3F1197-36, Provides Addl Info Re Four Items Identified During Recent SSFI 50-302/97-14 on 971006-1024 at Plant,Unit 3.Fourth Item Requiring Addl Info Involves Qualification of Large Bore safety-related Piping & Piping Support
| ML20198T202 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/10/1997 |
| From: | Holden J FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 3F1197-36, 50-302-97-14, NUDOCS 9711140225 | |
| Download: ML20198T202 (17) | |
Text
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.. Florida Power-EoAoa.W e J* .c; .L.-n November 10,1997 3Fil97-36 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washingten, DC 20555-0001
Subject:
Additional Information for SSFI Items, SSFI dated October 6,1997 through October 24,1997, NRC Inspection No. 50-302/97-14
Dear Sir:
Consistent with Florida Power Corporation's (FPC's) commitment at the October 31,1997, NRC Restart Panel Meeting in Crystal River, this correspondence provides additional information regarding four items identiUed during the NRC's recent Safety System Functional Inspection (SSFI) at Crystal River Unit 3 (CR-3). Specifically, on three of the four items, questions were raised by the NRC during the October 24,1997, exit meeting for a SSFI conducted October 6 through 24,1997. Three items from the exit meeting included: (1) the response to Generic Letter 88-17, (2) the crosstic of the liigh Pressure injection / Makeup discharge header, and (3) the FPC Generic Letter 96-06 response that installed containment penetration thermal expansion chambers. The fourth item requiring additional information involves the qualification of large bore safety-re;ated piping and piping supports.
Attachments A through D to this letter address each of the above-referenced items, respectively.
Attachment E contains the commitments made in this letter. FPC will provide appropriate corrective action and closure documentation within the CR-3 corrective action program, as necessary.
If you should have any questions on this information, please contact Mr. David Kunsemiller, Manager, Nuclear Licensing at (352) 563-4566 Sincerely, YV f
John J. lloiden, Director fW) )
Site Nuclear Operations JJil/Irm 9711140225 971110 I PDR ADOCK 05000302 Attachments a pon xc: Regional Administrator, Region 11 Senior Resident inspector
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Johns P. Jaudon CRYSTAL RIVER ENERGY COMPLEX: 16760 W. Power une Street
- Crystal River, Florida 34428 6708 * (352)795-6488 A FiorMe Progowss Conpeny
. U.S. Nuclear Regulatory Commission Attachment A 3F1197-36 Page 1 of 3 ATTACilMENT A FLORIDA POWER CORPORATION NRC INSPECTION NO. 50-302/97-14 SUPPLEMENTAL INFORMATION FOR ITEM NO.1 FPC GENERIC LETTER 88-17 RESPONSE APPEARS INCOMPLETE ITEM Both trains of the instrumentation utilized at CR-3 to monitor Reactor Coolant System (RCS) water level during a reduced inventory condition use the same RCS taps. The NRC questioned this arrangement as contradicting the guidance in Generic Letter (GL) 88-17.
BACKGROUND On October 17, 1988, the NRC issued GL 88-17, less of Decay licat Removal. In it, the NRC recommended that the nuclear plant licensees:
Provide at least two independent, continuous RCS water level indications whenever the RCS is in a reduced inventory condition. Water level indications should be periodically checked and recorded by an operator or automatically and continuously monitored and alarmed. Water level monitoring should be capable of being performed either:
(a) by an operator in the control room (CR), or (b) from a location other than the CR with provision for providing immediate water level values to an operator in the CR if significant changes oc;ur.
Observations should be recorded at an interval no greater than 15 minutes during normal conditions.
Additionally, in GL 88-17, Section 3.1.2.1, the NRC states:
We recognize that it may be difficult to provide independence in isolated instances.
Consequently, if the recommendation for independence resul's in an unnecessary hardship, we will consider compensatory means. For example, if a common tap is used for the liquid leg, a means of periodic draining or flusning capable of detecting blockage might be proposed as a means of diminishing the potential impact of dependency.
