05000336/LER-1996-029, :on 960715,discovered That Evaluation for Removal of Startup Rate Trip Feature Was Potentially non-conservative.Caused by Failure to Adequately Identify Design Basis Requirements.Modified Procedures

:on 960715,discovered That Evaluation for Removal of Startup Rate Trip Feature Was Potentially non-conservative.Caused by Failure to Adequately Identify Design Basis Requirements.Modified Procedures

ML20140C946
Person / Time
Site:	Millstone
Issue date:	06/03/1997
From:	Joshi R NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20140C938	List: 05000336/LER-1996-029, :on 960715,discovered That Evaluation for Removal of Startup Rate Trip Feature Was Potentially non-conservative.Caused by Failure to Adequately Identify Design Basis Requirements.Modified Procedures 05000336/LER-1996-029-01, Forwards LER 96-029-01,providing Update on Analyses & Investigation for Event.Commitments Made within Ltr,Listed
References
LER-96-029, LER-96-29, NUDOCS 9706100150
Download: ML20140C946 (5)

LER-1996-029, on 960715,discovered That Evaluation for Removal of Startup Rate Trip Feature Was Potentially non-conservative.Caused by Failure to Adequately Identify Design Basis Requirements.Modified Procedures

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(v), Loss of Safety Function
3361996029R00 - NRC Website

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' NRC FORM 366 U.s. NUCLEAR REGULATORY COMMisslON GPPHOtfED BY OMB NO. 3160-0104 (4-95)

EXPIRES 04/30/98 I"YoE'e's% "f8'*01'e l"J,2rnlo~n"M'M'at oWoIM"I"Yo"#SI 'oN59 EoNo^UE LICENSEE EVENT REPORT (LER) n'a"!8i s

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7-FActuTV NAME (1)

DOCKET NUMBER (2)

PAGE (3)

Millstone Nuclear Power Station Unit 2 05000336 1OF5 TITLE 14)

The Evaluation for Removal of Startup Rate Trip Feature was Potentially Non-Conservative EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8) sEQU AL RE N

MONTH DAY YEAR YEAR MONTH DAY YEAR U

U R

07 15 96 96

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01 06 03 97 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)

MODE (9) 5 20.22o1(b) 20.2203(a)(2)(v)

So.73(a)(2)(il 50.73(a)(2)(viii)

POWER 20.22o3(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii)

So.73(a)(2)(x)

LEVEL (10) 000 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)

So.73(a)(2)(iv)

OTHER

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20.22o3(a)(2)(iv)

So.36(c)(2)

So.73(a)(2)(vii) pf NRC Form 366A LICENSEE CONTACT FOR THIS LER (12) i NAME TELEPHONE NUMBER linclude Area Codel R. G. Joshi, MP2 Nuclear Licensing (860) 440-2080

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

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CAUSE

SYSTEM COMPONENT MANUFACTURER

CAUSE

SYSTEM COMPONENT MANUFACTURER PRD PRDS SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR YES SUBMISSloN 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 July 15,1996 at approximately 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> with the plant in Mode 5 at 0% power, it was discovered that the analysis for the removal of the High Startup Rate (SUR) Trip from Millstone Unit 2 in 1978 may not have addressed all possible control rod withdrawal events. The SUR Trip feature was removed in accordance with Amendment number 38 to the Technical Specifications issued on April 19,1978. This event is being reported pursuant to the requirements of 10 CFR 50.73(a) (2) (v),"any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe i

shutdown condition."

The cause of this event was the failure to adequately identify the design basis requirements of the SUR Trip. This event caused an inconsistency between the safety analysis and plant configuration for Cycles 2 and 3. Millstone 2 safety analysos for Cycle 4 and later recognized the removal of the SUR Trip. The investigation of this event demonstrated that Millstone 2 can operate without the SUR trip.

There were no automatic or manually initiated safety systems activated as a result of this event, This revision is a complete re-write to the LER.

9706100150 970603 PDR ADOCK 05000336 S

PDR NRC FORM [66 i4-95)

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1U.S. NUCLEAR REGULATORY COMMisslON 14 95)

LICENSEE EVENT REPORT (LER) 1 TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER 16)

PAGE (3)

SEQUENTIAL Revision R

NUMBER NUMBER 2OF5 Millstone Nuclear Power Station Unit 2 05000336 96

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01

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TEXT fit more space is required, use additional copies of NRC Form 366A) (11) 1 1.

