ML20127D218
| ML20127D218 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/04/1992 |
| From: | Eliason L NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9209140005 | |
| Download: ML20127D218 (11) | |
Text
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d' Northem States Power Company 414 Nicellot Mall
%nneapohs W nosota 55401 1927 leicphore (612) 330 ZOO September 4, 1992 10 CFR part 2 Appendix C U S Nuclear Regulatory Comrnission Attn:
Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PIANT Docket Nos. 50 282 License Nos. DPR 42 50-306 DPR-60 Response to Notice of Violation NRC Inspection 3cport Nos. 282/92011(DRP) and 306/92011(DRP)
Auxiliary Feedwater Pu:m Surveillance Test 1DE Your letter of August 5, 1992, which transmitted Inspection Report Nos.
282/92011(DRP) and 306/92011(DRP), requested responses to both an unresolved item and a violation, Our response to Unresolved Item (282/92011 01(DRS);
306/92011 01(DRS)) is included as an attachment to this letter.
The following is offered in response to the violation.
Violation During an NRC inspection conducted on May 27 through July 20, 1992, a violation of NRC requirements was identified.
11 accordance with the
- Ceneral Statement of Policy and Procedure for NRC Enforcement Actions,
10 CPR Part 2, Appendix C, the violation is listed below:
Technical Specification 4.8.A,8 requires that at 1 cast once every 18 months during shutdown, each auxiliary feedwater purnp shall be tested to verify that each pump starts as designed automatically upon receipt of each auxiliary feedwater actuation test signal. Actuation test signals are provided to the auxiliary feedwater pumps by the circuitry that senses that both associated main feadwater pump breakers are open and by both associated steam generator low water level circuits.
Contrary to the above, prior to June 4, 1992, the licensee had never tested the auxiliary feedwater pumps to verify that each auxiliary feedwater pump would start automatically upoe ceceiving an actuation signal from each associated steam generator low water level circuit or that each auxiliary feedwater pump would start automatically upon
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USKRC September 4, 1992
!"E" Nodhern States Power Company receiving an actuation signal from the circuity that senses that both associated main feedwater pump breakers are open.
This is a Severity Level IV Violation (Supplement 1).
Response to the Viola _t12D Backcround Each unit has two auxiliary feedwater pumps, one motor driven and one turbine-driven.
There are 5 automatic start signals for each of the pumps.
Technical 1 Specification 3.5, Tabic 3.5 3, Item 3 lists 4 automatic starts.
They are:
Steam Generator Low Lov Water Level low low icvel in either steam generator will start both pumps Undervoltage.on 4.16 KV Buses 11 and 12 (21 and 22 Unit 2)(Start turbine driven pump only)
-undervoltage on both buses will start the turbine driven pump Trip of Main Feedwater Pumps - trip of both main feedwater pumps will start both auxiliary feedwater pumps Safety Injection -L a safety injection signal will start both pumps There is another automatic' start signal. AMSAC, which is not required by
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Technical Specifications.
Since initial plant licensing, the safety injection start signal has been tested during the " integrated SI test" which is conducted each refueling. No other formal testing of automatic pump starts was done, nor was any_ required by, Technical Specifications prior to March 2, 1981.
On March 2, 1981 Technical Specifications were amended.
The amendment made several miscellaneous changes.
One of those changes'was the addition of i
specification 4.8.A.8.
This new specification-required that each auxiliary L
feedwater pump be started by each automatic. start signal every 18 months.
l This new specification was not implemented due to. inadequate review by plant
- staff, t
18, 1988, NRC Information Notico 88-83 informed licensees of L
Jon October inadequate testing of relay contacts in-safety related logic. systems, As a -
l result of review of the Notice, recommendations were made to. increase.the scopalof auxiliary-feedwater pumpLtesting.
The recommendations were mado based onTengineering judgment;. review of compliance with Technical
~
L Specifications was not done in conjunction with review of the Notice.
At the-time _of_._the. inspection, procedures were being developed to test the undervoltage and steam' generator low low water level starts.
No action.was-
-being~taken to test the trip from."both main feedwater pumps off" since this
- start 1was viewed-as backup or anticipatory to the steam generator low-low water 11evel start.
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USNRC September 4, 1992 I
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Northern States Power Cornpany On March 6, 1992, it was determined that the requirement for annual full flow testing of turbine driven auxiliary feedwater pumps in accordance with Technical Specification 4.8.A.2. was not being met, and Unit 1 Licensee Event Report 92 004 was issued. As corrective action for the event, we committed to "a comprehensive review of testing requirements...to ensure the Technical Specification surveillances are being met."
As a result of that review, other auxiliary feedwater system testing deficiencies were identified in a letter to the Operations Committee dated March 16, 1992.
