IR 05000317/1993031
| ML20059D652 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/27/1993 |
| From: | Eric Benner, Harris P, Lyons C, Larry Nicholson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20059D621 | List: |
| References | |
| 50-317-93-31, 50-318-93-31, NUDOCS 9401100029 | |
| Download: ML20059D652 (25) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
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l Report Nos.
50-317/93-31; 50-318/93-31 License Nos.
DPR-53/DPR-69 (
Licensee:
Baltimore Gas and Electric Company Post Office Box 1475 Baltimore, Maryland 21203 Facility:
Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Location:
Lusby, Maryland Inspection conducted:
November 15, 1993, through November 19, 1993 Inspectors:
Carl F. Lyon, Resident Inspector Paul W. Harris, Resident Inspector, Vermont Yankee Eric J. Benner, Reactor Systems Engineer, NRR
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Approved by:
Carry [)$. Nichofson, Chief Date Reactor Projects Section No. l A Division of Reactor Projects Inspection Summary:
This inspection report documents the special inspection of BG&E's response to an excessive service water system leak that occurred on September 29,1993, and BG&E's actions regarding a potential concern with closed loop safety-related cooling systems that are replenished from non-safety-related makeup systems. This report includes a description of
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the excessive leak event and the affected systems, a review of the safety significance, and
assessments of BG&E's corrective actions with respect to the issue and current system
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operations. An Executive Summary follows.
I 9401100029 931227 PDR ADOCK 05000317 G
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l EXECUTIVE SUMMARY
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i Calvert Cliffs Nuclear Power Plant. Units 1 and 2 l
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Inspection Repod Nos. 50-317/93-31 and 50-318/93-31
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On September 29,1993, BG&E removed the 11 service water (SRW) system from service l
following the discovery of excessive system leakage. When initially questioned by
l inspectors, operators had no guidance on the size of leakrate required to consider the system
inoperable. Inspectors found that in November 1991 BG&E had identified a concern that the
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i operating instructions for the SRW system and the component cooling water (CCW) system provided no limits for maximum system leakage. Since the makeup water sources to the systems were not seismically qualified or safety-related, existing system leakage under certain
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t accident conditions could disable the systems and prevent them from performing their design
safety function.
j BG&E evaluated the condition in November 1992 and concluded that the SRW system required a safety-related makeup source to ensure operability and that operability limits should be established based on the capacity of safety-related makeup. BG&E concluded that i
there was no operability concern with the CCW system because of historically low system leakrates.
i Equipment was staged and operators were given informal instruction and training on
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establishig a saltwater-SRW cross-connect to provide safety-related makeup in extreme
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emergenci9s. However, in September 1993, operators were still unaware of leakrate
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operability criteria for the SRW system, and the saltwater-SRW cross-connect procedure had l
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not been put into the operating instructions. These deficiencies were not corrected until October 1993.
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Inspectors assessed that continued operation of the SRW and CCW systems with the existing makeup water sources was safe, because adequate guidance was provided to direct operator
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actions based on escalating leakrates, periodic sampling was performed by chemistry and
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operations personnel to presage deteriorating system conditions, historical system performance was generally good, and contingency plans were provided for various makeup sources to assure defense-in-depth.
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While BG&E's identification of the concern over long-term operability of the SRW and CCW systems without a safety-related makeup source was commendable, the failure to promptly resolve the condition was a violation of NRC requirements. In addition, BG&E's basis for the operability assessments and the operability criteria for the systems did not
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include potential leakage through some boundary isolation valves. This was an unresolved
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item.
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Executive Summary Additional conclusions were summarized as follows. The operability screening criteria and procedures were satisfactory; however, they were not effectively implemented for this issue.
The 10 CFR 50.59 safety evaluation for cross-connecting saltwater to service water was satisfactory. The SRW instructions were inadequate prior to having operability guidance based on system leakrate and contingincies to provide makeup from the saltwater system.
BG&E's response to the excessive SRW leak in September 1993 was satisfactory; however, some weaknesses in timeliness and interdepartmental communications were noted.
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J TABLE OF CONTENTS Page 1.0 BACKGROUND S UMMARY................................ 1 2.0 PURPOSE......s.....................................
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3.0 SYSTEM DESCRIPTIONS.................................. 2
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4.0 ABBREVIATED SEQUENCE OF EVENTS....................... 3 i
5.0 CONCLUS ION S........................................ 7 5.1 The Operation of the SRW and CCW Systems with the Existing Makeup Water Sources is Safe........................... 7
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5.2 The Procedures and Criteria for Operability Screening were l
Satisfactory; however, They were not Effectively Implemented for this Issue........................................... 9 5.3 The 10 CFR 50.59 Safety Evaluation for Cross-Connecting Saltwater to
SRW was Satisfactory
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5.4 The SRW Operating Instructions were Inadequate Prior to having the Current Ixakrate Guidance and Contingency Plans in Place I1
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5.5 The CCW Operating Instructions were Satisfactory..............
I1 5.6 BG&E's Response to the Excessive SRW Leak on September 29 was Satisfactory; however, Some Weaknesses were Noted............
5.7 The Corrective Action Process for this Issue Was Weak...........
i 6.0 BG&E RESPONSE
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7.0 FOLLOW-UP OF PREVIOUS INSPECTION FINDINGS..............
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7.1 (Closed) Unresolved Item 50-317 and 318/93-28-01
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8.0 MANAGEMENT MEETING
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8.1 Preliminary Inspection Findings
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i ATTACHMENT 1: SERVICE WATER SYSTEM DIAGRAM l
A'ITACHMENT 2: COMPONENT COOLING WATER SYSTEM DIAGRAM i
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A'ITACHMENT 3: PERSONNEL CONTACTED DURING THE INSPECTION iv
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DETAILS 1.0 BACKGROUND SUMMARY
On September 29,1993, BG&E removed the 11 service water (SRW) system from service following the discovery of excessive system leakage. The leakrate was about 24 gpm, whereas normal system leakage was less than 2 gpm. When initially questioned by inspectors, operators had no guidance on the size ofleakrate required to consider the system inoperable. BG&E's investigation later found that four tubes in the 11 SRW heat exchanger were leaking.