~
. U.S. Nuclear Regul: tory Commission Attachment A 3F1197-36 Page 2 of 3 DISCUSSION-By letter dated January 4,1987, FPC committed to the following for level indication:
FPC , will complete a modification to provide single channel RCS water level instrumentation with indication in the control room in the range needed to monitor RCS water level in Refuel 7 beginning in Fall 1989. A portion of this modification was completed during the outage which will provide such capability during limited situations (RV head removed). FPC will evaluate the need for and best means to achieve a.:
additional channel of RCS level instrumentation. Installation will be completed by Refuel 8 presently for Fall 1991.
FPC Installed RC-201-LT and RC-202-LT, which provide both indication and alarms in the Main Control Room (MCR). This instrumentrion was added as part of Modification Approval Record (MAR) 87-07-28-02 and 87-07-28-03. As stated in FPC letter dated June 29,1990, the installation of the instruments was completed during Refuel 7. These modifications described the installation of this instrumentation as providing complete electrical independence. consistent with the intent of GL 88-17. It should be noted that even though this instrumentation is electrically independent, it utilizes common taps for both the reference and variable legs.
Additionally, as defense ir depth, FPC has a visual indication using tygon tubing connected to the RCS. This level indication is typically available in the MCR via video monitor or monitored by a dedicated operator. The tygon tubing was installed per MAR 79-04-07 and later enhanced by M AR 92-10-16-01. The level sensing line going to the tygon tubing utilizes the same RCS tap as the remote RCS level instrumentation (RC-201/202-LT). Plant procedure OP-301 addresses when and how both the tygon tubing and the remote RCS level instrumentation are placed into service.
The tygon tubing is the first instrument aligned. During this activity, the level sensing lines going to RCV-198/199/200 are flushed / drained for 15 minutes. At this point, the once through steam generators (OTSGs) and the pressurizer gas spaces are cross-tied together via the Reactor Coolant Drain Tank with a few pounds of nitrogen on the system. Then, the remote RCS level instrumentation is placed in service (RC-201/202-LT). In this step, Operations notifies I&C to place the instrumentation in service per SP-195, Operation of Level Transmitters with RCS Blanketed with Nitrogen. After the electronic strings are calibrated per SP-195, the transmitters are vented and the reference leg is filled / flushed with demineralized water.
Guidance in OP-301 describes how to flush the level sensing lines. A change to this procedure is being developed to drain the tubing runs instead of flushing since operational experi:nce has revealed that blockage has not occurred but air has bc:n found in the sensing lines. This action is consistent with GL 88-17 guidance.
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... U.S. Nuclear Regulatory Commission- Attachment A '
3Fil97-36 Page 3 of 3 CONCLUSION
-. Based upon the ~ installation _ of the' electrically independent instrumentation, .the
- flushing / draining utilized to compensate for the common taps, and the use of the MCR video displayed tygon_ tubing indication as defense'in depth, FPC concludes that its actions are consistent with the intent of GL 88-17.
REFERENCES 1.. M AR 87-07-28-02; Reactor Vessel Remote Level Indication 2.- MAR 87-07-28 03: Additional Narrow Range RV Level Indication
- 3. MAR 79-04-07, RC level Indication
- 4. MAR 92-10-16-01, RC Level Venting
- 5. Generic Letter 88-17, Loss of Decay lleat Removal, Dated October 17,1988
- 6. FPC Response, " Response to Generic Letter 88-17, - Loss of Decay Heat Removal,"
(3F0189-02), dated January 4,1989
- 7. FPC Response, " Programmed Enhancements For Generic Letter 88-17, Loss of Decay lleat Removal" (3F0690-15), dated June 29,1990_-
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U.S. Nuclear Regulatory Commission Attachment B 3Fil97-36 Page 1 of 3 ATTACIIMENT B FLORIDA POWER CORPORATION NRC INSPECTION NO. 50-302/97-14 SUPPLEMENTAL INFORMATION FOR ITEM NO. 2 CROSSTIE OF TIIE IIIGil PRESSURE INJECTION / MAKEUP DISCIIARGE HEADER ITEM The discharge piping for the liigh Pressure injection / Makeup Pumps are connected by a common crosstic header containing two normally-open manual valves (MUV-4 and MUV-8),
and two normally-open motor-operated valves (MUV-3 and MUV-9), with MUV-9 r.ormally deenergized. NRC inspectors questioned whether a postulated line break at any point in the discharge piping could compromise the capability to safely shut down the plant, especially since MUV-3 and MUV-9 are not stroke-time tested in the closed direction as part of the ASME Section XI Pump and Valve Inservice Testing Program.
