ML18107A503

LER 99-006-00:on 990729,determined That SG Blowdown RMs Setpoint Was non-conservative.Caused by Inadequate ACs for Incorporating Original Plant Licensing Data Into Plant Procedures.Blowdown Will Be Restricted.With 990826 Ltr

ML18107A503
Person / Time
Site:	Salem
Issue date:	08/26/1999
From:	Bezilla M, Bernard Thomas Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-99-006-02, LR-N990397, NUDOCS 9909030051
Download: ML18107A503 (6)

Text

" PSEG Nuclear LLC A P.O. Box 236, Hancocks Bridge, New J11!1y 08038-0236

~*.

0PSEG

.1\'u clmr LLC AUG 2 6 1999 LR-N990397 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 LER 272/99-006-00 SALEM GENERATING STATION - UNIT 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 Gentlemen: .

This Licensee Event Report entitled "Non-Conservative Setpoint for Steam Generator Slowdown Monitor" is being submitted pursuant to the requirements of the Code of Federal Regulations 10 CFR 50.73(a)(2)(i)(B), any operation or condition prohib!ted by the plant's Technical Specifications.

Sincerely,

1. 2115fty'l M. B. Bezilla Vice President - Operations Attachment C Distribution LER File 3.7 9909030051 990826

~DR ADOCK 05000272 PDR

NRC FORM366 U.S. NUCLEAR REGUL Y COMMISSION APPROVED MB NO. 3150-0104 EXPIRES 06/30/2001 (6-1998) Estimated burden per response to comply with this mandatory information collection request: 50 hrs. Refeorted lessons learned are incorporated into 1.,* LICENSEE EVENT REPORT (LER) the licensing process and ed back to industry. Forward comments regarding burden estimate to the Records Management Branch (T-6 F33),

U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to (See reverse for required number of the Paperwork Reduction Project (3150-0104), Office of Management and digits/characters for each block) Bud~et, Washin~ton, DC 20503. If an information collection does not disp ay a current y valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

IFACILITY NAME m I DOCKET NUMBER ~I IeAGE(~)

TITLE (4)

SALEM UNIT 1 05000272 OF 5 I Non-Conservative Setpoint for Steam Generator Slowdown Radiation Monitor EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR I SEQUENTIAL NUMBER IREVISION NUMBER MONTH DAY YEAR Salem Unit 2 05000311 FACILITY NAME DOCKET NUMBER 07 29 99 99 -006 - 00 08 26 99 HI

• *~ 'Of;' 10 rT<R D* (Ch *~t- '"~ nr

-100, more) (11)

I OPERA TING I THN K. v* *K T.'i; "'rn tMT' rTF.n PTTR.'i;TTA.NT TO THE R 1un K MODE(9) 1 I 20.220l(b) 20.2203(a)(l) 20.2203(a)(2)(v) 20.2203(a)(3)(i) x 50.73(a)(2)(i) 50.73(a)(2)(ii) 50.73(a)(2)(viii) 50.73(a)(2)(x) 20:2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 NWER

,.. 20.2203(a)(2)(ii) 20.2203(a)(4) 50. 73(a)(2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c)(l) 50.73(a)(2)(v)

~l,~ii.iJ:~ >'* 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

Specify in Abstract below or in NRC Fonn 366A LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include Area Code)

Brian J. Thomas, Licensinq Enoineer ( 856) 339-2022

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

COMPONE REPORTABLE REPORTABLE CAUSE SYSTEM NT MANUFACTURER TO EPIX CAUSE SYSTEM COMPONENT MANUFACTURER TO EPIX

<O:TTPPT l<'Mf;'.NT d. T RFPORT F.YPFFTF.O fl d\ MONTH DAY YEAR EXPECTED

'YES (If yes, complete EXPECTED SUBMISSION DATE). IX I No SUBMISSION DATE (15)

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

A safety evaluation in support of a calorimetric calculation change determined the SG blowdown radiation monitors' setpoint was non-conservative. The set point was based on a blowdown flow of approximately 60,000 lbm/hr. The plant has been operated at flows in excess of 60,000 lbm/hr. Upon discovery, SG blowdown flow was restricted to a maximum of 35,000 lbm/hr. Technical Specifications (TS) require, with non-conservative setpoints, discharge of liquid effluents via this pathway be suspended.