The SRW system is a closed loop safety-related system consisting of two separate safety-
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related subsystems. Each subsystem has its own head tank and can utilize any one of three safety-related pumps. The SRW system transfers heat from the main turbine generator (a non-safety-related function), the emergency diesel generators (EDGs), the containment air coolers, and the spent fuel pool heat exchangers to the open loop saltwater cooling system.
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The SRW system is required to remain operable following a plant event to provide cooling to the EDGs. According to the Updated Final Safety Analysis Report (UFSAR), the duration of an event may be as long as 30 days.
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Normally, makeup water is supplied to the SRW head tanks by a hard-piped path from the
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demineralized water and condensate systems. This makeup flow path is not seismically qualified. The pumps for the makeup flow path are not powered from the EDGs. The scenario of concern in this case is a seismic event causing a loss of offsite power. For this scenario, the hard-piped makeup to the SRW head tanks are assumed to be not available.
The vulnerability to a loss of offsite power was previously recognized by BG&E and compensated for by development of procedures to connect the head tanks to the fire
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protection header, which is supplied by a diesel-driven fire pump, using a fire hose. This measure effectively compensated for a lack of EDG backup to the primary makeup path.
i However, since the fire protection system is not seismically qualified, it could not be
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assumed to survive the postulated scenario.
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Inspectors found that in November 1991 BG&E had identified an issue that the operating instructions for the SRW system and the component cooling water (CCW) system provided no limits for maximum system leakage. The issue report noted that during an event
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insolving a loss of offsite power and a loss of power to the non-vital buses, there would be
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no automatic makeup to the SRW and CCW head tanks. The issue report recommended that the maximum amount of SRW and CCW leakage be limited to that amount sufficient to give operators time to initiate makeup by manually cross-connecting the fire protection header
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before the head tanks empty. BG&E calculated in November 1992 that with system leakage of 2 gpm and no makeup, net positive suction head to the SRW pumps could be lost in 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. A leakrate of 17 gpm with no makeup could result in loss of the SRW system in e
about 1-1/2 hours. As a contingency, BG&E planned to use makeup from the safety-related l
saltwater system to the SRW system in extreme emergencies. Nevertheless, in September 1993, operators were still unaware of a leakrate operability criteria for the SRW system.
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Inspectors initially reviewed the excessive SRW leak event in NRC Inspection Report 50-317 and 318/93-28. They identified apparent weaknesses in BG&E's corrective action process and interdepartmental communications, and potential system design deficiencies. These issues were an Unresolved Item (93-28-01) at the end of that report period.
2.0 PURPOSE The pmpose of this special inspection was to review BG&E's actions following identification of the potential concern regarding leakage from the closed loop safety-related systems without a safety-related makeup source, to assess whether there was adequate conGdence that the SRW and CCW systems could perform their design functions as required during certain plant events, and to compare the licensing and design requirements with the as-built SRW and CCW systems. In addition, inspectors reviewed BG&E's response to the excessive leak from the 11 SRW system on September 29,1993.
The review included partial walkdowns of the SRW and CCW systems, including pre-staged equipment for makeup water contingency plans; interviews of operations and engineering personnel involved in the excessive SRW leak on September 29, including supervisors; review of applicable technical specifications, procedures, instructions, ar.d event documentation; review of applicable corrective action process instruons and documentation; uv 7w of applicable etgineering' evaluations and bFSAR sections; and interviews of asso..ated engineering, licensing, and operations personnel.
3.0 SYSTEM DESCRWflONS The SRW system is a chemically treated fresh water, closed loop system whose purpose is to remove heat from turbine plant components, the containment air coolers, the emergency diesel generators, and the spent fuel pool, and transfer that heat to the saltwater system. The SRW system functions as one system in the Turbine Building and as two subsystems in the Auxiliary Building.
During normal plant operations, both Auxiliary Building subsystems are required to be in operation and function independently to the degree necessary to assure the safe operation and shutdown of the plant in the event of a component failure. Each subsystem includes a head tank, an electric-driven pump, and a SRW/ saltwater heat exchanger. A third SRW pump is
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provided as backup and may be electrically and mechanically cross-connected to supply either subsystem.
Two head tanks, one for each subsystem, provide the required net positive suction head for their respective subsystems. Head tank level is automatically maintained by level controllers which can receive makeup water from either the demineralized water or condensate systems.
Head tank level is indicated locally by sightglass, and electrically in the Control Room.
Attachment 1 is a simplified SRW system diagram.
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The CCW system functions as an intermediate cooling system and radiological barrier between certain reactor plant / aux.iliary system components and the saltwater system. It is a nonradioactive, chemically treated fresh water system. Some of the components cooled by l
CCW are the shutdown cooling heat exchangers; the letdown heat exchanger; the mechanical
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seal cooler, lube oil cooler, and thermal barrier for each reactor coolant pump; the control element drive mechanism coolers; the mechanical seal cooler, stuffing box jackets, and bearing housings for each high pressure safety injection (HPSI) pump; and the mechanical seal cooler, bearing house, and stuffing box jacket for each low pressure safety injection (LPSI) pump.
The CCW system will normally function to provide heat removal from the reactor plant during cooldown below 300 F (shutdown cooling), and provide heat removal from the
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reactor plant for long-term cooling and cooling to containment spray following a loss of
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coolant accident. During normal operation, only one CCW pump is required for component l
cooling circulation. This pump takes a suction on two separate return headers and discharges l
to two discharge headers. The water then flows through the inservice heat exchanger to the supply headers, branching off in parallel paths to the components to be cooled, and then to the return headers.
The component cooling head tank maintains sufficient net positive suction head to the l
component cooling pumps. The head tank also provides a surge volume, allowing for l
expansion and contraction of the system without interruption of normal operations. Makeup
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water for the system is added to the head tank automatically from either the demineralized water or condensate systems based on head tank level. The head tank is unbaffled.
