves radiation monitors by February 28,1997.

)

i V.

Additional Information

4 During the evaluation of this condition NNECO determined that the letter which submitted the License Amendment 4

Rrquest that was issued as License Amendment 129 stated that the response time was 5.29 seconds. NNECO 4

attributes this discrepancy to be a typographical error.

I

Similar Events

LER 96-042-00:

Incompletelv implemented Technical Specification Amendment Resultino in a Missed

]

Surveillance on RMS Monitors On October 28th,1996 at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, with the plant in Mode 5, it was discovered that j

Technical Specifications (TS) response time surveillance testing of Containment Fuel Drop Radiation Monitoring System (RMS) monitors 3RMS-RE41 and 3RMS-RE42 had not been

  • NRC FOIM 366A' U.S. NUCLEAR REGULATORY Commission (4-9 51 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION Millstono Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 4 of 4 96 046 00 TEXT tif more space is required, use additional copies of NRC Form 366A) (17) performed. The RMS monitors were declared inoperable and the containment purge & vent valves were shut as required by the applicable TS ACTION statement.

This condition was reportable pursuant to 10CFR50.73 (a)(2)(i)(B), as any operation or condition prohibited by the plant's Technical Specifications.

Response time testing of the Containment Fuel Drop Instrument channels will be completed prior to retuming the Containment Purge & Vent to service. The procedure goveming 1

licerJ9 amendment incorporation and implementation will be revised to clarify roles and respiisibilities. Licensing personnel will be trained on the requirements and responsibilities associated with developing and processing Technical Specification change requests.

1 LER 96-008-00:

Reactor Protection System Lead / Lao Time Constants Set Non-Conservatively On April 12,1996, at 14:30, with the plant iri Mode 5, it was discovered that time constants used on lead / lag cards for Overpressure Delta Temperature and Overtemperature Delta Temperature Reactor Trip setpoints as well as Steam Line Negative Rate -High Main Steam Line Isolation setpoints may be set non-conservatively. A subsequent review determined i

that these time constants specified in plant Technical Specifications for Overpressure Delta i

Temperature and Overtemperature Delta Temperature Reactor Trip setpoints were used as an input to calculate Safety Analysis Limits, thus affecting Limiting Safety System settings.

The root cause of the non-conservatively set time constants was the failure to identify conservative calibration requirements for Reactor Protection circuits in plant Technical

)

Specifications.

Manufacturer Data Ells System Codes Radiation Monitoring System - RM Reactor Building Environmental Control System - VA Ells Component Codes Indicator, Radiation - RI Transmitter, Radiation - RT K: man lonization Chamber model KDI-10

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