05000336/LER-2003-001, Re Pressurizer Water Volume Periodically Exceeded the Technical Specifications Limit

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Re Pressurizer Water Volume Periodically Exceeded the Technical Specifications Limit
ML031140346
Person / Time
Site: Millstone Dominion icon.png
Issue date: 04/14/2003
From: Sarver S
Dominion Nuclear Connecticut
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
B18865 LER 03-001-00
Download: ML031140346 (4)


LER-2003-001, Re Pressurizer Water Volume Periodically Exceeded the Technical Specifications Limit
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications

10 CFR 50.73(a)(2)

10 CFR 50.73(a)(2)(ix)(A)

10 CFR 50.73(a)(2)(iii)

10 CFR 50.73(a)(2)(x)

10 CFR 50.73(a)(2)(iv)(A), System Actuation

10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor

10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat

10 CFR 50.73(a)(2)(v), Loss of Safety Function

10 CFR 50.73(a)(2)(1)
3362003001R00 - NRC Website

text

Dominion Nuclear Connecticut, Inc.

Dominion-Millstone Power Station I

Rope Ferry Road Waterford, CT 06385 APR I 4 2003 Docket No. 50-336 B18865 RE: 10 CFR 50.73(a)(2)(i)(B)

U.S. Nuclear Regulatory Cornmission Attention: Document Control Desk Washington, DC 20555 Millstone Power Station, Unit No. 2 Licensee Event Report 2003-001-00 Pressurizer Water Volume Periodically Exceeded The Technical Specifications Limit This letter forwards Licensee Event'Report (LER) 2003-001-00, which documents an event at Millstone Power Station, Unit No. 2, on February 17, 2003. This LER is being submitted pursuant to 10 CFR 50.73(a)(2)(i)(B) as any operation or condition prohibited by the plant's Technical Specifications.

There are no regulatory commitments contained within this letter.

Should you have any questions regarding this submittal, please contact Mr. Paul Willoughby at (860) 447-1791, extension 3655.

Very truly yours, DOMINION NUCLEAR CONNECTICUT, INC.

Stephen P/arver, Director Nuclear Station Operations and Maintenance Attachment (1):

LER 2003-001 -00 cc:

H. J. Miller, Region I Administrator R. B. Ennis, NRC Senior Project Manager, Millstone Unit No. 2 Millstone Senior Resident Inspector

I NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004

, the NRC may not conduct or sponsor, (See reverse for reouired number of dioitslcharacters for each block) and a person Is not required to respond to, the information collection.

FACILITY NAME (1)

DOCKET NUMBER (2)

PAGE (3)

Millstone Power Station - Unit No. 2 05000336 1 OF 3

TITLE (4)

Pressurizer Water Volume Periodically Exceeded The Technical Specifications Limit EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

MO DAY YEAR YEAR SEQUENTIAL REV MO DAY YEAR FACILITY NAME DOCKET NUMBER Z jNUMBER NO.

I l

05000 02 17 2003 2003 - 001 - 00 04 14 2003 FACILITY NAME l DOCKET NUMBER 1

05000 OPERATING 1

THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) (11)

MODE (9) 20.2201 (b) 202203(a)(3)C(i) 50.73(a)(2)(il)(B) 50.73(a)(2)(ix)(A)

POWER 100 20.2201(d) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x)

LEVEL (10)

= 20 2203(a)(1)

= 50 36(c)(116)(Al

=

50.73(a)(2)(iv)(A) 73.71 (a)(4) 20.2203(a)(2)(i) 50 36(c)(1)(ii)(A) 50.73(a)(2)(v)(A)

_ 73.71 (a)(5) 20.2203(a)(2)Q(i) 50 36(c)(2) 50.73(a)(2)(v)(B)

_ OTHER

- q_

20.2203(a)(2)(iii) 150.46(a)(3)(ii) 50.73(a)(2)(v)(C)

Specify in Abstract below or

>^A; 5

20 2203(a)(2)(v) 50.73(a)(2)(1)(A) 50.73(a)(2)(v)(D)

In (If more space Is required, use addibonal copies of NRC Form 366A) (17) relief valves following any analyzed moderate frequency event in FSAR Chapter 14. Liquid discharge out the pressurizer relief valves could lead to a more serious plant condition due to the potential for the loss of Reactor Coolant System (RCS) [AB] fission product barrier integrity.

The limiting moderate frequency events with respect to pressurizer overfill are the loss of external load event presented in Final Safety Analysis Report (FSAR) Section 14.2.1 and the loss of normal feedwater flow event presented in FSAR Section 14.2.7. The FSAR Section 14.2.1 loss of external load event transient results in approximately a 10 percent increase in pressurizer level. The FSAR section 14.2.7 loss of normal feedwater transient results in approximately 5 percent increase in pressurizer level. Operating the plant due to the condition described above, with an actual pressurizer level approximately 3 percent higher than the maximum value allowed by the Technical Specifications, will not result in pressurizer overfill or the potential for the loss of the RCS fission product barrier integrity for either of these events.

The initial pressurizer level does not have a significant impact on the calculated peak RCS pressure for the limiting moderate frequency event with respect to RCS pressurization, the loss of external load event. The maximum calculated RCS pressure for the loss of external load event is 2717 psia, and is limited by the opening set pressure and relief capacity of the pressurizer safety valves. Operating the plant due to the condition described above, with an actual pressurizer level approximately 3 percent higher than the maximum value allowed by the Technical Specifications, will not result in a significant increase in the peak RCS pressure for this event.

Additionally, operating the plant due to the condition described above, with the actual pressurizer level approximately 3 percent higher than the maximum value allowed by the Technical Specifications, would result in a slight increase in the maximum liquid mass inventory contained within the RCS. The impact of this increased RCS liquid mass will not result in a significant increase in the calculated containment pressure consequences of the main steam line break or loss of coolant accident.

Based upon the above discussion this event has a low safety significance.

4. Corrective Action

The Engineering documents were revised to reflect the correct differential pressure calibration values for the pressurizer level transmitters. The results were subsequently incorporated into the surveillance procedure and on February 24, 2003 the two level transmitters were recalibrated to the correct values.

5. Previous Occurrences

No previous similar events were identified.

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