Attachment 2 is a simplified CCW system diagram.
4.0 ABBREVIATED SEQUENCE OF EVENTS 11/25/91:
BG&E issue report (IR) IRO-013-020 identified the concern that operating instructions (OIs) for the SRW and CCW systems provided no limits for maximum system leakage. The only criteria given in the Ols was to notify Plant Chemistry if leakage exceeded 2 gpm. The IR concluded that the i
maximum amount of SRW and CCW leakage must be limited to that amount that would give operators sufficient time to initiate makeup from the fire protection system before the head tanks emptied.
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IR5-011-534 identified that the long-term operability of the SRW system following an accident involving loss of offsite power was questionable, because l
there were no controls over allowable leakage and no makeup capability from a safety-related system. The IR noted that the system OI allowed continued operation without an operability assessment regardless of leakage rate, and that a leak rate of 16 gpm could empty a head tank in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The IR recommended that the requirements for long-term operation of the SRW l
system following accidents concurrent with a loss of offsite power be l
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determined, that leakage limits compatible with long-term operation be established, that operability limits be established, and that consideration be given to providing safety-related makeup. It also recommended that the CCW system receive similar evaluation and controls. The IR was initiated by the same employee who had written IRO-013-020, because he had seen no response to the first IR.
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IR5-011-534 was assessed by the Issue Report Review Group to be a
"significant condition adverse to quality," and was upgraded to a Program Deficiency Report (PDR). PDR 92488 was issued. IRO413-020 was closed to the PDR on 11/16/92.
11/11/92:
CCW and SRW system maximum leakage rate calculations M-92-234 and 235 were completed by BG&E Design Engineering Section (DES).
I1/17/92:
Initial operability determinations done by DES on the SRW system for PDR 92488 concluded that no immediate operability concern existed, partly because operator action could be taken to establish a makeap source.
11/20/92:
General Supervisor-Plant Engineering Section (GS-1 ES) memo to General Supervisor-Nuclear Plant Operations (GS-NPO) noted that DES questioned the ability of the fire protection system to provide makeup to SRW after a seismic event. The memo noted that DES, Operations, Compliance, and PES agreed that there was no immediate operability question, but that the appropriate short-term compensatory measure was to make arrangements to cross-connect saltwater to SRW. The memo included several actions to address the compensatory measure based on a 17.4 gpm leakrate, which would allow 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to install the cross-connect. An attachment to the memo detailed the necessary equipment, locations, and installation procedure to install the cross-connect. The memo noted that the process would be proceduralized into the SRW operating instruction, and that the leakrate determination procedure would be modified to provide additional actions at 7 gpm and 15 gpm.
Tools and 3/4 inch hoses were pre-staged for the saltwater (SW) to SRW cross-connect. The on-duty operations crew was briefed on the cross-connect procedure. Subsequent crews were briefed as they came on duty. The memo was provided to the Shift Supervisor for information. Procedure change
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requests were prepared to add the additional actions to the leakrate determinations and to add the cross-connect method to the operating instruction.
11/25/92:
BG&E memo CCSO-92-L85, prepared by the Calvert Cliffs Site Office of Combustion Engineering via DES, evaluated the operability of the SRW system with a known outleakage. It stated that "...the operability of a SRW
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system is based, in part, on the ability to add water to the system to makeup for out-leakage" and "...[a] makeup source that is seismically qualified is required to ensure that the SRW system is operable." The memo concluded i
that, "[as] it pertains to system out-leakage, the SRW System is operable as long the following is true: (1) Currently, the system out-leakage is equal to or less than 8.5 gpm, as measured weekly by the appropriate procedure (s) [ based on the capacity of the 3/4 inch hoses]. When the hose:: are replaced with 1-1/4 inch diameter hoses at 75 foot lengths, the system out-leakage is limited to less than or equal to 16.3 gpm. (2) A Standing Order is in place outlining the method of cross-connecting the SRW and SW systems, until the procedural changes occur. (3) The equipment necessary to cross-connect the SRW and SW systems is pre-staged in or near the SRW Room area." Inspectors noted that no " Standing Order" was ever put in place, but the 11/20/92 memo was provided to the shift supervisors.
12/15/92:
PDR 92488 was determined to not be a significant issue by the Quality Audits
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Unit (QAU) Review Team, and was downgraded to IR status (IR0-009-294).
The team concluded that it was a "what if" situation rather than an actual problem. They concluded that the makeup system was not required for SRW operability, otherwise it would have been safety-related.
2/12/93:
An initial operability assessment done by DES for the CCW system concluded there was no operability issue because there was no leakage from the system.
3/1/93:
IR5-Oll-534 was brought before the Plant Operations and Safety Review Committee (POSRC) at the request of the IR initiator. No safety concerns were noted, but the Committee requested additional information regarding why excessive leakage was not an issue for the CCW system.
3/9/93:
Operating instruction (01)-15, "SRW System," leakrate determination was changed to include required actions at 2 gpm,7 gpm, and 15 gpm leakrates.
3/24/93:
BG&E memo CCSO-93-557 provided a design evaluation of the SRW and CCW systems in response to PDR 92488. It noted that, "...a source of makeup water is required at all times to ensure SRW subsystem operability,"
and "... operability of a SRW system is based, in part, on the ability to add water from a seismically qualified makeup source in order to compensate for the out-leakage of the system." The memo reiterated the conclusions of BG&E memo CCSO-92-L85 of 11/25/92 regarding SRW operability. It also concluded that CCW system out-leakage was very low and there was no immediate concern for the operability of the CCW system.
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3/29/93:
The POSRC recommended approval of 10 CFR 50.59 Safety Evaluation 92-B-Oll-186-R00, which provided for installation of a cross-connection (by temporary hoses) between the saltwater and SRW systems.
