IR 05000397/1993029
| ML17290A633 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 08/13/1993 |
| From: | Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17290A632 | List: |
| References | |
| 50-397-93-29, NUDOCS 9309290224 | |
| Download: ML17290A633 (23) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION V
Report No:
Docket No:
License No:
Licensee:
Facility Name:
Inspection at:
Inspection Conducted:
Inspectors:
50-397/93-29 50-397 NPF-21 Washington Public Power Supply System P. 0.
Box 968 Richland, WA 99352 Mashington Nuclear Project No.
(WNP-2)
WNP-2 site near Richland, Washington July 19, 1993 August 2, 1993 R.
C. Barr, Senior Resident Inspector, WNP-2 D, L. Proulx, Resident Inspector, WNP-2 Approved by:
~Summar:
P.
H. Johnson, Chief Reactor Projects Section
Date Signed Ins ection on Jul 19 throu h Au ust
1993 Re ort No. 50-397 93-29 Areas Ins ected:
Special, announced, resident inspection of the licensee's discovery that a design deficiency in the suppression pool cooling (SPC)
mode of the residual heat removal (RHR) system that resulted in an apparent violation of TS 3.6.2.3,
"Suppression Pool Cooling," on August 6, 1990.
During the inspection, Inspection Procedures 90712, 92700, and 92701 were used.
Safet Issues Mana ement S stem SIMS Items:
None.
Results:
General Conclusions and S ecific Findin s:
On August 6, 1990, the Supply System failed to meet the requirements of Technical Specification 3.6.2.3,
"Suppression Pool Cooling" (Paragraph 10).
This appeared to have resulted from a design deficiency in the RHR system.
This deficiency went uncorrected from February 1987 to October 1990 due to ineffective evaluation and resolution of a clearly documented industry issue (Paragraph 10).
Additionally, after the Supply System initiated corrective actions to limit operation of the RHR system in SPC mode to a single train, WNP-2 plant operators violated procedures on three occasions by placing two 9309290224 930817 PDR ADOCK 05000397 Q
trains of the RHR system in the SPC mode (Paragraph 10).
Weaknesses:
This inspection highlighted the following problems that the NRC has previously identified at the Supply System:
inadequate evaluation and correction of industry issues, inadequate management oversight, poor interorganizati onal communication, and the failure to adhere to procedures.
Summar of Violations and Deviations:
Three apparent violations were identified (Paragraph 10), involving (1) operation of two trains of the RHR system in the SPC mode, a condition which renders the trains inoperable for post-accident response, in violation of TS 3.6.2.3, (2) inadequate evaluation and correction of a design deficiency associated with the SPC mode of RHR, resulting in the operation of two RHR trains in the SPC mode, and (3) failure to adhere to procedures, resulting in the operation of two RHR trains in the SPC mode (instead of one RHR train, as prescribed by procedure).
DETAILS Persons Contacted
- L. Oxsen, Deputy Managing Director
- V. Parrish, Assistant Managing Director for Operations
- J. Swai les, Plant Manager
- J. Gearhart, Director, guality Assurance W. Shaeffer, Operations Manager
- R. Webring, Manager, Technical Services Division
- R. Barbee, Manager, Systems Engineering
- R. Fuller, Licensing Engineer
- G. Smith, Manager, Operations Division
- J. Parker, Projects Manager (Acting)
- S. Kirkendall, Supervisor, Plant Support Engineering
- D. Overman, Engineer, Plant Technical J. Williams, Engineer, Plant Technical
- L. Grumme, Manager, Nuclear Safety Assurance J.
Massey, Engineer, Probability Risk guality Assurance P. Inserra, Supervisor, Plant Technical P. Harness, Supervisor, Mechanical Design Engineering The inspectors also interviewed various control room operators, shift supervisors and shift managers, maintenance, engineering, quality assurance, and management personnel.
- Attended the Exit Meeting held on August 3, 1993.
Descri tion and Desi n Basis of the Su ression Pool Coolin Mode of RHR'
The residual heat removal (RHR) system is one of several systems that protect the reactor core and fuel against overheating.
This system performs several functions which include the following operating modes:
low pressure coolant injection (LPCI), containment spray cooling (CSC),
suppression pool cooling (SPC)
and shutdown cooling (SDC).
The SPC mode cools containment by removing heat from the suppression pool (SP).