DescriDtion of Event l

On July 15,1996 at approximately 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />, with the plant in Mode 5 at 0% power, it was discovered that the analysis for the removal of the High Startup Rate (SUR) Reactor Trip from Millstone Unit Number 2 in 1978 may i

not have addressed all possible control rod withdrawal events. The SUR Trip feature was removed in j

accordance with Amendment number 38 to the Technical Specifications issued on April 19,1978, s

l On July 17,1996 at 2018 hours0.0234 days <br />0.561 hours <br />0.00334 weeks <br />7.67849e-4 months <br />, while in Mode 5 at 0% power, an immediate report was submitted pursuant to the requirements of 10 CFR 50.72(b)(2)(iii)(A), "any event or condition that alone could have prevented the j

fulfillment of the safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition."

During the investigation of this event, an error was identified in the current cycle Uncontrolled Control Element i

Assembly (CEA) Bank Withdrawal analysis. This second event concemed the use of power peaking factors in this analysis which did not bound all possible operating conditions. This second error was reported in an update i

report on August 12,1E96. Subsequent analysis has shown that this error is not reportable in that the use of incorrect peaking factor 5 would not allow operation of the facility outside design basis limits.

1 There were no automatn or manually initiated safety systems activated as a result of this event. Additionally, no l

operator actions were required in response to this event since the SUR Trip was removed in 1978 and this is a report of an historical event.

This event is being reported per 10 CFR 50.73(a)(2)(v), "any event or condition that alone could have prevented j

the fulfillment of the safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition."

I

II. Cause of Event

l The cause of this event was the failure to adequately identify the design basis requirements of the SUR Trip.

Ill. Analysis of Event The original plant design for Millstone Unit No. 2 included a Start Up Rate (SUR) Trip that was described in the initial pre-license submittals as an equipment protection trip. The Final Safety Analysis Report (FSAR) accepted by the Staff on May 10,1974, credited the Variable Over Power (VOP) Trip as providing protection against reactivity addition events. The SUR Trip was not credited in the FSAR. In 1977, Northeast Nuclear Energy Company (NNECO) determined that the best solution to problems conceming spurious actuations of the Reactor Protectiori System would be to remove certain equipment protection trips (including the SUR trip) and the l

setpoint requirements from the Technical Specifications. Confirmation that the SUR trip could be removed from the Technical Specifications was received from Combustion Engineering in a letter dated July 29,1977, which j

stated that "no credit was taken for the High SUR Trip in the safety analyses and that no safety limit is directly related to the trip." Using this information, NNECO, in the letter of September 2 1977, requested an amendment to the Technical Specifications which would eliminate the SUR Trip.

t i

NNECO received the amendrnent allowing removal of the SUR trip in Amendment number 38 to the Technical Specifications, with corresponding Safety Evaluation Report on April 19,1978. The SUR Trip was then removed

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from the plant prior to the start of Millstone Unit 2, Cycle 2.

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. -. --U.S. NUCLEAR REGULATORY COMMisslON i

(4-9M LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 0)

DOCKET LER NUMBER i6)

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SEQUENTIAL REVISION '

YEAR NUMBER NUMBER 3OF5 Millstone Nuclear Power Station Unit 2 05000336 96 029 01 TEXT (11 more space is required, use additional copies of NRC Form 366A) (17)

During the current Design Basis Review, it was questioned whether the physical removal of the SUR trip in 1978 placed Millstone Unit 2 outside of its design basis. ASEA Brown Boveri-Combustion Engineering (ABB-CE) informed NNECO by correspondence on July 15, July 18 and July 19,1996, that the original design of Millstone Unit 2 incorporated the protective action of a SUR trip to provide protection for subcritical Control Element Assemb!y (CEA) withdrawal. ABB-CE further explained that, with an operational SUR trip feature, an entire spectrum of events initiated from subcritical conditions were rendered non-limiting and therefore were not included within the plant accident analysis. As a result, the Cycle 2 and 3 safety analysis performed by 1

Combustion Engineering (C-E) addressed Uncontrolled CEA bank withdrawal events initiated from Hot Zero Power, and credited the Variable Over Power (VOP) Trip to terminate the event, without violation of the fuel design limits.