On March 26, 1992, the Operations Committee considered the question of operability of the auxiliary feedwater pumps based on the testing deficiencies identified. The Committee raised the operability concern, and asked line management to determine if a surveillance requirement had indeed been missed. Unfortunately, this Committee action was not recorded in the meeting minutes. When the resolution was brought back to the operations Committee, the Committee concluded that a surveillance requirement had been missed, and immediately declared all 4 auxiliary feedwater pumps inoperable.
A Supplemental Licensee Event Report was issued.
Renson for Violation As a result of our detailed review of this violation, three ceases were identified:
1.
The March 2, 1981, license amendment was not fully implemented because of inadequate review by plant staff.
No formal implementation procedure for license amendments existed at that time.
2.
As a result of review of NRC Information Notice A8 83, recommendations were made to increase the scope of auxiliary feodwater pump testing.
But the recommendations were made based on engineering judgment; review of compliance with Technical Specifications was not done in conjunction with review of the Notice.
3.
Following the comprehensive review of auxiliary feedwater system testing, the operations Committee failed to press for resolution of the operability question in a timely manner. The Committee raised the operability issue, but did not document its actions, nor did it assign the proper priority to its resolution.
The resolution process, which should have been prompt, took three months.
(Hundreds of operability questions are raised routinely through the Design Basis Reconstitution effort; nearly all of these are resolved with a positive outcome without the need for Operations Committee review. We believe that the Operations Committee expectation was that this item, too, would have a positive outcome, and for that reason did not press for a quick resolution.)
Corrective Stens that have been taken and Results Achieved On June 4,1992 all 4 auxiliary feedwater pumps were declared inoperable.
Testing to fulfill the requirements in Technical Specification 4.8.A 8 was satisfactorily performed and the pumps were declared operable.
4 USNRC September 4, 1992
- U" Northern States Power Company Corrective Stcos to Prevent Further Violationg
- 1. - The Technical Specification Change Review Committee, a-subcommittee of the Operations Committeo, is in place to review and implement license amendments.
This committee was formed in 1984.
We believe that if a License Amendment similar to the March 2, 1981, amendment were issued today, the amendment would be properly and completely implemented.
2.
New testing procedures-are being generated that will fulfill the requirements of Technical Specification 4.8.A.8.
These procedures will be in place when they are needed at the upcoming refueling outages for each unit.
- 3.-
The plant manager has reviewed this event with the plant staff and with the Operations Committee. A heightened sensitivity to operability issuet now exists, When operability issues arise now, the Committee is likely to assume inoperability until shown otherwise.
14.
Action has been initiated to include discussion of the event during continuin6 Engineer and Technical Staff training. This training will-serve to increase awareness of operability issues.
5.
-The Operating Experience Assessment form has been changed to insure that reviews of such iters as Information Notices include operability determinations.
6.
The comprehensive review of the Technical Specification surveillance requirements referenced (in the Background section) is ongoing.
A pilot review project is nearly complete.
Based on results of that effort, we expect the full-review to be complete by December 31, 1993.,1f other testing deficiencies are identified, you will be Informed through normal reporting processes, Date When Full Comollance will be Achieved Full compliance has been achieved.
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1)S!(RC September 4,1992 I*E" '
Northern States Power Company Please contact us if you have any questions related to our response to the subject inspection report.
M NwM" Leon R Eliason fro n- -
Vice President Nuclear Generation
. - -c: _ Regional Administrator III, NRC Senior Resident-Inspector, NRC NRR Project Manager, NRC J E Silberg NRC Document Control; Desk 9
Attachment:
Response to Inspection Report j
4 RESPONSE TO INSPECTION REPORT
- 1. Introduction This is a response to Section 3.f of the NRC Inspection Report on May 20, 1992 through July 20, 1992 received by NSp on August 10, 1992.
.The following discussion pertains.to Unresolved Item (282/92011-01(DRS);
306/92011 01(DRS)) concerning potential " hot short" conditions and pressurizer power operated relief valve control during a postulated Control Room Fire.
II. Review of Pressurizer Power Operaled Relief Valve Operation in Control Eggs Fire Scenario This issue was identified and closed out as part of the inspection of July 1988.- [ Report of Inspection on July 18 22, 1988 (Reports No. 50 282/88013(DRS);No. 50 306/88013 (DRS)) dated August 19, 1988.] The major concern centered on the possibility of a " hot short" condition during a control room fire-causing a loss of pressurizer level through the inadvertent opening of a pressurizer power operated relief valve. This might occur if the block valves were not shut and a hot short were to occur such that the dam ; 4 power operated relief valve circuit opened the valve and held it or n..This sould theoretically reduce pressuriser 1cvel e
below the indicating tange, Procedural enhancements were put in place to' require the pulling of fuses to the power-operated relief valve circuit to eliminate the effect of a hot short, Additional training of plant operators is being conducted to refocus plant'avareness of this requirement.