This mode of RHR system operation is manually p1aced in service to limit the temperature of water in the suppression pool following a design basis loss-of-coolant accident (LOCA), to control the suppression pool temperature during normal. operation of the safety relief valves (SRVs)
and the reactor core isolation cooling (RCIC) system, and to reduce the suppression pool temperature following an isolation transient.
When either train of RHR is in the SPC mode, the associated RHR pump takes a
suction on the suppression pool water via RHR-V-4A(B), pumps the water through the RHR heat exchanger, and returns the water to the suppression pool via RHR-V-24A(B).
The design basis of the SPC mode of RHR is to limit SP water temperature to less than 170'F following a LOCA.
Technical S ecifications A
licable to the Su ression Pool Coolin Mode Technical Specification (TS) 3.6.2.3,
"Suppression Pool Cooling," applies to the SPC mode of RHR.
The Limiting Condition for Operation (LCO),
which requires two operable loops of SPC, is applicable in operational conditions 1,
2 and 3.
Action Statement b. requires that the plant be in at least hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, when both SPC loops are inoperable.
Mater Hammer in Boilin Mater Reactor BWR RHR S stems
'n early 1980, nuclear utilities that operated BWRs reported numerous instances of water hammer in the RHR system both during preoperational testing and following issuance of their operating license.
The water hammer events resulted from the draining or leaking of water from the RHR system and the subsequent restart of the RHR pumps on a partially voided piping system.
Generally, the partial draining of the RHR system piping was due to inadequate filling and venting of the system, component deficiencies, such as leaking valves, or system design deficiencies.
Due to the number and significance of water hammer events being reported, the NRC and the industry distributed generic and specific information concerning the potential for water hammer in RHR systems of BWRs.
This special inspection report describes the Supply System's evaluation and correction of a specific RHR design deficiency'hat could have caused a
significant water hammer due to water draining from the RHR discharge piping following a coincident loss of offsite power (LOOP)
and LOCA when in the SPC mode of RHR.
Previ ous Histor of RHR Desi n Defici enc 92701 Potential for Mater Hammer Durin the Restart of RHR Pum s at BWR Nuclear.
Power Plants.
In April 1983, the NRC's Office for Analysis and Evaluation of Operational Data (AEOD) distributed to all nuclear utilities Engineering Evaluation E309,
"The Potential for Water Hammer During the Restart of RHR Pumps at, BWR Nuclear Power Plants," for information.
This evaluation recommended the following actions:
(1) that utilities very carefully examine the RHR system for configurations that could be vulnerable to water hammer; (2) that procedures and operator training specifically address the potential for water hammer during operation of the RHR system in the SPC and CSC modes; (3) that procedures and operator trai ning address the desirability of operating only one train of RHR in the SPC and CSC modes; and (4) that emergency procedures be developed for restarting the RHR system pumps following a draindown.
On August 31, 1983, the Supply System obtained Engineering Evaluation E309 based on its reference in Significant Event Report (SER) 55-83, and opened Operational Event Review (OER)
81078C to evaluate Engineering Evaluation E309.
The Nuclear Safety Assurance Group (NSAG) reviewed the AEOD evaluation and took no action other than those implemented for SER 55-83, described below.
Si nificant Event Re ort SER 55-83 On August 9, 1983, the Institute of Nu'clear Power Operation (INPO)
distributed SER 55-83 to all nuclear utilities, including the Supply il
J
System.
INPO distributed this SER based on their review of the NRC's AEOD Engineering Evaluation E309.
This SER informed utilities that water hammer in the SPC mode of RHR was likely to occur during non-emergency conditions.
The SER noted that the affected loop could be rendered inoperable due to void-induced water hammer following a LOCA and a
LOOP.
The Supply System received this SER and, to evaluate its applicability, opened item OER 81078B in the OER program.
The NSAG was assigned to evaluate this SER.
On September 14, 1983, following their assessment of the SER, the NSAG recommended the following five actions:
(1)
NSAG prepare and issue an addendum to the Monthly Operational Bulletin (MOB) article on water hammer to include a brief description of the scenario in the AEOD Evaluation; (2)
Revise PPM 2.4.2,
"Residual Heat Removal," to include a caution or limitation concerning water hammer potential following a loss of power to the RHR pumps; (3)
Revise PPM 5.2.2 and 5.2.3, and 5.2.4 to include a statement that RHR train B is the preferable train to use for drywell spray; (4)
Revise PPM 5.0.0 to warn of the potential of water hammer after a loss of power while operating RHR in the SPC or CSC modes; and (5)
Revise PPM 4.4.2. 1, "Loss of RHR Shutdown Cooling Mode Loops," to direct restarting of an RHR pump followed by slowly opening RHR heat exchanger bypass valves to minimize the effects of water hammer.