5/12/93:
DES presented POSRC the basis for the conclusion that excessive leakage was not a safety issue for the CCW system. The conclusion was based on the considerations that loss of CCW following a loss of coolant accident (LOCA)
does not result in failure to meet any critical safety function, that system leakage has historically been very low, and that a monthly leakage determination is performed to verify leakage is less than 2 gpm. At that leakrate, the head tank had sufficient volume to support system operation for about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. POSRC noted that the technical specifications required CCW to be operable, and nothing currently required an operability evaluation be performed if system leakrate increased. The Superintendent-Nuclear Operations stated that the leakrate determination would be changed to require an operability evaluation ifleakrate exceeded 2 gpm.
8/13/93:
IRO-0167-929 documented that the leakrate for 11 SRW header was 2.35 gpm and for 12 SRW header was 3.52 gpm. The IR noted that SRW heat exchanger tube leaks may have been responsible.
9/93:
1-1/4 inch diameter hoses were pre-staged at the SRW pump room.
9/23/93:
Unit 1 SRW 1eakrate was 14.1 gpm. Chemistry and system engineering personnel were notified as required by 01-15. An IR was not written, because one had been written in August. System engineering and operations personnel noted that maintenance was scheduled for bulleting / plugging 11 SRW heat exchanger tubes on 9/30.
9/29/93:
Between about midnight and 6:00 a.m., while performing surveillance test procedure (STP) 0-70-1, " Monthly Test of "A" Train Containment Air Coolers (CACs)...," operators noted that the frequency of pumping the containment sump increased from about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to about 45 minutes. After confirming that reactor coolant system leakrate had not changed, they postulated that one of the SRW relief valves for the 11 or 12 CAC may have lifted and failed to reseat during the STP, or that a SRW constant vent may have started leaking by. 11 and 12 CACs were isolated one at a time. SRW leakrate was measured and found to be about 24 gpm and unaffected by CAC isolation. After cycling the CAC SRW emergency outlet valves, the containment sump frequency returned to normal, but the SRW leakrate remained high. Operators surmised that the CAC SRW relief valves had rescated, but SRW was leaking from another source. After a system check for potential leakage paths, operators concluded that the most likely source was the 11 SRW heat exchanger tubes, particularly since they were aware that tube
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plugging was scheduled on that heat exchanger for 9/30 due to suspected tube leaks.
9/29/93:
At about 6:00 a.m., the GS-NPO became aware of the problem during his morning tour. After discussing it with the Shift Supervisor, they decided to isolate the 11 SRW heat exchanger after shift turnover to try to verify the leak source. Coincidentally, the principal auxiliary systems engineer entered the control room on his morning tour and was informed of the high SRW leakrate.
The NRC resident inspector was also made aware of the high leakrate during his morning tour. He questioned the operability of the system, and was told by the GS-NPO that no leakrate operability criteria had been provided to operators by the engineering staff.
9/29/93:
At 7:10 a.m., the 11 SRW system was taken out of service to inspect the SRW heat exchanger, and the applicable technical specification action statements were entered.
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9/29/93:
At about 9:30 a.m., following discussion between the principal auxiliary systems engineer and the GS-PES, PES recommended the 11 SRW system be declared inoperable because the leakrate exceeded 16.3 gpm. However, the system had already been declared inoperable when operators removed it from service.
9/29/93:
With the 11 SRW heat exchanger isolated, SRW leakrate was about 2.3 gpm.
Four leaking tubes were plugged and the system was returned to service at 5:15 p.m. Issue reports were written to document the concern that the CAC SRW relief valves or const t vents may have lifted during the STP, and to document the tube leaks it ne 11 SRW heat exchanger.
10/7/93:
The saltwater to SRW cross-connect procedure was added to 01-15.
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10/14/93:
The 16.3 gpm operability criteria was added to 01-15.
11/12/93:
A requirement was added to 01-16, "CCW System," to notify system engineering and to write an issue report with an operability concern if CCW j
leakage exceeded 2 gpm.
5.0 CONCLUSIONS 5.1 The Operation of the SRW and CCW Systems with the Existine Makeuo Water
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Sources is Safe i
Inspectors found that the SRW and CCW systen s were operated in accordance with the current licensing basis; however, the basis was silent with regard to makeup sources
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following an accident. Whereas a similar system in a newer plant may be required to have a head tank with a capacity based on some assumed system leakrate, or be required to have a seismically qualified makeup source, these requirements were not in the Calvert Cliffs design
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basis. The purpose of the head tanks in the Calvert Cliffs SRW and CCW s.vstems, as stated in the UFSAR, was to allow for thermal expansion of the system watcr wj to provide net positive suction head for the system pumps. The makeup source to both systems was noted to be demineralized water.
The SRW system had four proceduralized makeup sources. The demineralized water or the condensate system normally provided automatic makeup to the head tanks. These systems were not seismically qualified and the makeup pumps were not powered from vital buses.
The fire protection header could be manually cross-connected via hose to supply makeup to SRW. It also was not seismically qualified, but was considered by BG&E to be " rugged,"
and capable of surviving a seismic event. Inspectors identified no seismic concerns or access restrictions during walkdowns of the system. BG&E's assessment that the system would remain operable following a seismic event appeared to be based on reasonable engineering judgment. The system was supplied from a diesel-driven pump. In addition, BG&E had proceduralized a cross-connect via hose from the safety-related saltwater system to SRW.
Based on a review of trend data for the past four years, SRW system leakage was generally less than 2 gpm. However, historical data revealed numerous and sustained periods of higher leakage which could have significantly affected system performance, assuming loss of makeup and no operator action. For one week in March 1989, Unit 2 SRW leakage was about 18 gpm. On some occasions for both units, leakage exceeded 6 gpm for periods of many weeks. This was significant because makeup to the head tanks was automatically controlled by a level control valve, so system leakage could be masked from operators.