The cause of occurrence is inadequate administrative controls for incorporating original plant licensing data and plant testing data into plant procedures. Offsite Dose Calculation Manual (ODCM) effluent radiation monitor assumptions will be reviewed to verify consistency with actual plant operation.

Past operation with increased blowdown flow did not exceed 10CFR20 limits. Therefore, the health and safety of the public was not affected.

This event is reportable in accordance with 10CFR50.73(a)(2)(i)(B), any condition prohibited by the plant's Technical Specifications.

NRC FORM 366A U.S. NUCLEAR R GULATORY COMMISSION

.. (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET (2)

FACILITY NAME (1) NUMBER(2) LER NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER I NUMBER REVISION SALEM UNIT 1 05000272 99 0 0 6 00 2 OF 5 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17)

PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Steam Generator Slowdown System/Radiation Monitoring System{Wl/IL}*

• Energy Industry Identification System {EllS} codes and component function identifier codes appear as (SS/CCC)

CONDITIONS PRIOR TO OCCURRENCE At the time of discovery, Salem Unit 1 and Unit 2 were in Mode 1. (Power Operation) at approximately 100% Power.

DESCRIPTION OF OCCURRENCE A review of the Salem Updated Final Safety Analysis Report"(UFSAR) was being performed July 29, 1999, for the development of a 10CFR50.59 safety evaluation to use a Steam Generator (SG) blowdown flow of approximately 110,000 lbm/hr in the calorimetric calculation. This review identified that section 10.4.8.3 of the Salem UFSAR states, " ... it was further assumed that the blowdown rate was 35,000 lbm/hr (maximum rate) at the time the accident occurred." Radiation levels above the SG blowdown radiation monitors' setpoint initiate blowdown line isolation valve closure. The setpoint for the SG blowdown radiation monitors are set in accordance with the Offsite Dose Calculation Manual (ODCM). One of the assumptions in determining the SG blowdown radiation monitor setpoints is the amount of blowdown flow from the steam generators. The ODCM currently uses a value of 120 gallons per minute.(-60,000 lbm/hr) in determining the radiation monitor setpoints. Operating the plant with a blowdown flow in excess of the value used in the ODCM leads to a non-conservative setpoint of the radiation monitor. Technical Specification 3.3.3.8 action a. states:

"with a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable or change the setpoint so it is acceptably conservative."

The above actions of TS 3.3.3.8.a were not complied with when SG blowdown flow was increased to above 120 gallons per minute (- 60,000 lbm/hr). SG blowdown has been operated above the 120 gpm value since the early 1980's. Upon discovery of this issue, SG blowdown flow was restricted to a maximum of 35,000 lbm/hr.

Based on the above, this event is reportable under 10CFR50.73(a)(2)(i)(B), any condition prohibited by the plant's Technical Specifications.

NRC FORM 366A (6-1998)

" NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION

'*' (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET(2)

FACILITY NAME (1) NUMB~(2) LER NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER IREVISION NUMBER SALEM UNIT 1 05000272 99 0 0 6 00 3 OF 5 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17)

ANALYSIS OF OCCURRENCE In response to an NRC question during plant licensing, PSE&G assumed a maximum blowdown flow value of 35,000 pounds per hour in determining the radiological environmental significance of one of the steam generator (SG) blowdown flow pathways. During plant startup testing, the maximum individual steam generator blowdown flow rate was found to be greater than 40,000 pounds per hour.

The response to the NRC question was not revised based on the measured value. Additionally, the maximum value was not incorporated into plant operating procedures.

In 1980, a design change was performed to install the SG blowdown recovery system at Salem Unit 1.

This design change routed the blowdoWn flow to the main condenser instead of the SG blowdown tanks. The Unit 2 SG blowdown recovery system was installed in 1983. The maximum blowdown flowrate via this pathway was not measured until post installation testing was conducted in April 1998 on another steam generator blowdown system design change. The measured value was slightly less

. than 110,000 pounds per hour.

This measured value was not incorporated into the Offsite Dose Calculation Manual (ODCM) nor was the SG blowdown radiation monitor setpoint revised.

CAUSE OF OCCURRENCE The cause of occurrence is inadequate administrative controls for incorporating original plant licensing data and plant testing data into plant documents. The process for development of calculations and documenting/validating assumptions has changed significantly since the initial licensing of Salem Station. Procedures currently require that assumptions that need to be further evaluated are identified in calculations and that actions are undertaken to validate the assumptions used.