Recommendations 1,
2 and 3 were implemented.
On June 29, 1984, recommendations 4 and 5 were revised to change different procedures than were originally proposed.
Those procedures were subsequently revised.
NRC Information Notice IN 87-10
"Potential for Water Hammer Durin Restart of Residual Heat Removal Pum s" On February ll, 1987, the NRC issued Information Notice (IN) 87-10,
"Potential for Water Hammer During Restart of Residual Heat Removal Pumps."
The Supply System received the IN on February 25, 1987, and distributed the IN for information to the Managers of guality Assurance, Engineering, Operations, Maintenance and Licensing.
The Supply System opened OER 81078C and assigned the evaluation to the NSAG to evaluate the applicability of the IN to WNP-2.
To determine the applicability, the NSAG discussed the RHR-SPC mode of operation with a Shift Manager and a Control Room Supervisor (CRS).
These individuals stated that restrictions on the operation of the RHR system in the SPC mode were not necessary because WNP-2 had not experienced a high heating rate of the suppression pool as would. be caused by excessive main steam relief valve (MSRV) leakage.
Based on the discussions the NSAG personnel had with the shift personnel and the actions that had been previously implemented for SER 55-83, NSAG concluded that no further corrective actions were required as a result of receiving the I However,'n October 19, 1987, the NSAG evaluator, who performed the initial assessment of IN 87-10, found that, due to leaking MSRVs, the SPC mode of RHR was being used approximately daily, which was significantly more than assumed in the safety analysis.
The Supply System took no additional action based on this observation.
Gilbert Services Inc.
GSI Information Notice M-027-01 On March 2, 1987, Gilbert Services, Inc.,
an engineering company, distri-buted GSI Information Notice M-027-01,
"RHR Water Hammer," to the Supply System.
This notice, previously included as an attachment to IN 87-10, specifically described the plant conditions and design configurations that could result in a significant water hammer to the RHR system while in the SPC mode.
This notice stated,
"Since the SPC mode usage factor at many BWRs is higher than assumed in the initial design basis, the poten-tial for this scenario is significantly increased."
However, the Supply System took no additional actions as a result of receiving this notice.
En ineerin Evaluation E309
"Potential for Water Hammer Durin Restart of RHR Pum s at BWRs" On November 18, 1988, in response to an, informal contact with another licensee, the Supply System received AEOD Engineering Evaluation E309 for a second time.
An NSAG engineer opened OER 81078G.
During the screening process the engineer determined that report E309 had previously been entered into the OER program.
The NSAG engineer who screened this OER noted that the Supply System had changed procedures in 1984; therefore, he concluded that no additional actions were required.
0 eration of Both RHR Trains in the SPC Mode On October 2, 1990, a licensee engineer, who was evaluating containment cooling TSs as part of the Technical Specification Improvement Program (TSIP) recognized that the Supply System's corrective actions in response to NRC Information Notice 87-10 had not effectively addressed or resolved operation of the RHR system in the SPC mode.
Based on this observation, the Supply System revised OER 81078E.
This resulted in two changes to PPM 2.4.2 (Revision 16).
Specifically, Caution 5.22 stated,
"During non-emergency conditions, do not align more than one RHR loop in the suppression pool cooling mode at a time.
A design basis LOCA coincident with a Loss of Offsite Power that occurs while one or more RHR loops are in the suppression pool cooling will void portions of the affected RHR loops.
Water hammer may occur in the affected RHR loops when the diesel generators restart the RHR pumps."
Also, Cautions in Paragraphs 6.7 and 6.8 stated,
"Do not operate more than one loop of RHR in the SPC mode at a time during non-emergency conditions.
The affected loop may be rendered inoperable due to void induced water hammer following a LOCA."
On November 4, 1992, while investigating Problem Evaluation Request (PER)
292-1191,
"Elevated Suppression Pool Airspace Temperatures Due to Leaking MSRVs," a Supply System engineer discussed NRC IN 87-10 with a represen-tative of another nuclear utility.
During that discussion, the Supply System engineer concluded that IN 87-10 may not have been adequately dispositioned.
He opened PER 292-1243 to assess the potential for water
e
hammer in the RHR system while in the SPC mode.