BACKGROUND The original configuration of the liigh Pressure injection / Makeup Pumps discharge piping consisted of four manual valves (MUV-3, MUV-4, MUV-8 and MUV-9), with valves MUV-3 and MUV-4 being normally closed. This allowed liigh Pressure Injection Pump 1B (MUP-IB) to serve as the normal makeup pump while providing separation between High Pressure injection Pumps I A and IC (MUP-1A and MUP-lC) in the discharge piping to maintain two independent Emergency (' ore Cooling System (ECCS) flowpaths. The required flowpaths consisted of two separate pumps each with two separate injection lines to the Reactor Coolant System.
By letters dated April 24, 1978, and May 5,1978, FPC submitted Licensee Event Report (LER) 78-019/0lT-0 and follow-up information identifying that a reanalysis of Small Break Loss-of-Coolant Accident (SBLOCA) scenarios had resulted in the determination that the current ECCS configuration, and specifically High Pressure Injection flow delivery capability, was not adequate to mitigate the effects of a specific spectrum of small breaks. By letter dated June 14,1978, FPC submitted a justification for continued operation which included addition of operator actions to emergency operating procedures.
By letter dated July 21,1978, and subsequent correspondence, FPC proposed configuration changes, including hardware and procedural modifications, and the addition of simple operator actions to resolve the SBLOCA concerr.s. These actions resulted in the ability to crosstie either MUP-1A or MUP-lC to all four injection lines. By letter dated May 29,1979, the j NRC issued a Safety Evaluation Report approving FPC's proposed solution. Valves MUV-3,
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,. U.S. Nuclear Regulttory Commission . Attachment B ;
3Fil97 Page 2 of 3 l
. 'f MUV-4, MUV-8 and MUV-9 were thereafter left open as part of this permanent solution to
'the SBLOCA issue, in - 1980, valves MUV-3 ~ and - MUV-9 were modified to install . non-safety related_ motor operators to ensure the capability to remotely isolate leaks that may occur in the High Pressure injection / Makeup Pumps discharge piping. Also, both valves were added to- the: ASME Section XI_ Pump and Valve Inservice Testing Program, which resulted in stroke-time testing- 1 i
in the closed direction on a quarterly basis, and remote position verification testing every two years, in mid-1991, FPC reassessed this stroke-time testing methodology and determined that .
i since the valves were passive, thev did not have to be stroke-time tested. -However, as a conservative measure, FPC contir.M to perform remote position verification testing to ensure the valves could be remotely verified open (their required safety function position).
Between 1982 and 1984, during performance of the Safe Shutdown Analysis and Appendix R i Fire 5tudy, FPC determined that valve MUV-9 was required to be de nergized to ensure adequate makeup capability was preserved during various postulated fire scenarios which might spuriously close the valve. Therefore, based on requirements of the 10 CFR 50, Appendix R Fire Study, valve MUV-9 is presently maintained open and deenergized.
l DISCUSSION ,
r A review of the circumstances surrounding the decisions made to open valves MUV-3 and MUV-9 and- to maintain valve MUV deenergized, including review of the docketed correvondence discussed above, leads FPC to conclude that a postulated line break at any point in the discharge piping would not compromise the ability to safely shut down the plant.
flowever, FPC recognizes that the capability to safely- shutdown the plant, given a leak or break in the I-ligh Pressure Injection / Makeup Pumps discharge piping, is best achieved by assuring the reliability and availability of valves MUV-3 and MUV-9 to close upon demand.
L in fact, current operating procedures have been developed ba,ed on using valves MUV-3 and
. MUV-9, either remotely or locally, to isolate leaks or breaks in the High Pressure injection / Makeup Pumps discharge piping.