4 Therefore, Millstone Unit 2, Cycles 2 and 3, with the safety analysis performed by Combustion Engineering, implicitly credited the SUR trip to prevent Uncontrolled CEA bank withdrawal events initiated from suberitical conditions from becoming limiting. Therefore, the removal of the SUR trip during Cycles 2 and 3 allowed i

Millstone Unit 2 operation outside of the assumptions of the safety analysis. There was no documentation in the FSAR to explain the dependence on the SUR trip to prevent Uncontrolled CEA Bank Withdrawal events initiated from subcritical conditions from becoming limiting. Also during the discussions between NNECO and i

C-E related to removing the SUR trip from the Technical Specifications, no documentation has been found that

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explained this dependence on the SUR trip. An additional causal factor of this event is poor communications by NNECO with C-E. No documentation has been found that NNECO informed C-E that the SUR trip had been physically removed, instead of being just removed from the Technical Specifications. NNECO had opportunities j

at the start of Cycles 2 and 3 to identify the removal of the SUR trip to C-E during the initiation of the cycle i

designs, but no evidence was found that this communication had actually occurred.

l The actual safety significance of this event is low since no uncontrolled CEA withdrawal events occurred from a i

subcritical condition which would have required the SUR trip. The potential safety significance is high, in that if i

such an event had occurred, there is no assurance that fuel design limits would not have been exceeded.

a As a result of this discrepancy in the safety analysis for Cycles 2 and 3 (1978 through 1980) for the Uncontrolled CEA Bank withdrawal event from subcritical due to the SUR trip removal, a review of the Westinghouse supplied safety analysis for Cycles 4 through 9 (1980 through 1989) was also performed. Westinghouse began supplying fuel for Millstone 2 in Cycle 4 (1980). At that time Westinghouse analyzed the Uncontrolled CEA Bank withdrawal event from subcriticalin the Basic Safety Report (WCAP 9660). The Westinghouse supplied j

analyses did not credit the SUR trip. The Westinghouse supplied safety analysis indicated acceptable results for i

CEA withdrawal incidents from subcritical without the SUR Trip. The Westinghouse analysis uses the VOP Trip l

to terminate the event, with no fuellimit DNBR or centerline melt violations.

As a result of this discrepancy in the safety. analysis for 1978 through 1980 for the Uncontrolled CEA Bank Withdrawal event from subcritical due to the SUR trip removal, a review of the Siemens Power Corporation (SPC) supplied safety analysis for Cycles 10 through 13 (1989 to present) was also performed. SPC began supplying fuel for Millstone 2 in Cycle 10 (1989) and the safety analysis for the current fuel cycle (Cycle 13) is supplied by SPC. The SPC supplied Safety Analysis does not credit the SUR trip. The SPC supplied safety analyses indicated acceptable results for

• Uncontrolled Control Rod Assembly Withdrawal from a subcritical or low power startup condition" without the SUR Trip installed. The SPC analysis uses the VOP Trip to terminate the event, with no fuel DNB or centerline melt limit violations. It was SPC's intention to make this analysis bound Mode 2 and Mode 3 by using the most limiting parameters from either Mode as the input to this analysis. For 4

i example, the initial power level used in the analysis is 1E-9 of full power However, as a result of the review performed associated with this event an error was discovered in this analysis. This review determined that non-conservative power peaking factors were used in the Siemens analysis of the CEA bank withdrawal event for l

NRC FORM 306A (4-DM

- U.s. NUCLEAR REGULATORY COMMISSION (4 at UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER 4)

PAGE (3)

SEQUENTIAL REVislON EAR NUMBER NUMBER 4OF5 Millstone Nuclear Power Station Unit 2 05000336 96

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01 TEUT (If more space is reqwred, use additional copies of NRC Form 366A) (17)

Modes 2 and 3. As a result, the analwis did not consider the increased radial and axial peaking factors for rod insertion to the power dependent insertion iimits (PDll's) allowed in Mode 2, and did not address the impact on 4

radial and axial peaking factors from potential Mode 3 CEA bank withdrawal configurations. This error caused the CEA bank withdrawal analysis to not bound operations in Modes 2 and 3. This error existed for Cycles 10 through 13.

