ML18106A219

LER 97-011-00:on 971119,failure to Test Isolation of Sgb & SS on Automatic Start of Afps Was Noted.Caused by Inadequate Incorporation of DBA Mitigation Features Into Plant Tp.Pse&G Committed to Complete GL 96-01 Evaluations by 971231

ML18106A219
Person / Time
Site:	Salem
Issue date:	12/19/1997
From:	Bernard Thomas Public Service Enterprise Group
To:
Shared Package
ML18106A218	List: ML18106A217 ML18106A219
References
GL-96-01, GL-96-1, LER-97-011-02, LER-97-11-2, NUDOCS 9801050381
Download: ML18106A219 (5)

Text

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION A VED BY OMB NO. 3150-0104 (4-951 EXPIRES 04/30/98 ESTllllATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COll.ECTION REQUEST: 60.0 HRS. REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED LICENSEE EVENT REPORT (LER) BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTNATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F331, U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC (See reverse for required number of 20666-0001, AND TO THE PAPERWORK REDUCTION PROJECT 13160-0104). OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC digits/characters for each block) 20603.

FACILITY NAME 111 DOCKET NUMBER 121 PAGE 131 SALEM GENERATING STATION UNIT 1 05000272 1 OF 5 TITLE 141 Failure to Test Isolation of Steam Generator Slowdown and Sampling System on the Automatic Start of tno Ar 1vi1; ....... , C',-- '***-"--- Cr ...........

EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR I SEQUENTIAL NUMBER I REVISION NUMBER MONTH DAY YEAR FACILITY NAME SALEM UNIT 2 05000311 DOCKET NUMBER 11 19 97 97 - 011 - 00 12 19 97 OPERATING N THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR I: (Check one or more) (11)

MODE (9) 20.2201 (b) 20.2203(a)(2)(v) 50. 73(a)(2)(i) 50. 73(a)(2)(viii)

POWER 000 20.2203(a)(1) 20.2203(a)(3)(i) X 50. 73(a)(2)(ii) 50. 73(a)(2)(x)

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Brian J. Thomas, Licensing Engineer 609-339-2022 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER "l:.":=' iJ,=t-1 _c_Au_s_E --t-SY_s_TEM_-t-_co_M_PO_N_ENT_+-M-AN_u_F_Ac_ru_R-ER-+-R-F_o_:r_P:_~s_LE--11 YES SUPPLEMENTAL REPORT EXPECTED (14)

I EXPECTED SUBMISSION MONTH DAY YEAR I (If yes, complete EXPECTED SUBMISSION DATE).

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

DATE (15)

On November 19, 1997r the Salem Technical Specification Surveillance Improvement Project (TSSIP) identified that the Steam Generator blowdown isolation valves (GB4 valves) and the Steam Generator blowdown sampling isolation valves (SS94 valves) were not being functionally tested to verify that these valves would isolate on an automatic start of the Auxiliary Feedwater (AFW) pumps.

This concern was identified following a review of the applicable drawings and surveillance test procedures in accordance with the guidelines of NRC Generic Letter (GL) 96-01. Isolation of these valves is assumed to occur in the evaluation of the Loss of Normal Feedwater accident.

The cause of this occurrence has been attributed to the inadequate incorporation of design basis accident mitigation features into plant testing procedures. The evaluation of surveillance test procedures in accordance with the guidelines of GL 96-01 is continuing and will be completed by December 31 r 1997.

This event is reportable under 10 CFR 50.73(a)(2)(ii)(A) for the past operation of Salem in an unanalyzed condition that significantly compromised plant safety.

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1050381 971219 ADOC~ 05000272 PDR

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-951 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1 I DOCKET NUMBER (2) LER NUMBER 161 PAGE (3)

SALEM GENERATING STATION UNIT 1 05000272 VEAR I SEQUENTIAL NUMBER I REVISION NUMBBI 2 OF 5 97 - 011 - 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Auxiliary Feedwater System (AFW) {BA/-}

Steam Generator Slowdown {WI/-}

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

CONDITIONS PRIOR TO OCCURRENCE At the time of identification, Salem Unit 1 was defueled and Salem Unit 2 was in Mode 1 at 100%

power. Since Salem Unit 1 was defueled at the time of identification, there was no immediate operability impact. However, since Salem Unit 2 was operating at 100%, compensatory actions were taken as described below until testing of the isolation function could be completed.