1 F - CONCLUSION
- Based on the fact that valves MUV-3 and MUV-9 were stroke-time tested in the closed direction until'1991, have been full-stroke exercised at least every two_ years since that time,
- and are operated at least quarterly during the performance of High Pressure injection / Makeup Pump testing, there is reasonable assurance that the valves are capable of performing their design function, in addition requirements for quarterly stroke-time testing in the closed direction for valves MUV-3 and MUV-9 are being added to the ASME Section XI inservice Testing Program, and will be added to surveillance procedures SP-340C and SP-340F prior to ,
March 31,1998. ~1n the interim, operability and availability are as stated above, and no further action is required.
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. LU.S. Nuclear _ Regul tory Commission:
Attachment B ,
3Fil97 Page 3 of_3 P
- REFERENCES
~ 1978, "LER 78-019/0lT-0: ECCS Small Break Analysis Concerns"
'2. 3F0778-05; Letter from FPC to NRC dated July 21, -1978, "ECCS Small Break '
Analysis Permanent Solution"
- 3. - 3N0579-08, Letter from NRC to FPC dated May 29,1979, " Safety Evaluation Report Approving Small Break LOCA Analysis Permanent Solution"
Corrective Actions for ECCS Small Break Analysis Concerns" 5.. NPSE91-0183, Interoffice Correspondence dated June 4, .1991, "MUV-3 Packing Leak, and MUV-3 and MUV-9 Inservice Testing Requirements"
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- _U.S. Nuclear Regulatory Commission Attachment C
- 3F1197-36 Page 1 of 5 ATTACIIMENT C FLORIDA POWER CORPORATION
. NRC INSPECTION NO. 50-302/97-14 SUPPLEMENTAL INFORMATION FOR ITEM NO. 3 ADDITION OF CONTAINMENT PENETRATION TIIERMAL EXPANSION Cli AMBERS, RESPONSE TO GL 96-06 ITEM
-Generic Letter 96-06 requested that containment penetration piping be evaluated to determine if thermally induced overprcurization could jeopardize the ability of accident mitigating systems to perform their safety functions or could lead to a breach of containment integrity via bypass leakage, in response, FPC modified Containment Penetration No. 333 piping (Makeup and Purification System letdown line) by adding a thermal expansion chamber (MURS-1) outside containment between the containment penetration and valve MUV-49. With regard to this action, NRC inspectors questioned whether this modification: (1) meets CR-3 Design and Licensing Basis, and (2) created the possibility of a new type of accident not described in the Final Safety Analysis Report (FSAR).
BACKGROUND By letter dated September 30, 1996, the NRC issued Generic Letter 96-06, which requested, in part, that all licensees evaluate containment penetrations that may be susceptible to overpressurization due to thermal expansion of trapped fluids after containment isolation in response to a design basis accident. The NRC provided additional clarification in a November 22, 1996, letter based on the results of an October 29, 1996, working meeting with industry representatives.
By letter to the NRC dated December 13, 1996, FPC added the completion of Generic Letter 96-06 related actions as part of Restart Issue D-8. By letters dated January 27,1997, April 30, 1997, and October 17, 1997, FPC provided responses to Generic Letter 96-06 which included: (1) a description of the thermal expansion chamber concept and design, including the use of rupture discs internal to the therrnal expansion chambers, to be incorporated during the required modifications; (2) an explanation for installing the thermal expansion chambers outside of the Reactor Containment Building; and (3) a listing of the affected containment penetrations.
Modification Approval Record (MAR) 96-10-04-01 was developed and approved to install the thermal expansion chamber and rupture disc for Containment Penetration No. 333. Upon
. U.S. Nuclear Regulatory Commission Attachment C 3F1197-36 Page 2 of 5 completion, the specific design basis information for this MAR will be incorporated in Topical Design Basis Document TDBD-9/3, " Containment isolation."