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Nevertheless, inspectors found that periodic system leakrate determinations were performed by the operations unit and the chemistry section at a sufficient frequency to allow assessment and corrective action for increased leakage. Operations personnel perform a weekly SRW system leakage determination. Chemistry personnel generally sample SRW at least three times per week to determine hydrazine concentrations, but may sample several times daily if hydrazine concentration is changing. Inspectors reviewed chemistry trend graphs and interviewed chemistry supervisors and found that changes in hydrazine concentration correlated well with system leakrate. Chemists and operators noted several past occasions where reduced hydrazine level was the first indication of an abnormal SRW leak. Inspectors found that Plant Chemistry's estimation ofleakrates based on hydrazine correlated well with
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actual measured leakrates.
Control room indications of important system parameters, such a head tank level and system pressure, were provided for system assessment by operators. Operators were knowledgeable of makeup sources, including implementation of manual cross-connects to the fire protection I
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header or saltwater system. Adequate guidance was provided in the operating instructions for operator actions and engineering evaluations based on escalating leakrates, including operability limits.
Inspectors found that preventive maintenance on the SRW system was satisfactory to assure system reliability and continued operability. Heat exchanger performance was monitored and tube plugging criteria were established commensurate with industry guidance. Tubes were checked for leakage generally every quarter as part of regular heat exchanger cleaning, and leaking tubes were plugged. Housekeeping and material condition of the system was good.
Inspectors found similar circumstances in the CCW system, with some exceptions. The saltwater system was not an approved makeup source to CCW. Operability limits are not included in the opera %g instructions for CCW, but guidance did direct operator action and evaluation for a leaktate above 2 gpm. BG&E was evaluating the need for operability criteria in the CCW operating instruction. The CCW system leakrate was historically zero.
l Review of trend data revealed that on rare occasions the leakrate was 0.1 gpm, and once got to 0.3 gpm. Operations personnel performed a monthly leakrate determination on CCW, but Plant Chemistry sample frequency was similar to the SRW system. Inspectors also noted that, while the SRW system was required to support EDG operations, the CCW system was not required to safely shut down the plant.
Continued operation of the SRW and CCW systems with the existing makeup water sources was assessed by the inspectors to be safe. The conclusion was supported primarily by the fm' dings that a safety-related makeup source was now available to the SRW system, that operability guidance based on the capacity of the safety-related makeup source was provided to operators, and that de CCW system was not required for safe plant shutdown. In addition, adequate guidance was provided to direct operator actions based on escalating leakrates, periodic sampling was performed by chemistry and operations personnel to presage deteriorating system conditions, historical system performance was generally good, and contingency plans were provided for various makeup sources to assure defense-in-depth.
j 5.2 The Procedures and Criteria for Operability Screening were Satisfactory: however.
They were not Effectivelv Implemented for this Issue The inspectors reviewed procedures QL-2-100, " Issue Reporting and Assessment," QL-3-102, " Program Deficiency Reporting," Calvert Cliffs Instruction (CCI)-315, " Functional Evaluation / Operability Determination," and Issues Assessment Unit (IAU) Instruction -02,
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" Issue Report Processing by the IAU," and concluded that satisfactory instructions were provided to assure prompt and comprehensive operability determinations. The procedural guidance was detailed and clearly documented, and included both industry and regulatory guidance. NRC C -ic letter 91-18, " Resolution of Degraded and Nonconforming l
Conditions and Operfoility," was implemented into procedures and effectively used when l
BG&E credited manual operation to assure the continued post-seismic operability of the SRW l
system. Multi-level and multi-discipline reviews were required to assess operability. In
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addition, appropriate guidance was provided to allow continued system operability based on reasonable expectation that the system would remain operable, despite degraded system j
performance or the lack of a conclusive engineering determination. During operability l
reviews, assessment of whether an issue represented a nuclear safety concern was also performed.
QL-2-100 (formerly CCI-169) screening criteria for operability concerns identified that if the j
l problem of concern could affect technical specification equipment and the equipment may not l
be capable of performing its intended safety function, or does not or could not meet i
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operational of functional requirements, then an operability concern existed. IAU-02 required that activities or conditions that could affect the qualification or operational characteristics of
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installed components classified as safety-related be screened as Nuclear Safety Significant and l
be brought to the attention of Supervisor-IAU who would contact the Plant Operations and
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Safety Review Committee (POSRC) Chairman. QL-3-102 (formerly CCI-l16) screening
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criteria required that a PDR be created for an issue which adversely affected nuclear safety.
Nevertheless, although DES clearly had an operability question and considered several j
measures necessary to ensure continued operability of the SRW sytem, the issue was screened as not being an operability concern by the originater, the reviewing supervisor, the
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Issues Assessment Unit, the Issue Report Review Group, the Quality Audits Unit, and the i
POSRC.
The procedures required timely operability assessments during initial issue report reviews.
The evaluations conducted for the reports were apparently not of sufficent detail to conclude
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that the elevated system leak rates represented a condition adverse to continued SRW system operation. Inspectors assessed that the failure to identify the condition as an operability concern contributed to the lack of timeliness by BG&E in resolving the issue.
5.3 The 10 CFR 50.59 Safety Evaluation for Cross-Connecting Saltwater to SRW was Satisfactory l
The inspectors audited the 10 CFR 50.59 safety evaluation (BG&E log no. 92-B-011-186-R00) written for the installation of the saltwater to service water (SRW) cross-connection and determined that no technical specification change or unreviewed safety question existed. The safety evaluation assessed the diversion of salt water from the saltwater system and the impact of salt water on the SRW system. The proposed change was safe, appropriately based on past operating experience, and implemented in accordance with Calvert Cliffs Instruction (CCI)-702, " Change Control Process Overview." Comprehensive reviews were performed for corrosion, erosion, material compatibility, temperature of the makeup water, silting and impurities. These reviews were based on sound and discriminate engineering judgements. The assessments were supported by multi-discipline expertise and well documente.4 The SRW Operatine Instructions were Inadeauate Prior to having the Current Leakrate Guidance and Contingency Plans in Place Inspectors noted that operating instruction (01)-15, "SRW System," was changed on 10/7/93 to add the contingency for cross-connecting the safety-related saltwater system to the SRW system. This supplemented the non-safety-related demineralized water, condensate, and fire protection system makeup sources. The OI required escalated operator action based on increasing system leakage. On 10/14/93, an operability leakage limit of 16.3 gpm was added to the OI, along with additional required operator actions. Periodic leakrate determinations done by operators and frequent chemistry sampling provided reliable indications of system abnormalities. Historical system leakrates were generally low, between 1-2 gpm. As a result, inspectors assessed that the current SRW operating instructions were satisfactory.