We believe the present plant configuration and procedures to be acceptable for the following reasons:
A.
Procedural Direction
-The plant Control Room Fire Procedure (Prairie Island Operations Manual Section F5 Appendix-B) instructs-operators to close the
-pressurizer power operated relief valve block valves as-part of the immediate actions in response to a control room fire and evacuation.
This will prevent an inadvertent power operated relief valve opening due to a hot short from affecting the plant.
B.
Compensatory Monsures Further _ assurance of the control of the pressurizer power opcrated -
relief valve interface was obtained by including the pulling of fuses to the power' operated relief valve circuit as an immediate procedural action in response to a control room evacuation due to a fire. This would climinate-the effects of any hot short in the control room 5 ---.-
4 Attachmnt Page 2 of 6 power operated relief valve circuit. [ Reference Step 3.D.ii of Section F5 Appendix B.]
C.
Training Training is being beld for all plant operators as part ot Training Cycle 92 07 to enhance the familiarity and proper response of operators to this scenario in the unlikely event of a control room fire and evacuation.
D.
Compatibic with other Plant Approaches NSP understands that other plants such as Davis Besse and Nine Mile Point facilities have similar procedural actions and requirements that have been the basis of exemptions to Appendix R requirements.
E.
Previously Approved By NRC These actions were considered acceptable at the time of the inapection since:
"The licensee's (NSP] plant specific configuration for pulling the pressurizer power operated relief valve fuses was examined and found es follows:
- The fuse panels (4) are readily accessibic, e The fuses are cicarly identified in the panels,
- The fuse pancis have sufficient space to permit ready / easy access for pulling the fuses, o Fuse pullers are installed in each panel, and e The operators are trained and experienced in removing / pulling
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fuses.
Based upon the above conditions, discussions with NRR representatives, and the previous acceptance of removing / pulling fuses to achieve hot shutdown at other nuclear power plants, the licensee's control of the power operated relief valve interface was found to be satisfactory." { Report of Inspection on July 18 22, 1988 (Reports No. 50 282/88013(DRS); No. 50 306/88013 (DRS)) dated August 19, 1988.]
F.
Allowed By Generic Letter 86 10 Page 18 of Generic Lotter 8610 notes that credit for additional actions beyond the reactor trip in a control room fire must be shown;
- 1) to be feasible and 2) not susceptible to being negated by subsequent spurious actions.
1 1
A
4 At tactnet Pass 3 of 6 In the case of Prairie Island, the shutting of power operated relief valve' block valves:
- 1) Is 'easible since it requires the actuation of the block valve switches which can be performed as quickly as a reactor trip, has
-been incorporated into emergency procedures, and is the subject of operator training.
- 2) _ Will not be negated by subsequent spurious actuation since both the power operated relief valve and power operated relief valve block valves would have to sustain hot short conditions, a highly unlikely condition. Note that the spurious openin6 of the power operated reliof valve would require essentially three failure modes: 1) a hot short of the power circuit, 2) a hot short of the control power circuit, 3) sustained conditions of sufficient duration to impact plant parameters.
-C.
Boundin6 Analysis
-Existing analyses indicate that pressur:.zer level will remain in the indicating range during the time period required to perform compensatory measures (fuse pulling) in the remote chance that a flow path through the power operated relief valve were to exist. This is being confirmed by a reanalysis using all pertinent parameters for this scenario.
M.
Low Probability _of Occurrence The probability of this event occurring at all is extremely low. A review of the likelihood of a control room fire and spurious operation' of the power operated relief valve supports this-conclusion.
Control Room Fire u
The probability of-control room fire requiring evacuation is _quite small. This is,because:
- 1. EThe Control Room is continuously manned by qualified operators. In the event of a fire in the Control Room, it would most likely be rapidly detected and extinguished. The Control Room is equipped with an ionization-type smoke detection system, portable fire
.extirmtisher's,-and fire hose. station standpipe-system.
This system of fire detection and extinguishment was_ reviewed and approved by the NRC as an exemption to the requirenert of a fixed fire--suppression system'in the Control Room (Reference NRC letter Jto NSP dated 2/2/83).
2.
The'NRC Fire Protection Safety Evaluation Report (SER), dated
. September-6,L1979_ evaluated specific plant areas _for combustibles, consequences of lack of fire' suppression,.and the adequacy of the
> fire protection. systems. This report states:
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Attachment Page 4 of 6 any unsuppressed fire in a control room cabinet has the potential for damaging significant anounts of safety-related equipment including systems required for safe shutdown.
However, the cabinets and consoles are compartmentalized to
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provide protection of redundant division.
The SER further concluded that the " separation provided between n--
redundant circu'.ts is adequate."
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3.