To correct this error, limiting radial and axial peaking factors were identified for potential Mode 3 CEA bank withdrawal configurations, and CEA bank configurations allowed by the Power Dependent insertion Limits in Mode 2. The resulting limiting axial and radial peaking factors were then used in the corrected Rod Withdrawal analysis, which concluded that fuel limits.or DNBR and centerline melt were still met. The safety significance of this error is low, since the fuel limits for DNB and centerline melt protection were still met for Cycles 10 through 13 after correction of the error.

Based on the completion of the revised Uncontrolled CEA Bank Withdrawal event from subcritical analysis, the l

VOP Trip provides fuel protection against exceeding DNBR or fuel centerline melt limits. This analysis includes CEA bank withdrawal events initiated from either Mode 2 or Mode 3 conditions. The revised analysis i

demonstrates that Millstone 2 can continue to operate without the SUR trip. A submittal to the NRC providing the updated analysis is planned. Also in this submittal to the NRC, NU will request NRC concurrence that it is acceptable to continue to operate without the SUR trip. This request will be made, since the original NRC 4

approval for iemoval of the SUR trip prior to Cycle 2 was based on the premise that the SUR trip was solely an equipment protection trip. As discussed above, during Cycle 2 and 3 this was not correct, since C-E was implicitly crediting the SUR trip to prevent CEA Bank Withdrawal events from suberitical from becoming limiting.

IV. Corrective Action

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

1.

Review the Cycle 3 C-E " ground rules" document to determine whether any other issues similar to the SUR trip that could have been improperly understood from Cycle 2 and 3. This corrective action will be complete prior to entry into Mode 4 from the current outage.

Modify procedures to send to the safety analysis supplier (typically the fuel supplier) a copy of all relevant Technical Specification amendments /NRC Safety Evaluation Reports that could reasonably affect the safety analysis, as they are received from the NRC. This will ensure that safety analysis suppliers are kept informed of the final determinations on Technical Specification changes. This corrective action will be complete prior to entry into Mode 4 from the current outage.

3.

Update the Millstone Unit 2 FSAR Section 14.4.1 for the Uncontrolled CEA Bank Withdrawal event from subcritical/ low power to reflect the revised SPC analysis. This corrective action has been completed.

V.

Additional Information

Similar Events LER 93-016: This LER identified the discovery of errors in assumptions for shutdown cooling flow rate used for the boron dilution accident analysis. The boron dilution event was reanalyzed for Modes 4,5 and 6 using a maximum of two operable charging pumps when RCS cold leg temperature is less than 300 degrees Fahrenheit, and increasing the required shutdown margin requirement for operating Mode 5.

NRC FORC A 366A (4-95)

f 1

NRC FOAM 366A U.s. NUCLEAR REGULATORY Commission (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION j

FACILITY NAME (1)

DOCKET LER NUMBER i 6)

PAGE (3)

SEQUENTIAL REVislON YEAR NUMBER NUMBER 5 OF 5 Millstone Nuclear Power Station Unit 2 05000336 l

96

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01 TEXT fit more space is required, use additional copies of NRC Form 366A) (17) l 1

l LER 86-010: This LER identified the discovery of errors in assumptions for the CEA withdrawal from subcritical accident analysis. In particular, the error concerned the number of reactor coolant pumps assumed to be operating in operating Modes 3,4 and 5. The CEA withdrawal from subcritical accident analysis was reanalyzed and a change to the Technical Specifications was approved by i

the NRC in April 1987.

l LER 85-001: This LER identified the discovery of an error in core power distribution (axial shape index) j assumptions for Small Break Loss of Coolant Accident (SBLOCA) analysis. The SBLOCA analysis was reanalyzed with the existing Technical Specification axial shape index imits and acceptable i

results were obtained.

4 LER 83-007: This LER identified the discovery of an error in assumptions in the Steam Generator Tube Rupture 4

l (SGTR) radiological consequences analysis. This error concemed the assumptions for using low steam generator pressure and the atmospheric dump valves in manual mode. The SGTR accident i

was reanalyzed using higher steam generator pressures and the atmospheric dump valve in j

automat.: mode. Acceptable results were obtained.

Enerav Industry identification System (Ells) codes Control Rod Drive Mechanism - AA Reactivity Control Systems - JD-