However, before the saltwater makeup source and the operability guidelines were proceduralized, the operating instructions were not adequate. An operability issue existed which was not resolved by design and plant engineering personnel in a timely manner, and the resolution was not communicated to operators via procedural guidance in a timely manner.
5.5 The CCW Onerating Instructions were Satisfactory Inspectors noted that operating instruction (01)-16, "CCW System," provided three non-safety-related makeup sources to CCW: the demineralized water, condensate, or fire protection systems. The OI provided a leakrate limit for operator action and for initiating an operability review. Periodic leakrate determinations done by operators and frequent chemistry sampling provided reliable indications of system abnormalities. Historical system leakrates were generally near zero. BG&E was evaluating the need for adding operability limits based on leakrate to the OI.
BG&E had determined that the CCW system was not required to safely shutdown the plant.
If the SRW system failed; however, CCW would be required for the recirculation phase j
following a loss of coolant accident. SRW provided cooling to the containment air coolers (CACs), which could remove containment and core heat following certain accidents if the shutdown cooling heat exchangers were unavailable. If the SRW system failed, CCW i
cooling to the shutdown cooling heat exchangers would provide heat removal.
Since CCW was not required to safely shutdown the plant, inspectors assessed that the CCW operating instructions were satisfactory.
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5.6 BG&E's Resoonse to the Excessive SRW Irak on Seotember 29 was Satisfactory:
however. Some Weaknesses were Noted Based on interviews with BG&E personnel and review of logs and documentation, the inspectors concluded that BG&E's response to the excessive SRW leak on September 29 was satisfactory; however, some weaknesses in communications and timeliness were noted.
A good safety perspective was exhibited by the control room staff in immediately assessing the integrity of the primary system on September 29, when they identified that the l
containment sump fill rate was higher than normal. This contributed to an assessment of the safety significance of the leak and focused operators toward determining the source of the I
water. Operators demonstrated good system knowledge and awareness of plant material condition in their preliminary conclusion that the leakage was from the SRW system.
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Appropriate consideration was placed on evolutions in progress, specifically the surveillance test procedure, in reaching their conclusion. An assessment of SRW system leakage was made and independently verified by the Shift Technical Advisor that correlated the leakage l
with the containment sump fill rate. Troubleshooting was immediately commenced and l
conducted in accordance with appropriate procedural guidance.
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Inspectors noted, however, that during a system leak determination, 01-15, "SRW System,"
required that "...if leakage is greater than 15 gpm, then... notify plant chemistry, notify the l
system engineer, and submit issue report with an operability /reportability concern." Also, Calvert Cliffs Instruction (CCI)-315, " Functional Evaluation / Operability Determination,"
required that, immediately upon notification of an operability issue, the Shift Supervisor _ shall notify the General Supervisor - Nuclear Plant Operations (GS-NPO). CCI-315 required the
performance of a timely functional evaluation by the engineering staff to support an
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operability determination. Based on interviews with BG&E management, the intent of the notifications and reports required by 01-15 and CCI-315 was to assure comprehensive and multi-disciplined reviews to assure an integrated and focused review of degraded plant or system conditions. Nevertheless, the system engineer, chemist, and GS-NPO were not informed of the high leakrate until about six hours after it was discovered. An issue report l
was not written until about nine hours after the leak was discovered. When interviewed, the l
Shift Supervisor said he did not feel he had sufficient information to report until he had
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completed troubleshooting, at which point he knew personnel would be arriving for the l
normal work day. Despite the apparent lack of timeliness in notification, the Superintendent-i Nuclear Operations and the GS-NPO stated that their expectations with regard to compliance with the procedure had been fulfilled. There were no actual safety consequences to the delay j
l in notification.
l The operability concern regarding continued system operation with leakage rates in excess of the maximum fill rate of the saltwater to SRW cross-connection was not communicated to the Shift Supervisor in a timely manner. In fact, inspectors found that the GS-NPO and operators were unaware of any operability limits with regard to SRW leakrate. This resulted in a delayed entry into the technical specification action statement.
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When the Auxiliary Systems Principal Engineer went into the control room on the morning of September 29, he discussed the troubleshooting efforts and high SRW leakrate with the Shift Supervisor and GS-NPO. He was aware of the 16.3 gpm operability criteria considered by Design Engineering Section. He did not discuss the operability of the system with the Shift Supervisor or GS-NPO. When interviewed, he said that he wanted to assess the troubleshooting effort and discuss it with his engineers. He also did not want to take the system out of service without good reason, as it would have been an unnecessary transienh Inspectors noted that a system may be declared " inoperable," that is, incapable of performmg its safety function, and still remain in service in a degraded condition.
The Auxiliary Systems Principal Engineer discussed the high leakrate with the General Supervisor-Plant Engineering Section (GS-PES) about two hours after discovering it. At that time, the GS-PES called the GS-NPO and recommended that the system be considered inoperable based on the high leakrate. By that time, operators had already declared the system inoperable because the heat exchanger had been isolated. Inspectors considered the failure to communicate the operability concern or criteria between Design Engineering, Plant
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Engineering, and Nuclear Plant Operations Sections to be a weakness. The weakness resulted in the failure to recognize that the SRW system was operating in a significantly degraded mode requiring an operability assessment. There were no actual safety consequences to the weakness, because normal makeup sources were always available.
Despite weaknesses in timeliness and interdepartmental communications, the response of the operators to the excessive SRW leak on September 29 was satisfactory.