DESCRIPTION OF OCCURRENCE On November 19, 1997, the Salem Technical Specification Surveillance Improvement Project (TSSIP) identified that the Steam Generator blowdown isolation valves (GB4 valves) {WI/-} and the Steam Generator blowdown sampling isolation valves (SS94 valves) were not being functionally tested to verify that these valves would isolate on an automatic start of the Auxiliary Feedwater (AFW) pumps {BA/-}. This concern was identified following a review of the applicable drawings and surveillance test procedures in accordance with the guidelines of NRC Generic Letter (GL) 96-01.

The review of the AFW pump start circuitry by TSSIP, verified that a portion of the circuitry is verified to be operational in accordance with surveillance test procedures. However, there was no testing being performed to demonstrate that the GB4 valves and SS94 valves would isolate when an AFW pump started.

The Updated Final Safety Analysis Report (UFSAR) section 10.4.7.2.2 states in part that, "when either of these pumps (AFW) are started automatically, a signal is sent to close Steam Generator blowdown and sampling systems' isolation valves." The review of the AFW pump circuitry design drawings identified that the actual circuitry did exist in the plant to perform this function. However, since this function was not being tested, this feature of the system could not be relied upon to perform its function. As a result, a review of the accident analyses was performed to determine the impact to plant operation.

NRC FORM 366A (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION 14-96)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 111 DOCKET NUMBER 121 LER NUMBER 161 PAGE 131 SALEM GENERATING STATION UNIT 1 05000272 YEAR I SEQUENTIAL NUMBER I REVISION NUMBBt 3 OF 5 97 - 011 - 00 TEXT (If more space is required, use additional copies of NRC Form 366AI I 1 71 DESCRIPTION OF OCCURRENCE (cont'd)

This review identified that the loss of normal feedwater accident is the most affected by an assumed failure of steam generator isolation. This accident analysis assumes that the only mass and energy exiting the steam generators is through the steam generator safety valves. Consequently, the AFW injected into the steam generators is heated to its saturated temperature and then vaporized before exiting the steam generators. The steam generator blowdown lines are assumed to be isolated during the loss of normal feedwater accident. If the steam generator blowdown lines are not isolated, then the AFW flow entering the steam generators may exit the steam generators through the blowdown lines without being heated to its saturated temperature and without being vaporized. This would significantly reduce the heat removal capability of the AFW injected into the steam generators.

The current loss of normal_ feedwater accident analysis assumes, with the isolation of blowdown, the single failure of the turbine driven AFW pump (TDAFWP). In addition, credit is conservatively not taken for one of the two motor driven AFW pumps (MDAFWP). These assumptions result in an analysis conservatively crediting only one MDAFWP delivering 220 gpm of AFW flow to two steam generators and no AFW flow to the remaining two steam generators.

If steam generator blowdown is not isolated for the steam generators on the start of the AFW pumps, then the flow of water exiting the steam generators via the blowdown system could be conservatively assumed to not contribute to the heat removal of the RCS. At Salem, the normal steam generator blowdown flow is 40,000 lbm/hr (-80 gpm) per steam generator. However, to improve chemistry conditions in the steam generators, the steam generator blowdown flow is periodically increased to the maximum value of 100,000 lbm/hr (-200 gpm) per generator. Assuming that the loss of 11ormal feedwater accident occurred during the period of operation of maximum steam generator blowdown flow and blowdown flow was not isolated (since the design feature was not tested), then the essential net result of AFW flow delivered to the steam generators would be 20 gpm (220 gpm - 200 gpm).

This flow rate was less than the amount of flow required in the current accident analysis to ensure the plant would meet the acceptance criteria of the accident analysis.

At the time this -issue was identified;-Salem-Unit"-2 was--operating-at 100%-power*and Salem Unit 1 was defueled. To ensure Salem Unit 2 would continue to operate within analyzed limits, the following immediate actions were initiated based on engineering judgment. Salem Operations issued Temporary Standing Order (TSO) 97-09, on November 19, stating that Steam Generator blowdown flow will be limited to a maximum of 40,000 lbm/hr.