DISCUSSION FPC is not required to comply with the current 10 CFR 50, Appendix A, General Design Criterion 55. CR-3 received its construction permit on September 25, 1968. Section 1.4 of the FSAR describes the applicable General Design Criteria for the design of CR-3. These General Design Criteria are based on the draft 10 CFR 50.34, Appendix A, General Design Criteria which had been proposed by the Atomic Energy Commission at the time the Crystal River Unit 3 Preliminary Safety Analysis Report was issued. Consistent with SECY-92-223,
" Resolution of Deviations identified During the Systematic Evaluation Program," dated September 8,1992, the NRC determined that piants with construction permits issued prior to May 21,1971, must satisfy their original General Design Criteria, as previously committed to the NRC, and that exemptions from the current 10 CFR 50, Appendix A, General Design Criteria were not necessary. Based on its construction permit, dated September 25, 1968, SECY-92-223 is applicable to Crystal River Unit 3.
FSAR Section 1.4 contains CR-3 specific commitments and discussions on how the draft 10 CFR 50.34, General Design Criteria are met. The applicable General Design Criteria are described in FSAR Section 1.4.51, " Reactor Coolant Pressure Boundary Outside Containment," and Section 1.4.53, " Containment Isolation Valves."
Question i The NRC's first question challenged whether the modification to install a thermal expansion-chamber on the letdown line piping meets CR-3 Design and Licensing Basis. FSAR Section 1.4.53 specifies that containment penetrations that require closure for the containment isolation function shall be protected by redundant valving, and references FSAR Section 5.3.2 for further details. FSAR Section 5.3.2 describes the allowed containment isolation valve combinations and defines the four types of containment penetrations employed at CR-3. Type I containment penetrations employ one inside and one outside containment isolation valve for isolating lines that are connected directly to the Reactor Coolant System (RCS) following an accident. FSAR Table 5-4 currently identifies Containment Penetration No. 333 as a Type I penetration with automatic containment isolation valves both inside and outside containment.
The Type I penetration boundary before and after installation of this modification, which is consistent with the CR-3 Design and Licensing Basis, includes the piping between valves MUV-40, MUV-41, MUV-505 (inside containment) and MUV-49 (outside containment). The original 3/4" drr.in line connection in the piping between containment and valve MUV-49 was isolated with a normally closed and capped valve. Installation of this modification, replaces the 3/4" drain line connection and normally closed and capped isolation valve with 3/8" tubing between this connection and the expansion chamber. The expansion chamber has a drain valve which is normally closed and capped. Although the design of the thermal expansion chamber represents an extension of the original drain line pressure boundary (including a normally
. U.S. Nuclear Regulatory Commission Attachment C 3F1197-36 Page 3 of 5 closed and capped test connection), the penetration will continue to meet the requirements for a Type i penetration boundary as defined by FSAR Section 5.3.2. It should als be noted that the rupture disc internal to the thermal expansion chamber is not credited as a containment boundary, Regarding the potential effect of an extended length of tubing, FPC recognizes that, originally, MUW49 was located as close as practical to the containment to minimize the effects of a high energy line break should it occur between MUV-49 and containment. Although the 3/8" tubing runs now approximately twelve feet between the letdown piping and the thermal expansion chamber, the tubing is small enough that even if it should become completely severed at any location, the resultant leakage rate would not exceed the capacity of one makeup pump. Therefore, the importance of locating the thermal expansion chamber as close to containment as practical is minimal. It should also be noted that locating the letdown line thermal expansion chamber (MURS-1) as far as twehe feet away from the letdown line piping was based on reserving sufficient space for a major modification to the high pressure injection system to be installed in a future outage. Relocation of the expansion chamber closer to containment would require later relocation when this major modification is installed.
The decision to locate the thermal expansion chambers outside containment was made to allow surveillance of the rupture discs on a periodic schedule to assure their integrity without having to enter the Reactor Building during power operations. If a rupture disc degrades, FPC would have the option of closely monitoring the condition or replacing the rupture disc.