5.7 The Corrective Action Process for this issue Was Weak The concern that the SRW and CCW operating instructions (OIs) provided no limits for maximum leakage was identified by BG&E issue report (IR) IR0-013-020 on 11/25/91. IR5-011-534 on 10/14/92 identified that long term operability of the SRW system following an accident involving a loss of offsite power was questionable, that the makeup sources were not safety-related, and that the OI allowed continued operation without an operability assessment regardless of actual leakrate. It also identified that, while leakage rates are generally less than 2 gpm, the leakage rates exceeded 5 gpm on many occasions and went as high as 16.1 gpm with the system still considered operable. A leakrate of 16 gpm could empty a SRW head tank in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Nevertheless, the issue reports were screened as being of no operability concern.
Design Engineering Section evaluated the issue in November 1992 and concluded that no immediate operability concern existed, but that contingencies for a safety-related makeup source and leakage limits should be put in place to ensure SRW operability. Plant Engineering, Compliance, and Operations representatives concurred, and an initial operability determination was provided to the General Supervisor-Nuclear Plant Operations. The same day, equipment to accomplish a saltwater-SRW cross-connect was staged, operator trammg i
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on the cross-connect was begun, and procedure change requests were prepared.
Nevertheless, operators were not given any operability guidance or basis for the cross-connect.
The procedure changes were not prioritized as intendert. Based on interviews with some of the personnel who prepared the initial operability determination, the procedure changes were expected to be completed in about a week. This expectation was not communicated to procedure change personnel, who prioritized the changes for completion in the 3-6 month timeframe.
The Program Deficiency Report (PDR) subsequent to the second IR was downgraded by the
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QAU Review Team. The team did not review any of the corrective actions associated with the PDR prior to downgrading it. In fact, none of the corrective actions were yet complete.-
A contingency plan to cross-connect saltwater to SRW was first put in place in November 1992, but a 10 CFR 50.59 Safety Evaluation to support the cross-connection was not apprceed until March 1993. Escalated actions based on increasing SRW leakrates were not proceduralized until March 1993. The saltwater-SRW cross-connect and the operability -
criteria based on SRW leakrate were not proceduralized until October 1993. The requirement for an operability determination based on CCW leakrate was not proceduralized until November 1993.
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The inspectors concluded that BG&E's identification of the concern over long-term operability of the SRW and CCW systems without a safety-related makeup source was commendable. However, the failure to promptly resolve the concern was a violation of 10 CFR 50, Appendix C, Criterion XVI, which requires that " Measures shall be established to assure that conditions adverse to quality...are promptly identified and corrected." (VIO 93-31-01). The failure to conduct a timely operability determination, the failure to conduct a timely safety evaluation for the cross-connection, the failure to proceduralize the cross-connection within the operating instruction and provide operability guidance in a timely manner, and some weaknesses in BG&E's response to the excessive SRW leak in September 1993 stemmed from a failure of the corrective action program to resolve the identified issue promptly.
In addition, inspectors noted some potential weaknesses in BG&E's basis for the operability
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assessments and the operability criteria for the systems. Regarding the SRW system, Program Deficiency Report (PDR) 92488 stated the Design Engineering Section position that the SRW system was inoperable if system out-leakage exceeded 16.3 gpm, based on maximum makeup flow available through the 1-1/4 inch saltwater-SRW cross-connect hose.
This was the basis for the 01-15 change on 10/14/93 to require that a SRW subsystem be declared inoperable if system leakage was determined to be greater than 16.3 gpm.
01-15 measured system leakage by isolating makeup to the SRW head tanks and comparing i
the head tank levels after an hour has elapsed. A figure was provided in the procedure to
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calculate the volume lost from the change in head tank level and allow the operator to calculate a leak rate. The leak rate was measured for the existing system configuration.
During normal operations, the safety-related portion of the SRW system communicates with the non-safety-related portion of the SRW system, and so the measured leakage was from both portions of the system, which appeared to be conservative.
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However, given that the maximum allowed out-leakage was based on available makeup flow i
during a seismic event, it would appear that the " measured system leakage" should have taken into account leakage from the system that would be expected during a seismic event.
The non-safety-related portion of the SRW system is not seismically qualified and, therefore, must be assumed lost during a design basis earthquake. SRW system turbine building isolation valves SRW-1600, -1637, -1638, and -1639 (supply control valves) and SRW-323, -324, and -325 (return check valves) would then become the seismically qualified SRW i
pressure boundary. Current inservice test (IST) practice using surveillance test procedure (STP) O-67-C, "SRW System Turbine Building Header Test," allows 10 gpm of leakage through these valves per subsystem. Past practice allowed 10 gpm through the worst
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subsystem supply control valve and 10 gpm through the worst subsystem return check valve, resulting in a potential total subsystem leakage of 20 gpm.
BG&E's position was that leak rates through the above isolation valves did not need to be included in the determination of a maximum leak rate measured by 01-15, because they have performed a Seismic Qualification Utilities Group (SQUG) walkdown of the non-seismic portions of the SRW system and have concluded that it will withstand a' seismic event without
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collapse or gross rupture. Identified potential break points were modified to prevent failure.
The NRC has not approved the SQUG methodology used to conclude that the non-seismic portion of the SRW system will withstand a seismic event, and the NRC is currently evaluating the isolation provisions for the SRW system at Calvert Cliffs. Inspectors considered it a weakness that BG&E did not include leakage through the SRW turbine
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building isolation valves in its system leak rate determination, based on the unapproved l
assertation that the non-seismic portions of the SRW system would withstand a seismic event.
Inspectors understood that the leak rate criteria will be re-evaluated by BG&E upon completion of NRC review ofisolation provisions for the SRW system during a seismic event.
Similarly, valves 1-CV-1645,1-CV-1646,2-CV-1645, and 2-CV-1646 were not in the IST
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program. These valves provided SRW to the 12 EDG (Swing Diesel) through seismically qualified piping. The position of the valves was determined by pressure switches which' seek the pressurized header. Since these valves were not in the IST program, their leakrate was indeterminate and, as such, could potentially be an unanticipated " cross-connect" of the Unit
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I and Unit 2 SRW systems. The maximum allowable leakrate determination did not take into account potential leakage through these valves.