On December 22, 1992, a
Supply System engineering evaluation concluded that a water hammer could cause failure of a train or trains of RHR operating in the SPC mode during a LOOP coincident with a LOCA.
The evaluation concluded that the water hammer could cause failure in portions of the associated RHR piping and RHR heat exchanger, resulting in loss of the SPC, suppression pool spray (SPS),
drywell spray (DWS), and LPCI.
On January 13, 1993, during a review of WNP-2 logs, a Supply System licensing engineer found instances in which the Supply System had operated both trains of RHR in the SPC mode.
The Supply System made the required 4-hour and 30-day notifications to the NRC.
Additionally, the Supply System performed a root cause evaluation, as documented in Non-conformance Report (NCR) 292-1243, to determine the causes of inadequate resolution of the issue involving RHR operation in the SPC mode.
The licensee determined that on September 30, 1991, July 6, 1992, and July ll, 1992, the Supply System had operated trains A and B of the RHR system in the SPC mode following HSRV testing, in violation of the proce-dural requirements.
In these instances, both RHR trains were operated in the SPC mode for approximately 4g, 6,
and 2Q hours, respectively.
In addition, the licensee identified that on August 6, 1990, the Supply System had operated trains A and B of the RHR system in the SPC mode for approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, which is longer than allowed by the 12-hour TS action statement.
On April 14, 1993, Supply System Engineering completed calculation NE-02-93-01,
"RHR B Water Hammer Evaluation, Suppression Cooling Node."
The purpose of the calculation was to determine the resultant forces in the RHR piping system due to a concurrent LOCA and LOOP while in the SPC mode.
This analysis indicated that a maximum force of 130,000 pounds would be exerted downward at the RHR heat exchanger.
The calculation concluded that, if in the SPC mode of operation during a coincident LOCA and LOOP, the RHR loop would fail.
Specifically, the maximum force allowed on the RHR pump nozzle is 8,363 pounds.
The reflected pressure waves from the water hammer would exceed that value substantially.
Therefore, any RHR train operating in the SPC mode during such an occurrence would be rendered inoperable (see paragraph 8.d for additional discussion of the effects of a water hammer).
Licensee Event Re ort LER Followu 90712 92700 LER 93-01 Revision 0 and Revision
Closed
- "Ino erable Su ression Pool Coolin Due to Potential Waterhammer."
In this LER and its revision, submitted pursuant to
CFR 50.73(a)(2)
(i)(B) and (v)(B), the Supply System reported that Technical Specifica-tion 3.6.2.3, which establishes limits for the SPC mode of operation of the RHR system, had been exceeded.
The licensee's conclusions, as reported in this LER, wer e based upon the findings of PER 292-1243 and the formal root cause evaluation as documented in NCR 292-1243.
In the LER and the NCR, the licensee identified the following root causes for the August 6, 1990, violation of TS 3.6.2.3:
~
Inadequacy in the original design analysis.
~
Hanagement methods which did not ensure appropriate technical input during review and closure of OER 81078E.'he licensee identified the failure to follow procedures as the cause of operating two inoperable trains of the RHR system in the SPC mode on September 30, 1991, July 6, 1992, and July 11, 1992.
In the LER, the licensee discussed one immediate corrective action, five followup corrective actions and four future corrective actions to preclude exceeding TS limits.
The following briefly describe the more significant of these actions:
~
On December 22, 1992, the Supply System issued a "Night Order" directing shift personnel to declare any train of RHR inoperable when it is operating in the SPC or SPS mode.
~
Prior to the 1993 refueling outage (R8), the Supply System determined the appropriate maintenance activities required to significantly reduce MSRV leakage.
~.
Supply System and contract craftsmen refurbished 13 HSRVs during the R8 outage.
~
Supply System engineers performed an evaluation to determine the long-term solution to the RHR design deficiency.
~
The Supply System trained all operating crews and operating crew supervisors regarding mandatory procedure compliance.
The Supply System also identified actions that they had taken following the event, but prior to the discovery of the event, that would likely have prevented the event had the actions been implemented in 1987.
These actions included changing the OER process and the
CFR 50.59 process.
The licensee concluded that this event did not affect the public health and safety.
The licensee concluded that the'safety significance of these violations was negligible because the probability of a coincident LOOP and LOCA when in the SPC mode of RHR (2.9E-7)
was approximately two orders of magnitude less than the core damage frequency (CDF) (5.4E-5).