All of the new components for this modification have been classified as " Nuclear" (N1) in accordance with FSAR Section 1.3.2.12; were derigned in accordance with the requirements of USAS B31.1.0-1967; and were fabricated, shop-tested, erected and inspected to B31.7-1969, including Code Case B31-83 (October 1970). This corresponds to the same requirements as the system to which they have been attached. Additionally, as discussed previously, the tubing, thermal expansion chamber, and normally closed and capped drain valve form a passive containment isolation boundary that is not susceptible to any active failures. Also, as previously discussed, the tubing is sufficiently small that any passive failure, including complete severance of the tubing or rupture of the expansion chamber, would result in a leakage rate that would not exceed the capacity of one makeup pump.
Question 2 The NRC's second question involved whether thT ification created the possibility of a new type of accident not described in the FSAR. . ~ AR Section 9.1.2.5 discusses leakage considerations for the Makeup and Purification System. As described in this section, rupture of small bore piping or tubing beyond the first isolation valve or outside the reactor building is not considered a Loss of Coolant Accident, and therefore is not required to be analyzed as a Design Basis Accident.
For the worst-case (double-ended) break of the 3/S" tubing, analyses performed by FPC have determined that the resultant leakage rate would be approximately 50 to 75 gpm. These leakage rates are well within the flow capability of one makeup pump. Therefore, Reactor
4 s '
. O.S. Nuclear Regulatory Commission Attachment C 3F1197-36 Page 4 of 5 Coolant System pressure is expected to be unaffected by this event, and an automatic reactor or turbine trip will not occur and no other automatic Engineered Safeguards Features will actuate. In addition, operator actions to successfully terminate the leak could be performed from the Control Room before any appreciable leakage had occursed. Therefore, a rupture of the expansion chamber components would be classified as a leak event, and would not be considered a Design Basis Accident requiring additional accident analyses.
CONCLUSION FPC has designed and installed the MURS-1 thermal expansion chamber and rupture disc in accordance with the same requirements as the letdown system to which it has been attached.
In addition, all other expansion chambers meet the same design and installation requirements as the systems to which they are attached. The design of the associated modification meets the diaft General Design Criteria specified in FSAR Sections 1.4.51 and 1.4.53, as well as requirements specified in FSAR Chapter 5 for containment penetrations.
In addition, this modification does not create the possibi!ity of a new accident since all postulated failure modes would result in an isolable leak, and would not result in actuation of any Engineered Safeguards Features required to protect the reactor or containment integrity.
Therefore, this modification meets all of the applicable FPC Design and Licensing Basis requirements, and no further action is required.
REFERENCES
- 1. 3N0996-18, Letter from NRC to FPC dated September 30,1996, "NRC: Generic Letter 96-06: Assurance of-Equipment Operability and Containment Integrity During Design-Basis Accident Conditions"
- 2. 3 Nil 96-23, Letter from NRC to FPC dated November 22,1996, " Meeting with NEl and Licensees to Discuss ' Generic Letter (GL) 96-06, ' Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions'"
- 3. 3F1296-05, Letter from FPC to NRC dated December 13,1996, " Crystal River Unit 3 Restart Panel and Action Plan"
- 4. 3F0197-05, Letter from FPC to NRC dated January 27, 1997, "120 Day Response to NRC Generic Letter %-06, ' Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident' Conditions'"
- 5. 3F0497-33, Letter from FPC to NRC dated April 30,1997, " Supplemental Resporse to NRC Generic Letter 96-06, ' Assurance of Equipment Operability anc' Containment Integrity During Design-Basis Accident Conditions'"
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, U.S. Nuclear Regulatory Commission Attachment C L 3F1197-36 Page 5 of 5 6.- 3F1097-05, Letter from FPC to NRC dated October 17,1997, " Final Report on NRC ,
Generic Letter 96-06, ' Assurance of Equipment Operability and Containment Integrity. ,
. During Design-Basis Accident Conditions'"
- 7. Temporary Change (TC) No. 548 to TDBD 9/3, " Topical Design Basis Document for.
Containment Isolation"
- 8. Framatome Technologies- Engineering Information Record 51-5000846-00, "CR-3 Makeup System SSFI Question 59, Follow-up Number 1" i=
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. U.S. Nuclear Regul: tory Commission Attachment D 3Fil97-36 Page 1 of 4 ATTACIIMENT D FLORIDA POWER CORPORATION
.NRC INSPECTION NO. 50-302/97-14 SUPPLEMENTAL INFORMATION FOR ITEM NO. 4 QUALIFICATIONS OF PIPING AND PIPING SUPPORTS ITEM During the NRC SSFI, the NRC inspectors requested more information concerning incomplete records pertaining to large bore piping analysis and support calculations.