Regarding the CCW system, isolation valves 1-CV-3840,1-CV-3842,2-CV-3840, and 2-CV-3842 were not leakrate tested. These valves isolated the Reactor Coolant Waste Evaporator
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during a loss of coolant accident. The piping to the Reactor Coolant Waste Evaporator was l'
not seismically qualified. The valves were stroke-time tested in accordance with the ASMEl Code and BG&E's IST program. The NRC has concurred that these valves are not required
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to be leak rate tested given their function (seat leakage is not consequential; the valves need only be checl:ed for " gross failure to close").
This potential leak path was not factored into the leakrate criteria in OI-16, "CCW System." Subsequent to the inspection, inspectors I
noted that BG&E planned to revise SRW and CCW system boundary valve testing to include
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the above valves.
BG&E had partially based their determination and implementation of leakrate criteria and their operability assessments of the SRW and CCW systems on the fact that the systems had
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historically low leakrates. The failure to include these boundary isolation valves in their evaluation is an Unresolved Item. (URI 93-31-02)
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6.0 BG&E RESPONSE
In response to deficiencies identified with regard to the resolution of Program Deficiency Report (PDR) 92488, the BG&E Quality Audits Unit conducted a complete' evaluation of l
issues originally included in the PDR system that were subsequently transferred to the issue
report (IR) system. Based on a review of actions completed either prior to, or after transfer.
to the IR system, all issues were effectively addressed.
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BG&E had previously identified weaknesses in elements of the Corrective Action Program,-
based on findings from their own quality assurance audits as well as the NRC. BG&E-
Quality Assurance Audit 93-12, dated September 22,1993, concluded that the Corrective I
Action Program was ineffective in preventing significant conditions adverse to quality :
l (SCAQs) from recurring. Additionally, SCAQs were not being resolved in a timely marmer:
and were recurring despite completion of corrective actions. Similar problems had been-
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noted the previous year in Audit 92-12. NRC Systematic Assessment of Licensee Performance (SALP) Report 50-317 and 318/91-99 noted that timely and effective corrective actions were not always achieved due to ineffective management contmls and oversight.
NRC SALP Report 50-317 and 318/92-99 noted that follow-through on corrective actions for some events was not always fully effective.
In response to these concerns, the Plant General Manager initiated action on October 20,1993, to address and remedy the deficiencies in the Corrective Action Program.
Inspectors reviewed the plan and found it to be a credible schedule for identifying, prioritizing, and resolving the issues, and assessing the effectiveness of the corrective actions. Emphasis was appropriately placed on management and supervision and issue -
resolution. The estimated completion date for the action plan was December 1994. BG&E considered that the deficiencies noted in their response to the concern with the SRW and CCW systems, existing system leakage, ud makeup water sources was another example of Corrective Action Program weaknesses that the action plan was intended to correc.
7.0 FOLLOW-UP OF PREVIOUS INSPECTION FIhTINGS 7.1 (Closed) Unresolved Item 50-317 and 318/93-28-01 The item concerned apparent weaknesses in BG&E's corrective action process and interdepartmental communications, and potential system design deficiencies. The item will -
be reviewed as part of the assessment of BG&E's response to violation 50-317 and 318/93-31-01.
8.0 MANAGEMENT MEETING During this inspection, periodic meetings were held with plant management to discuss inspection observations and findings. Attachment 3 lists the personnel contacted during the inspection. At the close of the inspection, an exit meeting was held to summarize the conclusions. No written material was given to the licensee and no proprietary information related to this inspection was identified.
8.1 Preliminary Insoection Findings A violation was identified with regard to the failure to promptly resolve the concern over the long-term operability of the SRW and CCW systems with existing leakage and without a safety-related makeup source, as discussed in section 5.7. An unresolved item was identified with regard to the failure to include some boundary isolation valves in the leakrate determination for operability guidance, as discussed in section.
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ATI'ACHMENT 3 j
I PERSONNLL CONTACTED DURING THE INSPECTION BG&E Personnel A. Anuje, Supervisor, Quality Audits Unit *
C. Cruse, Plant General Manager *
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T. Delaney, Principal Engineer, Auxiliary Systems Engineering Unit (ASEU) *
G. Detter, Director, Nuclear Regulatory Matters *
i J. Gines, ASEU *
i N. Haggerty, Licensing Unit A. Henni, Mechanical Engineering Unit (MEU) *
C. Ludlow, Principal Engineer, MEU M. Milbradt, Compliance Unit D. Shaw, Plant Engineering Section A. Simpson, ASEU *
R. Szoch, Principal Engineer, Instrumentation and Controls Engineering Unit
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J. Carroll, Chairman, Plant Operations and Safety Review Committee *
C. Faller, ASEU *
i W. Kemper, Plant Engineering Section *
J. Imhr, Operating Experience Review Unit *
B. Montgomery, Principal Engineer, Licensing Unit *
j C. Sly, Compliance Unit *
G. Strauss, Plant Testing Unit *
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L. Tucker, General Supervisor - Plant Engineering Section *
R. Wenderlich, Superintendent - Nuclear Operations *
J. Hill, General Supervisor - Nuclear Plant Operations R. Branch, Procedures Management Unit J. Denton, Plant Operations Unit C. Dunkerly, Issues Assessment Unit j
M. Junge, Quality Audits Unit C. Andrews, Supervisor, Procedure Develop & Mods Accept Unit (PDMAU)
D. Lynch, PDMAU j
C. Drumgoole, PDMAU
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NRC Personnel P. Wilson, Senior Resident Inspector *
M. Boyle, Chief, Reactor Projects Section.l A *
D. Mcdonald, Senior Project Manager, NRR R. Capra, Director, Project Directorate I-1, NRR W. I2fave, Plant Systems Branch, NRR P. Baranowsky, Chief, Trends & Patterns Analysis Branch, AEOD
- Denotes those present at the exit meeting on November 19, 1993.
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