The inspectors reviewed the LER and NCR, discussed the licensee's findings and conclusions with the personnel who drafted these documents, verified selected corrective actions, and walked down selected portions of the RHR system.
The inspectors found that the NCR and the LER were thorough and critical and identified the root and contributing causes of the event.
During their walkdown of the RHR system, the i'nspectors found no evidence of damage due to water hammer.
The inspectors noted that the licensee's calculation of probability did not use the longest times the RHR had been operated in the SPC mode; therefore, the event probability was greater than 2.9E-7 before the procedure was corrected in December 1992 (see paragraph 9).
f
8.
NRC Event Followu 92700 92701 As followup, the inspectors discussed this issue with managers, operators and engineers who were involved in the various aspects of the issue; reviewed the various licensee event evaluations associated with it; reviewed selected OER evaluations that were associated with potential water hammer in the RHR system; reviewed calculation NE-02-93-01,
"RHR B
Mater Hammer Evaluation, Suppression Cooling Mode"; and reviewed selected RHR operating and surveillance procedures to address the following areas:
a ~
0 erator Trainin OER 81078B, recommendation 1, suggested the issuance of an addendum to the Honthly Operational Bulletin (HOB), to include a brief description of the water hammer scenario which could result from draining the RHR system while in the SPC mode.
The HOB was issued on September 25, 1984.
The HOB stated, H
"As a supplement to the November 1982 HOB article on water hammer, this article discusses additional scenarios which may cause water hammer in the RHR system....
Possible consequences of water hammer caused by a postulated loss of power to the RHR pumps following a LOCA are:
1.
Failure of a containment spray penetration resulting in release of significant amounts of radi oacti vity to the reactor bui lding environs if operating in the containment spray mode.
2.
Failure of instrument lines which disables control instrumentation and allows loss of radioactive water if operating in the SPC mode.
MNP-2 RHR system design is susceptible to draindown similar to the events described above."
The inspectors found that training had been conducted on this issue.
However, at no time during the training did anyone question whether the procedure change adequately resolved the issue of water hammer in the SPC mode of RHR.
The inspectors noted that, contrary to the HOB item 1, the 1993 engineering calculation proved that the containment penetrations would remain intact, but that other sections of piping and components may be damaged.
b.
Procedure Revisions OER 81078B, recommendation 2, suggested revising PPM 2.4.2 to
"include a limitation or caution concerning water hammer potential following a loss of power to the RHR pumps."
Recommendation 2 was implemented; however, the recommendation did not adequately address the subject of operating in SPC with a coincident LOCA and LOOP.
The procedure change, Limitation 2.4.2.4.H, stated, "If during a power interruption, any RHR PUMP DISC PRESS HIGH/LOM alarm is received and adequate core cooling is assured, place the associated
pump control switch in Pull-to-Lock until the subsystem can be filled and vented."
In addition to the reviews associated with the OER, the reviews performed during the procedure approval process for Revision 3 of PPM 2.4.2 failed to detect that the procedure change did not address the initial issue of water hammer in the SPC and SPS modes with a concurrent LOOP and LOCA.
The reviews were performed by individual contributors, supervisors, and oversight groups, including the Plant Operating Committee.
Additionally, reviews of the other OERs on this subject that relied on this procedure change did not identify this deficiency.
n PPM 2.4. 2, Revisions 2 through 8, di d not include di recti on on establishing the SPC mode of RHR.
However, during their review of WNP-2 logs the inspectors noted that the licensee did operate in the SPC mode of RHR.
Based on discussions with licensee personnel, the inspectors found that during the time these revisions were effective, operators considered transitioning from any other mode of RHR to the SPC mode to be a simple task that required no procedure.
It appeared that-not having a procedure which referred to the SPC mode of operation contributed to the inadequate revision to PPM 2.4.2 because there was no section of the procedure associated with the SPC mode of operation to modify.
On September 26, 1990, the Supply System revised PPM 2.4.2 to include two cautions to warn operators about potential water hammer when operating the RHR system in the SPC mode.
Collectively, these cautions clearly described the issue and provided specific operational direction.
However, the inspectors noted that WNP-2 procedures were not consistent with industry practice because operational direction is included in caution statements.
Industry practice generally provides operational direction in procedure steps instead of caution statements.
The Supply System root cause analysis, Non-Conformance Report (NCR)
292-1243, noted that one of the Shift Managers involved with not adhering to PPM 2.4.2, Revision 16, was aware of the caution but, because of other considerations, decided to run two SPC loops at the same time.