NRC FORM 366A (4-95) i L_ ___

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION 14-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1 I DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SALEM GENERATING STATION UNIT 1 05000272 YEAR I SEQUENTIAL NUMBER I REVISION MJMBER 4 OF 5 97 - 011 - 00 TEXT (If more space is required, use additional copies of NRC Form 366A) ( 171 DESCRIPTION OF OCCURRENCE (cont'd)

Based on the review of the loss of normal feedwater accident, a determination was made that if the second MDAFWP was taken credit for in the loss of normal feedwater analysis (the TDAFWP would remain the single failure) and steam generator blowdown would be limited to a maximum of 40,000 lbm/hr per steam generator then the Salem Units would remain within their analyzed limits. This determination was made based on the engineering judgment that the heat removal capability of two

=

MDAFWPs providing a net flow of 560 gpm (220 gpm - 80 gpm 140 gpm per steam generator) to four steam generators is in excess of the heat removal capability of one MDAFWP providing a total of 440 gpm to two steam generators. A RETRAN analysis was completed, on December 5, to confirm the conclusions of the engineering judgment concerning the sensitivity of the loss of normal feedwater accident results to net delivered AFW flow rates. The results of this analysis confirm that the plant remains within the conclusions presented in the Updated Final Safety Analysis Report (UFSAR) for the loss of normal feedwater accident.

As a result of the above information, a determination was made that Salem Units 1 and 2 had potentially operated in the past in an unanalyzed condition that could compromise plant safety.

Therefore, this condition is being reported under 10 CFR 50.73(a)(2)(ii)(A).

CAUSE OF OCCURRENCE The cause of this occurrence has been attributed to the inadequate incorporation of design basis accident mitigation features into plant testing procedures.

PRIOR SIMILAR OCCURRENCES A review of LERs for Salem Units 1 and 2 issued in the previous two years did not identify any previous similar occurrences of failure to adequately test design basis accident mitigation features.

However several LERs have been issued as a result of missed surveillances due to inadequate implementation of.Technical-Specification-requirements: *-These-I.:£ Rs are-discussed in LER 272/96-005. The identification of these programmatic issues resulted in the initiation of the Technical Specification Surveillance Improvement Program (TSSIP) described in LER 311/95-008.

NRC FORM 366A (4-95)

NRC FORM 366A (4-96)

• LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME 111 DOCKET NUMBER 121 LER NUMBER 161 PAGE 131 SALEM GENERATING STATION UNIT 1 05000272 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 5 OF 5 97 - 011 - 00 TEXT llf more space is required, use additional copies of NRC Form 366AI 1171 SAFETY CONSEQUENCES AND IMPLICATIONS Failure of the steam generator blowdown and sampling systems to isolate during a loss of normal feedwater accident would lead to a reduction in heat removal capability of the AFW system. In the event that a loss of normal feedwater accident occurred and steam generator blowdown and sampling did not isolate, Operators would be alerted to this situation by an increasing Reactor Coolant System (RCS) temperature. The increasing RCS temperature would alert the Operators to evaluate the heat sink capability and lead the Operators to isolate the steam generator blowdown lines. Subsequent to the identification of this issue, testing was performed on Salem Unit 2 that satisfactorily demonstrated that steam generator blowdown will isolate on the start of an AFW pump.

Based on the above discussion, the health and safety of the public were not impacted.

CORRECTIVE ACTIONS

1. Public Service Electric and Gas (PSE&G) committed to complete the Generic Letter 96-01 evaluations by December 31, 1997, as stated in letter LR-N970346 dated May 28, 1997.

2. Temporary Standing Order (TSO) 97-09 was issued on November 19, 1997, to limit steam generator blowdown flow to 40,000 lbm/hr. This TSO remained in effect until the testing, discussed in corrective action 3, was satisfactorily completed.

3. Procedure TS2.SE-SU.AFW-0001 (Q), Rev. 0, was issued on November 23, 1997 for the testing of the Unit 2 steam generator blowdown and sampling system isolation on the start of the AFW pumps. Testing was satisfactorily performed on November 24, 1997.

4. Surveillance test procedures S1 (2).0P-ST.AF-0009(Q) were revised on December 5, 1997 to include the necessary steps to test the isolation of steam generator blowdown and sampling following the automatic start of the AFW pumps.

NRC FORM 366A (4-95)