ML18016A397
| ML18016A397 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 04/22/1998 |
| From: | Scott Flanders NRC (Affiliation Not Assigned) |
| To: | Robinson W CAROLINA POWER & LIGHT CO. |
| References | |
| IEB-96-002, IEB-96-2, TAC-M95593, NUDOCS 9804270390 | |
| Download: ML18016A397 (31) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 22, 1998 Mr. W. R. Robinson, Vice President Shearon Harris Nuclear Power Plant Carolina Power & Light Company Post Office Box 165, Mail Code: Zone 1
New Hill, North Carolina 27562-0165
SUBJECT:
COMPLETION OF LICENSING ACTION FOR NRC BULLETIN96-02,
, "MOVEMENT,OF HEAVYLOADS OVER,SPENT FUEL, OVER.FUEL IN.THE.
REACTOR CORE, OR OVER SAFETY-RELATED EQUIPMENT," DATED APRIL 11, 1996, for SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 (TAC NO. M95593)
Dear Mr. Robinson:
On April 11, 1996, the U.S. Nuclear Regulatory Commission (NRC) issued NRC Bulletin (NRCB) 96-02, "Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or Over Safety-Related Equipment," to all holders of operating licenses.
The NRC issued NRCB 96-02 for three principal reasons:
1.
Alert addressees to the importance of complying with existing regulatory guidelines associated with the control and handling of heavy loads at nuclear power plants, 2.
Request that all addressees review their plans and capabilities for handling heavy loads in accordance with existing regulatory guidelines and within their licensing basis as previously analyzed in the final safety analysis report, and 3.
Require addressees to report to the NRC whether and to what extent they have complied with the actions requested in this bulletin, Also the bulletin requested that Carolina Power 8 Light (CP8 L) determine whether current activities were within the licensing basis and to submit a license amendment request as necessary.
In addition, by a Request for Additional Information, the NRC staff also requested that CP8L prepare and submit an evaluation of the plant's heavy-load activities in the moving of dry storage casks.
This evaluation was to determine ifa tipping-over hazard exists while dry storage casks are being moved by plant cranes.
In response to NRCB 96-02, you provided letters dated May 13, 1996, February 11, 1997, and March 14, 1997, for Shearon Harris Nuclear Plant. These submittals provided both the information requested and the responses required by NRCB 96-02.
NRC staff review of the re, ponses to NRCB 96-02 finds that, overall, the responses are acceptable; therefore, TAC No.
M95593 is closed.
The staffs summary review of the licensee responses to the bulletin is enclosed.
.~'Q5
'7804270390 980422 PDR ADQCK 05000400 8
PDR JrQ Po I
Mr. W. Robinson The NRC willcontinue to review the issue of heavy loads through an ongoing Task Action Plan for heavy loads. Any additional information required for the completion of the Task Action Plan willbe obtained on a plant-specific basis.
Ifyou have any questions regarding this matter, please contact Scott Flanders at (301) 415-1172.
Sincerely, (Original Signed By)
Docket No. 50-400
Enclosure:
As stated cc: See next page
~SU rDocket:FilM PUBLIC PD II-1 Reading File J. Zwolinski L. Plisco, Region II M. Shymlock OGC P. Ray ACRS B. Thomas Scott Flanders, Project Manager Project Directorate II-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation See next page cc: Harris Service List FILENAME-G:>HA RIS(HAR95593.clo OFFICE PM:P II NAME SFla rs DATE
//7498 COPY Yes/No LA:PDII-1 EDunningto
/%98 es No AD:PDII-1 PKuo f/%98 Yes/No
Mr. W. R. Robinson Carolina Power 8 Light Company Shearon Harris Nuclear Power Plant Unit 1 CC:
Mr. William D. Johnson Vice President and Senior Counsel Carolina Power 8 Light Company Post Office Box 155'aleigh, North Carolina 27602 Resident Inspector/Harris NPS c/o U.S. Nuclear Regulatory Commission 5421 Shearon Harris Road New Hill, North Carolina 27562-9998 Ms. Karen E. Long Assistant Attorney General State of North Carolina Post Office Box 629 Raleigh, North Carolina 27602 Public Service Commission State of South Carolina Post Office Drawer Columbia, South Carolina 29211 Regional Administrator, Region II U.S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303 Mr. Mel Fry, Acting Director Division ofRadiation Protection N.C. Department of Environment, 3825 Barrett Dr.