The Prairie Island Nuclear Generating Plant Fire Hazards Analysis (HUO indicates that tha fire loading in the Control Room is light; i.e. 8601 Btu /sq.ft. cables, and 110 Btu /sq.ft. misc.
The Y
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FHA, Table 6-2, further indicates the Control Room fire severity 19 " trivial."
i Sourious Valvg_ Operation For the power operated-relief valve spurious valve operation to s
occur, the following series of events and conditions must exist:
1.
Fire-in Control Room occurs which r3 quires evacuation, 2.
The operators-(which are very familiar with contrcl room evacuation procedures) fail to shut the power operated relief valve block valves, 3.
The panel which encloses the circuitry for the pressurizer power operated relief valve is burned in such a way as to cause a " hot short,"
4,- A sustained " hot short" necurs to continually hold the power n
operated elief valve og.mu.
5.
This scenario woulo ' ave to cccur within the short per od of time prior to pulling of power fuses to the affected power.perated relief valve control circuitry.
A brief review of this scenario by NSP Probabilistic Risk nssessment staff-and consultants indicated that the likelihood of this sequence of events is extremely small. (Probably less than 10-6 7
occurrences / reactor year).
In summary, the likelihood cf a control room fire as well as the existence of. conditions which would cause en open power operated relief valve-is very remote.
III.
Other Hot Short Scenarios The following discussion pertains to the other hot s) ort scenarios that were identified by the plant design basis reconstittt in project and Jaddressed in the inspection report.
l
4 At tacheent Page 5 of 6 A.
Loss of 4KV Breaker Control A number of circuits for control power to 4160 Volt load breakers were identified as susceptible to failure during a control room fire. The issue was reviewed by the plant Operations Committee and a safety evaluation was prepared to ensure interim suitability of the plant procedures and compensatory measures concerning the fire impact on 4KV circuit breaker control. Long term resolution of this issue includes a review of the circuit design and possible modification to provide additional fuse protection for these
- circuits, b
B.
12 Diesel-driven Cooling Water Pump Circuit A control room fire hot short failure mode was identified that could lead to failure of the 12 Diccel-driven Cooling Water Pump to start when required. The issue was reviewed by the plant Operations Committee and a safety evaluation was prepared to ensure int 3 rim suitability of the plant procedures and compensatory measures 3
>O concerning the fire impact on pump control. The long term corrective I '
action was to require pulling of the appropriate fuses to prevent R 96 start sequence failure of the pump control circuit.
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C.
Reactor llead Vent Valves A control room fire hot short failure mode was identified tnat could lead to inadvertent opening of the reactor coolant head vent solenoid valves. The issue was reviewed by the plant Operations Committee and a safety evaluation was prepared to ensure interim suita'oility of the plant procedures and compensatory measures concerning the fire impact on valve operation. The long term corrective action was to require pulling of the appropriate fuses to prevent inadvertent operation of the valve control circuits.
D.
Other Issues Noted by Inspectors The concerns noted by the inspection team such as drawing errors and shutdown list availability are being addressed by the ongoing design basis reconstitution project which includes a review of 10 CFR Part 50, Appendix R documentation, procedures, plant configuration and controls. The circuit drawing discrepancies noted in the inspection report are being addressed by a planned "as-built walkdown" of several safety related electrical circuits in conjunction with the upcoming plant outage.
IV. Ongelne Activities The existing c:nalysis of the various possible effects of a contcol room fire ou plant parameters is under review. This review will include the 1
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potential ramifications 1of an open pressuricer power operated relief valve condition and ensure thit sufficient _ time is available to complete all actions' required during a control room fire scenario. Preliminary indications are that the analysis will indicate that sufficient time is available4 to complete the necessary actions.
A review of the Prairie Island 10 CFR Part 50, Appendix R analyses and documentation is in progress as part of the Design Basis Reconstitut1 n program. This effort will provide added assurance tC the adequacy of all fire protection documentation and designs at the plant.
- An _ Exemption Request to the requirements of 10 CFR 50, Appendix R will be prepared'for. submission to the NRC based on-the above reviews. Long term-corrective actions requirco by the 10 CFR 50, Appendix R program,'the idesign: basis reconstitution effort, and this recent inspection will be incorporated as appropriate-for the exemption request.
ll, Conclusion Based on the above review, no further actions are considered necessary
- beyond thefongoing actions of the analysis preparation, Design Basis review, and exemption request _Given the procedural and training enhancements, compatibility with industry and previous NRC decisions, analysia, _andLlow probability of occurrence, NSPfconcludes that the plant
- is_ adequately! protected from the.affects of a control room fire.
The-careful self assessment of 10 CFR 50, Appendix R compliance in
,rogress'at Prairie Island isl designed to ensure that the plant documentation, design, and procedures aro adequate for all anticipated
- fire scenarios ~. This will-provide additional confidence in the controls
- and configuration of fire protection programs and equipment at Pratrie Island.
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