HACKGROUND in mid-19%, FPC recognized a potential concern with availability or lack of adequate engineering calculations to support the qualification of large bore safety-related piping and pipe supports. To address this concern, FPC determined that a programmatic assessment of our documentation and as-built condition was needed. In October 1996, FPC brought in an independent reviewer (Wais and Associates, Inc.) to evaluate the adequacy of existing documentation to support current licensing and design basis of large bore piping and associated supports, and to ensure reliable plant operation. The assessment included a review of the analysis documentation, interviews with key personnel, and a walkdown of specific areas of the plant.
The final report identified scveral discrepancies involving inconsistent application of methodologies and reconciliation with the as-built conditions and a concern regarding inadequate control of documentation. The assessment concluded that individually, the issues have minor safety significance, however, due to the number of findings, and the lack of adequate documentation, the collective significance could constitute a safety concern.
Therefore, FPC initiated a Precursor Card to address the specific issues identified in the Wais Report and to assess the extent of condition.
DISCUSSION
- Precursor Card (PC) 97-0048 was issued to document the Wais Report findings. A Suspected Design Basis issue evaluation was conducted for PC 97-0048 and concluded that no Design Dasis Issue existed. This evaluation determined that notwithstanding the Wais and FPC fmdings, the systems, structures and components were still capable of performing their safety
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. U.S. Nuclear Regulatory Cc.mmission Attachment D 3F1197-36 Page 2 of 4 function. For piping systems, this is interpreted to be the loss of pressure boundary integrity and/or the ability to deliver rated How, in part, this conclusion is based on historical data. Also, specific portions of piping systems have been requalified or evaluated in the recent past. While a number of code compliance issues have been identified over the years, subsequent assessments and analyses have not identified a piping system that was incapable of meeting its design basis requirements. Some of the qualification efforts addressed issues that are similar to those discussed in the Wais Report. Also, no items were found to represent an operability concern. Where code compliance concerns were identified, they were promptly corrected.
To further ensure that no design basis issues existed, and to evaluate the collective significance of all of the piping and pipe support related issues, FPC initiated a second third party review (Dr. John D. Stevenson). This assessment concluded that "...there is no safety concern with respect to the "as designed" and "as installed" safety-related large bore pipe at CR-3." The basis for his conclusion included over three weeks of CR-3 documentation review and walkdowns as well as referencing past industry efforts. The following are excerpts from the basis of his conclusion:
- 1) The U.S. NRC sponsored the Systematic Evaluation Program (SEP). The purpose of the SEP was to evaluate the seismic adequacy of structures, mechanical and electrical distribution systems and mechanical and electrical distribution components in older operating nuclear power plants. The results of the SEP ultimately led to the Unresolved Safety Issue A-46 which excluded piping due to the lack of safety concerns (except for piping flexibility between structures). This conclusion regarding the seismic safety of installed safety-related piping in older nuclear power plants was later confirmed in NUREG/CR-4334.
- 2) A comparison of the way CR-3 piping is supported relative to the SEP plants and approximately 20 other A-46 plants identified no safety issues. Additionally, it was noted, CR-3 has the second lowest probabilistically defined mean design basis ground acceleration levels of all nuclear power plants in the U.S. CR-3 has a 0.65x10~'/yr probability (NUREG-1488) for a mean peak ground acceleration of 0.lg. This probability value is more conservative by a factor of more than 2.0 as compared to most nuclear plants operating in the U.S. today and is less than the NRC's A-46 reevaluation criteria of 10"/yr.
- 3) CR-3 piping design, because of the excessive conservatism contained in both the definition of SSE equivalent zero period ground acceleration, the design basis floor spectra and allowable piping stresses, results in failure probabilities which are several orders of magnitude less than currcatly required by ASME and NRC piping design criteria and has therefore resulted in a conservatively safe design.