Raleigh, North Carolina 27609-7721 Ms. D. B. Alexander Manager Performance Evaluation and Regulatory Affairs Carolina Power 8 Light Company 412 S. Wilmington Street Raleigh, North Carolina 27601 Mr. Bo Clark Plant General Manager - Harris Plant Carolina Power 8 Light Company Shearon Harris Nuclear Power Plant P.O. Box 165 New Hill, North Carolina 27562-0165 Mr. J. W. Donahue Director of Site Operations Carolina Power 8 Light Company Shearon Harris Nuclear Power Plant Post Office Box 165, MC: Zone 1 New Hill, North Carolina 27562-0165 Mt. Robert P. Gruber Executive Director Public Staff NCUC Post. Office Box 29520- -.
Raleigh, North Carolina 27626 Chairman of the North Carolina Utilities Commission Post Office Box 29510 Raleigh, North Carolina 27626-0510 Mr. Milton Shymlock U.S Nuclear Regulatory Comm.
Atlanta Federal Center 61 Forsyth Street, SW, Suite 23185 Atlanta, Georgia 30303 Mr. Stewart Adcock, Chairman Board of County Commissioners of Wake County P. O. Box 550 Raleigh, North Carolina 27602 Health and Natural Resources Margaret Bryant Pollard, Chairman Board of County Commissioners of Chatham County P. O. Box 87 Pittsboro, North Carolina 27312 Mr. Chris A. VanDenburgh, Manager
, Regulatory Affairs Carolina Power & Light Company Shearon Harris Nuclear Power Plant P.O. Box 165, Mail Zone 1
New Hill, NC 27562-0165 Mr. Johnny H. Eads, Supervisor Licensing/Regulatory Programs Carolina Power 8 Light Company Shearon Harris Nuclear Power Plant P. O. Box 165, Mail Zone 1
New Hill, NC 27562-0165
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The following summarizes the results of the staffs review of licensees'esponses to NRC Bulletin (NRCB) 96-02, "Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or. Over Safety-Related Equipment," dated April 11, 1996, and its associated Requests for-Additional Information (RAI). The bulletin reminded licensees of their responsibilities for ensuring that heavy load-handling operations are performed safely.
It also requested that licensees review their plans and capabilities for handling heavy loads, and ensure that their load-handling operations are in accordance with existing regulatory guidelines and the plant's licensing basis.
Also requested was that licensees identify and present schedules for licensing actions needed to support'implementation of their heavy load-handling operations involving spent fuel dry storage casks.
The licensees also were to provide schedules for moving dry storage casks.
The RAI requested that selected licensees evaluate the hazards associated with an in-plant tip-over of spent fuel dry storage casks that could dislodge the cask lid and spent fuel elements.
This summary closes the staff's review of licensee responses to both the bulletin and the associated RAI. Future issues regarding the handling of heavy loads willbe addressed generically under the Heavy Loads and Crane Issues Task Action Plan (TAP) and on a plant-specific basis as needed.
Plant-specific reviews needed in the future may require the staff to obtain additional information from individual licensees.
Bac rou d NRCB 96-02 was issued as an urgent generic communication that requested licensees'esponses to the following:
(1)
For licensees planning to carry out activities involving the handling of heavy loads over spent fuel, fuel in the reactor core, or safety-related equipment within the next 2 years from the date of the bulletin, provide the following: A report within 30 days of the date of the bulletin that addresses the licensee's review of its plans and capabilities to handle heavy loads while the reactor is at power (in all modes other than cold shutdown, refueling, and defueled) in accordance with existing regulatory guidelines.
State whether the activities are within the licensing basis and, ifnecessary, submit a schedule for requesting a license amendment.
Additionally, indicate whether changes to Technical Specifications (TSs) are required.
(2)
For licensees planning to perform activities involving the handling of heavy loads over spent fuel, over fuel in the reactor ccrc, or over safety-related equipment while the reactor
'I Enclosure Report (FSAR), submit a license amendment request 6-9 months in advance of the planned movement of the loads to give the staff sufficient time to perform an appropriate review.
(3)
For licensees planning to move dry storage casks over spent fuel, over fuel in the reactor core, or over safety-related equipment while the reactor is at power (in all modes other than cold shutdown, refueling, and defueled) include, in item 2 above, a statement of the capability of performing the actions necessary for a safe plant shutdown in the presence of a radiological source term that may result from a breach of the dry storage cask, damage to the fuel, or damage to safety-related equipment due to a load drop inside the facility.