The NCR noted that the other considerations were past experience in performing this activity and the perceived urgency to complete power ascension.
The inspectors discussed not adhering to PPM 2.4.2 with that Shift Manager referenced in NCR 292-1243.
He stated that while he was aware of the caution statement in PPM 2.4.2, he did not have specific knowledge of the actual concern of a coincident LOCA and LOOP when operating in the SPC mode of RHR.
The Shift Manager noted that he understood how the system responded during a LOCA and, based on his knowledge of the RHR system response during a LOCA, concluded that it was acceptable to operate two trains of RHR in the SPC mode.
The inspector questioned the Shift Manager as to why he did not resolve the conflict in the caution, as required by PPM 1.2.3, prior to proceeding with operating two trains of RHR in SPC.
The Shift Manager noted that at the time he thought there was no conflict because, as Shift Manager, he believed he had the latitude to depart
from procedures when hi s experience indicated that it was acceptable.
He stated that had the procedure more clearly stated the concern and the operating restrictions, he would not have departed from the procedure.
0 erational Reviews In reference to OER 81078E, the NSAG based their review of IN 87-10 almost exclusively on the input from two senior licensed operators.
These individuals stated that restrictions on operation of the RHR system in the SPC mode were not necessary because MNP-2 had not experienced high heat loads from excessive main steam relief valve (NSRV) leakage.
It appeared that the NSAG personnel involved in this evaluation did not verify the validity of these statements.
The inspectors reviewed MNP-2 logs and found that MSRVs were leaking in February 1987, and the SPC mode of RHR was being used, contrary to what the licensed operators stated.
On October 19, 1987, the NSAG engineer who performed the initial review of this OER recognized and documented that SPC was being used daily; however, no additional action was taken.
The inspector noted that this information was reviewed by the NSAG supervisor.
En ineerin Anal sis The inspectors found that Supply System calculation NE-02-93-01, Revision 0, used accepted codes and conservative assumptions. 'he calculation determined that in many cases the forces which would be generated by a water hammer would exceed the faulted allowable values for the piping support system.
In some cases the calculated forces would exceed the allowable amount by greater than 200~.
The calculation showed that the RHR heat exchanger and the RHR pump nozzle were two of the more limiting components.
The inspectors questioned whether a potential water hammer could cause pipe breaks or broken pipe joints, with resulting leakage.
These leaks would result in bypassing the primary containment by pumping potentially highly radioactive water from. the SP to either the RHR pump room or the RHR heat exchanger room, in addition to depleting the inventory of SP water used for accident mitigation.
The licensee noted that each room has a
sump with a high level alarm which annunciates in the control room.
The alarm response proce-dures would then direct the operators to isolate the leak.
The inspectors verified the operability of the sump alarms and the adequacy of the applicable procedures.
Conclusions The licensee's reviews of OERs 81078B, C and G, and procedure changes were weak.
The licensee did not verify that data were accurate, and their followup was not probing when inaccuracies were identified.
Additionally, management and supervisory oversight, including that provided by the Plant Operating Committee, was weak in that the reviews did not verify that the design deficiency was
o
-10-resolved by the implemented procedure changes.
The inspectors also concluded that weaknesses in the content, clarity and format of procedures had contributed to the operators'ailure to follow them.
The inspectors concluded that the October 1990 change to PPN 2.4.2 (Revision 16)
had adequately described the potential and the
.
conditions for water hammer when in SPC.
That revision also provided adequate direction to the operators on restricting operation of SPC to one train of RHR.
Therefore, it,appeared that the operating crews who placed two trains of the RHR system in SPC contrary to procedure requirements on September 30, 1991, July 6, 1992 and July ll, 1992, acted contrary to PPH 2.4.2.
The inspectors noted the personnel involved with non-adherence to the procedure did not attempt to clarify the intent of the caution prior to operating two trains of the RHR system in the SPC mode.
The inspectors concluded that communications and followup between organizations were weak.
For example, even though the NSAG assigned responsibility for disposition of the recommendations to Operations, the OER information was distributed to other organizations which could have questioned the thoroughness of the corrective action.
Additionally, due to poor inter-organizational communication, operators were never effectively trained on this issue.
9.
Safet Si nificance 92701 This event was significant because the margin of safety for maintaining containment integrity following a coincident LOCA and LOOP was significantly reduced by operating with a design deficiency in the SPC mode of RHR.