PRIOR SIMILAR OCCURRENCES A review of LERs for Salem .Units 1 and 2 back to 1990, and Hope Creek LERs for 1997, 1998 and 1999 identified the following LERs as similar occurrences:

LER 272/90-032-00 identified that the 1R13A and 1R13B (CFCU service water radiation monitors) alarm setpoints were not correct. In the 1984/1985 time frame a design change was performed that replaced 1R13A & B detector crystals. When the design change was implemented, the program did not ensure adequate verification that new design parameters would be incorporated into applicable configuration control documents. Although the calibration and functional procedures were revised to reflect the new alarm setpoint, the Offsite Dose Calculation Manual (ODCM) was not revised to reflect the change in the 1R13A and B setpoint: Subsequently the procedures were revised to agree with the NRC FORM 366A (6-1998)

NRC FORM 366A U.S. NUCLEAR GULATORY COMMISSION I

(6-1998)

't LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET(2)

FACILITY NAME (I) NUMBER(2) LER NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER IREVISION NUMBER SALEM UNIT 1 05000272 99 0 0 6 00 4 OF 5 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17)

PRIOR SIMILAR OCCURRENCES values of the ODCM which contained the incorrect setpoint. One of the corrective actions consisted of performing an audit of the ODCM to verify the validity of the information and setpoints for the RMS channels identified in the ODCM.

As a result of the ODCM review, LER 311/91-011-00 was generated to identify that the Non Radioactive Liquid Waste Discharge Radiation Monitor (2R37) setpoint was not correct. The 2R37 channel alarm and warning setpoints incorporate the maximum Clearwell Pump flowrate. In 1985, a design modification replaced these pumps with higher capacity pumps. The corresponding 2R37 setpoint was not revised to reflect the increased flowrate. As a result of this LER, the design change process administrative procedure was revised to require the assessment of the impact of a design change on the ODCM.

Although the ODCM review identified that the 2R37 setpoint was incorrect, this review did not identify the incorrect assumption for the SG blowdown radiation monitors.

SAFETY CONSEQUENCES AND IMPLICATIONS Upon determination that Steam Generator (SG) blowdown flow was being operated at a value above the assumed value of 35,000 lbm/hr stated in the UFSAR, SG blowdown was restricted to a maximum of 35,000 lbm/hr.

A review of past operation of Salem Units 1 and 2 during periods of steam generator primary to secondary leakage was performed to determine if the non-conservative setpoint for the SG blowdown radiation monitors allowed a release of radioactive material greater than the 10CFR20 requirements.

Based on a review of the semi-annual effluent release reports, previous evaluations, plant chemistry database, radiochemistry logs, plant control room logs and other pertinent information, there are no indications that liquid releases to the environment via the steam generator blowdown system exceeded the MPC values in 10CFR20. Therefore, the non-conservative steam generator blowdown radiation monitor setpoint was not safety significant.

UFSAR section 15.4.4.5 discusses the environmental consequences of a steam generator tube rupture.

This discussion assumes steam is released through the steam generator safety relief valves. The activity in the steam generators is assumed to be released directly to the environment via the safety relief valves. For a steam generator tube rupture event, the amount of steam generator blowdown flow does not alter the consequences.

Therefore, the health and safety of the public was not affected.

NRC FORM 366A (6-1998

NRC FORM 366A U.S. NUCLEAR R GULATORY COMMISSION (6-1998)

't LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET(2)

FACILITY NAME (1) NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER IREVISION NUMBER SALEM UNIT 1 05000272 99 0 0 6 00 5 . OF 5 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17)

CORRECTIVE ACTIONS

1. Until completion of the evaluation to justify an increase of Steam Generator Slowdown flow above 35,000 lbm/hr, blowdown is being restricted to a maximum of 35,000 lbm/hr.

2. An evaluation is being performed to properly assess operation of the plant with SG blowdown flow up to 120,000 lbm/hr. Appropriate changes to the ODCM, radiation monitor setpoints and the UFSAR will be performed. Changes are expected to be complete by October 15, 1999.

3. A review of the assumptions of the remaining effluent radiation monitors contained in the ODCM will be performed. Actual plant process parameters will be reviewed to ensure they are consistent with the assumptions in the ODCM. This review will be completed by February 29, 2000.

NRC