'E (4)
For licensees planning to perform activities involving the handling of heavy loads over spent fuel, over fuel in the reactor core, or over safety-related equipment while the reactor is at power (in all modes other than cold shutdown, refueling, and defueled), determine whether changes to the TSs will be required to allow the handling of heavy loads (e.g., the dry storage canister shield plug) over fuel assemblies in the spent fuel pool and submit the appropriate information 6-9 months in advance of the planned movement of the loads for NRC review and approval.
Dsc sso The levels of'detail in the licensees'esponses to NRCB 96-02 varied significantly. Although some licensees presented detailed information about their heavy load-handling operations, some licensees (Catawba, Crystal River, Farley, Indian Point 2, Salem, St. Lucie, Summer, Dresden, Fitzpatrick, Hope Creek, LaSalle, Quad Cities, and WNP-2), either omitted information pertinent to the staffs review in their submittal or referenced previous submittals associated with NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants."
However, all of the licensees responded to the bulletin.
In response to the bulletin, all the licensees reviewed their plans and capabilities to handle heavy loads and indicated that their plans and capabilities are adequate.
Some discussions about licensees'lans and capabilities to move heavy loads addressed the plant mode of operation (at power or during shutdowns), the type of crane used (non-single-failure-proof, single-failure-proof, or upgraded cranes), and the methods and procedures for implenienting the guidelines in NUREG-0612, Phase I. Allthe licensees indicated that their load-handling operations are in accordance with the guidelines in NUREG-0612, Phase I.
The bulletin requested that licensees determine whether their load-handling operations are within the licensing basis of the plant. Some licensees stated that their operations are within the licensing basis; other licensees committed to evaluate their licensing basis: Some licensees identified issues to be addressed with the NRC through licensing actions (amendment requests or 10 CFR 50.59 evaluations), and projected schedules for submitting the actions for NRC review. Following the responses to the bulletin, a few licensing actions have been reviewed and approved by the NRC concerning the bulletin. The issues involve proposed changes to TSs, scope changes to accident analyses, changes in loads and load paths, and updates to UFSAR requirements.
The bulletin also asked licensees to determine iftheir movement of heavy loads involves potential load drop accidents that were not evaluated previously in the FSAR and, ifneeded, submit a license amendment request.
Most licensees stated that they move only analyzed loads.
Some licensees indicated that they performed load drop or consequence analyses or both though the guidance in Generic Letter (GL) 85-11 canceled the. need to perform any analyses.
Some licensees committed to evaluate the heavy loads identified previously when they responded to NUREG-0612..Despite the analyses performed, all the licensees stated that they satisfy the recommended guidelines in Section 5.1.1 of NUREG-0612.
Licensees moving heavy loads at power and using load drops and consequence analyses indicated that they have. adequate capabilities to safely shutdown the plant ifa heavy load drop-occurs causing a release of radiation or damage to safety-related equipment.
The bulletin also requested that licensees identify plans and schedules for moving spent fuel dry storage casks.
Some licensees stated that'they planned to move casks in the near future; other licensees indicated that they had not yet considered onsite dry cask storage.
Based on requests in the bulletin, the staff reviewed the licensee responses to identify:
(1) plant mode during the handling of heavy loads (at'power or during plant shutdowns);
(2) type of crane used to liftheavy loads; (3) evaluation of the licensing basis for handling heavy loads, including planned licensing actions associated with heavy loads (i.e., license amendment requests); (4) plans and schedules for moving heavy loads (particularly spent fuel dry storage and transportation casks); and (5) the type of analysis performed (load drop analysis or consequence analysis or both). Although the bulletin did not specifically request this information, the staff believes that this type of information covers the areas of concern about the licensees'eavy load-handling operations.
On the basis of its review, the staff noted the following points.
odeD a
dlin Review of the responses to the bulletin revealed that approximately 38 percent of the plants (21 PWRs and 20 BWRs) plan to move heavy loads at power. Some of these plants indicated that they move analyzed heavy loads at power and unanalyzed heavy loads during plant shutdowns.
These plants also indicated that heavy load movements over safety-related equipment are minimized to the extent practicable, and their procedures do not allow movements of heavy loads over fuel or over the reactor core in accordance with NUREG-0612. Some PWR licensees (i.e., Callaway, Shearon Harris, and Calvert Cliffs) indicated that their heavy load movements involve casks moved within a separate fuel building. As indicated by the licensees, the movement of casks in PWRs that have a separate fuel building involves little or no cask travel over systems needed for safe shutdown functions. As a result, a dropped cask would not cause significant damage to safe shutdown equipment and, therefore, would have negligible effect on the licensees'bility to shut down the plant safely.