The conclusions from the above assessment also substantiate that the safety-related large bore piping and supports will perform their intended safety function. As previously stated, the majority of the issues identified to date involve missing or lack of qualification documentation l
l
o U.S. Nuclear Reguletory Commission Attachment D 3F1197-36 Page 3 of 4 to support current licensing basis and design basis requirements. Previous evaluations of piping for plant modifications such as increased valve / operator weights and evaluations of previously identified degraded conditions, such as inoperable supports and non-conservative analytical errors, have not identified any operability concerns.
FPC plans to perform a comprehensive and rigorous inspection and revalidation /requalification of CR-3 large bore safety-related piping and pipe supports. A project scoping report is currently in development that will denne the scope, duration and cost of this effort. This effort will address the concern with lack of adequate engineering documentation as well as address the various technical issues associated with the Wais Report.
The specific Pipe Stress and Support calculations to support the resolution of identified issues are being developed. In addition, a project plan for establishing the programmatic tasks (e.g.,
computer program purchases, master list development, reviewing licensing commitments, project staffing, etc.) is underway. FPC currently anticipates that the program will take approximately 4-6 years (2-3 fuel cycles) to implement. This program will be modeled similar to the " Isometric Update Program (IUP)" recently completed at Arkansas Nuclear One Units 1 and 2. That IUP began in 1987 and ended with the last modification package being installed spring of 1997.
CONCLUSION All currently identified code compliance issues associated with large bore piping and supports will be corrected prior to restart. Modifications are being implemented where necessary. Due to the identification of past code compliance issues, FPC plans to submit a voluntary report summarizing our findings. Additionally, any emergent findings will be assessed on a case by case basis.
As discussed above, FPC plans to implement a comprehensive and rigorous inspection and revalidation /requalification project of large bore safety-related piping and pipe supports. This project will span the next 4-6 years due to the significant extent of effort required. FPC believes this plan is commensurate with the level of potential safety concern and is similar in method, scope and duration with that performed by other utilities who have addressed this issue in the past. To date, no operability concerns have been identified based on the above issues at CR-3. flowever, if future code compliance problems are identified during the revalidation effort, they will be documented and operability evaluations performed in accordance with our corrective action program (Precursor Cards).
Based up(m our assessment and results of the assessments of third party reviewers, FPC concludet that no condition exists that would chslienge the ability of systems, structures or components to perform their safety function of preventing or mitigating design basis events.
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--,- U.S. Nuclear Regulitory Commission - Attachment D 3Fil97-36 : Page 4 of 4 REFERENCES ,
- 1. Wais and Associates, Inc. . Report Number %04-002 dated October 28,1996,
" Evaluation of Pipe Support Documentation for Crystal River 3" '
- 2.- John D. Stevenson, Consulting Engineer, Report Number 97.174-1 dated October 30, .
1997, " Overview cf Safety Related Large Bore Piping and Piping Support Design and Construction Currently Existing at Crystal River-3 Nuclear Power Plant"
- 3. FPC Precursor Card (PC) 97-0048 -
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_U.S. Nuclear Regulitory Commission . Attachment E
- 3Fil97-36' Page1ofI !
ATTACHMENT E List of Regulatory Commitments ,
The following table identifies those aciions committed to by Florida Power Corporation in this >
document.7 Any other actions discussed in the submittal represent intended or planned actions by Florida Power Corporation. They are described to the NRC for the NRC's information and are not regulatory commitments.' Please notify the Manager, Nuclear Licensing of any
- questions regarding this document or any associated regulatory commitments.
Section Commitment Due Date ite m 2 FPC management commitment to add quarterly March 31,1998 :
stroke time testing in the closed direction for valves MUV-3 and MUV-9 to the ASME Section XI e Inservice Testing Program.
Item 4 FPC is committing to do a comprehensive and A project scoping ,
rigorous inspection and revalidation /requalification of report is currently in large bore safety-related piping and pipe supports at development that will CR-3. This project will span the next 4-6 years. define the scope, ,
duration and cost of this effort.
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