The licensee noted that the probability of a coincident LOOP and LOCA when in the SPC mode of RHR is approximately 2.9E-7, assuming less than 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> per week in the SPC mode.
This is approximately two orders of magnitude less than the core damage frequency (CDF) probability (5.4E-5).
Therefore, this event did not significantly increase the overall probability of an accident.
However, on occasion the Supply System has operated in the SPC mode of operation more than 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> per week.
The NRC considers that the failure to meet the requirements of TS 3.6.2.3,
"Suppression Pool Cooling," on August 6, 1990, was a significant safety issue because the Supply System operated the RHR system such that if a coincident LOCA and LOOP had occurred, both trains of the RHR system would not have been capable of performing their intended safety functions.
Also, the Supply System's resolution of a clearly documented industry issue was not timely or effective, even though there were multiple opportunities to correct this deficiency.
The NRC considers that the licensed operator's failure to adhere to the plant procedures, involving the operation of both trains of the RHR system in the SPC mode, to be a significant safety issue.
The procedures clearly stated that two train operation of RHR in the SPC mode was forbidden except in emergencies; however, operators did not adhere to these procedures on three non-emergency occasions, resulting in increased probability of a significant even.
A arent Viol ati ons 92701 Three apparent violations of regulatory requirements were identified during this inspection, as discussed in the foregoing paragraphs.
The violations are summarized as follows:
a.
In February 1987, the licensee apparently violated the requirements of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," by failing to implement effective corrective actions for the RHR system design deficiency which was discussed in NRC IN 87-10 and in industry documents distributed to the Supply System.
Criterion XVI requires that measures be established to assure that conditions adverse to quality and nonconformances are promptly identified and corrected.
Significant conditions adverse to quality must have the cause of the condition determined and corrective action taken to prevent repetition.
The potential vulnerability of the RHR system to a significant water hammer event while in the RHR-SPC mode of operation is a significant condition adverse to quality.
The Supply System's failure to restrict the operation of the RHR system in this mode of operation or to correct the RHR system design deficiency is an apparent violation of this requirement (50-397/92-29-01).
b.
On August 6, 1990, the Supply System apparently violated Technical Specification 3.6.2.3,
" Suppression Pool Cooling," by operating two trains of RHR in the SPC mode for approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following HSRV testing with a design deficiency that could have rendered both trains inoperable following a coincident LOOP and LOCA.
Specifi-cally.
TS 3.6.2.3 requires that the suppression pool cooling mode of the residual heat removal (RHR) system shall be operable with two independent loops, with each loop consisting one OPERABLE pump; and an OPERABLE flow path capable of recirculating water from the suppression chamber through an RHRSW heat exchanger.
The TS Action'tatement 3.6.2.3.b.
requires that the plant be in at least hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when both SPC loops are inoperable.
The licensee's failure to meet the requirements of TS 3.6.2.3 or to satisfy the associated action statement is an apparent violation (50-397/93-29-02).
Co The WNP-2 operators'ailure to adhere to a procedural requirement of PPN 2.4.2, Revision 10, on September 30, 1991, July 6, 1992, and July ll, 1992, when trains A and B of RHR were operated in the SPC mode, is an apparent violation of TS 6.8. 1, which states that
"Written procedures shall be established, implemented and main-tained..."
The Supply System's failure to meet this TS requirement is an apparent violation (50-397/93-29-03).
The above apparent violations highlight the following recurring problems that the NRC has observed at the Supply System:
inadequate evaluation and response to industry issues, inadequate management and quality oversight, poor interorganizational communication, and failure to adhere to procedures.
Three apparent violations were identifie ll.
Exit Meetin 30703 On August 3, 1993, the inspector met with licensee representatives (as noted in Paragraph 1) to discuss inspection findings related to the violation of TS 3.6.2.3,
"Suppression Pool Cooling."
During the exit meeting the licensee noted the following:
(1) the Supply System identi-fied the violation during the investigation of an associated problem; (2) the Supply System recognized that there was a design deficiency in the RHR system because they had been discussing an associated problem with another utility, and this discussion represented improved perform-ance with respect to previous isolation from the rest of the industry; and (3) the Supply System's evaluations and corrective actions have been prompt and thorough following their recognition of the design deficiency.
The licensee did not identify as proprietary any of the materials discussed with or reviewed by the inspectors during this inspection.