Approximately 39 percent of the plants (28 PWRs and 15 BWRs) indicated that they move heavy loa'ds at plant shutdowns, and about 23 percent of the plants (23 PWRs and 2 BWRs) did not clearly indicate the plant status when heavy loads are moved. A few of
these licensees (e.g., Oyster Creek) that plan to move heavy loads during plant shutdowns also indicated that they plan to perform dry runs at power, before initiallyloading the cask.
The staff finds that although some licensees have committed to move only analyzed loads at power, they may not adequately consider the adverse safety consequences of a load drop during the movement of heavy loads.
Some licensees'nalyses consider methods that may be used to preclude a load drop (e.g., enhancements to the load handling system, including upgrades to brakes, instrumentation, and controls, and the use of energy-absorbing structures throughout the load path).
However, they may not consider the adequacy of their capabilities needed to mitigate or manage the adverse
- consequences of a load drop. Some examples of such capabilities are the abilities to shut down the plant safely, continue normal operation, maintain personnel access to various areas in the plant, and mitigate potential accidents that could expose individuals to releases.
The staff is also concerned that some licensees may not adequately address the potential consequences of a load drop during practice runs of cask movements while the reactor is at power. A drop of an empty cask during practice movements could result in similar adverse consequences to the operation of the plant as does the actual movement of a fully loaded spent fuel cask, Therefore, it is the staffs view that activities involving actual heavy load movements or practice runs of moving spent fuel dry storage casks are to be evaluated by the licensee for potential accidents and consequences.
In addition, the staff is concerned with BWR licensees that move heavy loads while the reactor is at power because, in general, the safety-related systems required for safe shutdowns are susceptible to damage from a dropped heavy load. These licensees should exhaust all options of establishing safe load paths to minimize the risk of affecting safe shutdown equipment in the event a heavy load is dropped.
(2)
T e
of r
In the responses to the bulletin, approximately 27 percent of the plants (6 PWRs and 23 BWRs) indicated that they use single-failure-proof cranes to liftheavy loads; 14 percent of the plants (12 PWRs and 3 BWRs) indicated that they have upgraded the reliability of their load-handling system in accordance with NUREG-0612, Section 5.1.6 (see explanation below); and about 8 percent of the plants (5 PWRs and 4 BWRs) indicated that their crane is non-single-failure-proof.
However, almost half the plants (49 PWRs and 7 BWRs) did not clearly indicate the type of crane they use.
NUREG-0612, Section 5.1.6, "Single Failure Handling System," provides the alternative of upgrading an existing crane in lieu of complying with certain recommendations of NUREG-0554, "Single Failure Proof Cranes for Nuclear Power Plants," to achieve improved reliability in load-handling systems.
Accordingly, several licensees have upgraded their overhead load-handling crane to single-failure-proof status, or they have improved reliability by increasing the factors of safety or by providing redundancy in certain active components of the cranes.
A few licensees (i.e., Oyster Creek, Dresden, Yankee Rowe)
4 ~
have indicated that they are considering upgrading their cranes or installing new cranes to achieve single-failure-proof capability.
Licensee information regarding the types of overhead cranes used at the plants indicates that many plants have either single-failure-proof cranes in accordance with NUREG-0554, "Single-Failure-Proof Cranes for Nuclear Power Plants," or cranes upgraded in accordance with guidelines in NUREG-0612 (Section 5.1.6, and Appendix C, "Modification of Existing Cranes)."
Although several plants were not clear about the type of crane they possess, none of the plants indicated that they have cranes and liftingsystems that were inadequately designed, installed, and tested.
- ~
~
/
The staff concludes that many licensees previously performed adequate evaluations of their crane design for lifting heavy loads and the evaluations were accepted by the staff.
However, the staff is concerned that some facilities could have weaknesses in their load-handling operations.
These weaknesses may include insufficient training of personnel involved in the lifting and rigging procedures, procedures lacking in requirements. for evaluating loads and ensuring that the design limitations of the hoisting system are not
- exceeded, insufficient inspection and preventive maintenance of cranes and lifting devices, and inadequate review of loading capacities.
The staffs view is that the potential exists for any of these weaknesses to result in a single failure involving heavy loads being dropped and causing adverse consequences.
As a result, future staff reviews willbe focused on licensees'valuations of their cranes and liftingdevices, and related methods and procedures used for complying with the requirements of NUREG-0612.
o ice si 8
i I'
v Load Review of the responses to the bulletin indicated that all of the licensees believe that their heavy toad-handling operations are in accordance with the licensing basis of the facility.
Approximately 24 percent of the plants (10 BWRs and 16 PWRs) did not address the licensing basis in their responses.
The staff is concerned that some plants that believe
'heir load-handling operation is within the plant's licensing basis may, in fact, be outside the licensing basis.
For example, the staff's reviews of Oyster Creek's (OC's) load-handling operations determined that OC would have operated beyond its licensing basis.
This is because OC was planning to move loads that exceeded the size of the loads previously evaluated in the FSAR. Approximately 10 percent of the licensees indicated that they willreview and modify their licensing basis as needed.
As indicated in the submittals, licensees'eviews of the licensing basis resulted in one or more of the following:
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identification and analysis of new heavy loads beyond the loads previously addressed in the licensing basis,
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commitments to only move heavy loads that were previously analyzed,
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determinations that heavy load-handling operations deviated from previous commitments and the licensing bases, and
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determinations that change the TSs are needed.
Licensees'eviews of their plans and capabilities to handle and control heavy loads have resulted in some licensees undertaking licensing actions to implement their load-handling operations.
The following are examples of planned licensing actions noted in the responses to NRCB 96-02:
~ce eee a
e s
= Brunswick:.
License amendment request to make the FSAR consistent with actual
-'lant operations (completed).
Fitzpatrick:
Changes to the TSs to allow the movement of spent fuel dry storage casks at power (schedule TBD).
Nine Mile Point: Design change involving reracking of the spent fuel pool. (Schedule TBD).
North Anna:
Various license amendments regarding heavy load-handling issues (schedule TBD).
Oyster Creek:
TS changes to remove the weight restriction for liftingthe dry storage canister (DSC) shield plugs over fuel in the DSC. (completed).
Watts Bar:
Design change for reracking of the spent fuel pool (currently under review).
The staffs review of the information submitted indicates that some licensees'oad-handling operations may have been implemented inconsistently with the licensing basis of the facility. Some plants either have inadvertently deviated from their load-handling procedures, implemented procedures that are inconsistent with the licensing basis, or misinterpreted the design features of their load-handling system.
The staff also believes that since the issuance of NUREG-0612, many changes have evolved in licensees'lans to handle heavy loads. As a result, several licensees have identified changes in their load-handling operations that were not previously addressed in their licensing basis.
Therefore, on an "as needed" basis, the staff willcontinue to perform audits and inspections in order to evaluate licensees'ovement of heavy loads.
(4)
I s or Movin D
or Approximately 17 percent ofthe plants (10 PWRs and 9 BWRs) indicated that they plan to store spent fuel dry storage casks.
Most of these plants plan to move casks within 2 years from the date of the bulletin. The remainder of the licensees either did not address the issue or have not yet begun planning for the storage of spent fuel.
(5)
L D
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iP Approximately 33 percent of the plants indicated that they have performed load drop and consequence analyses in support of their plans to move heavy loads. The remaining plants did not show that any analysis exists.
In the future, the staff willreview the load drop and consequence analyses on an as-needed plant-specific basis.
The staff has found that several licensees have done load drop and consequence analyses though Generic Letter 85-11 canceled Phase II of NUREG-0612, and dismissed the need for licensees to perform these analyses.
The results of the analyses have led some licensees to modify their load-handling operations, including upgrading the crane and associated components of the liftingsystem, and modifying the load paths:
~Gus o The staff finds that NRC Bulletin 96-02 achieved its objective of getting licensees to evaluate their load-handling activities to ensure that they are performed safely and in the best interest of protecting the health and safety of the public. The bulletin was very effective in getting licensees to review their plans and capabilities, licensing bases, and regulatory guidelines for carrying out activities involving the movement of heavy loads.
Although the licensee responses to the bulletin contained various levels of detail regarding load-handling operations at their plants, sufficient information was available to enable the staff to reach the conclusions noted below.
Although several licensees have increased the reliability of their load-handling systems, the staff willcontinue to review load-handling operations, on an as-needed basis, to ensure that licensees adequately address their ability to preclude load drop accidents.
As determined through earlier NRC reviews, licensees have reliable lifting systems as required by NUREG-0612.
However, licensees need to continue to address other activities surrounding the crane operation that could help to minimize weaknesses in their load-handling operations that may
, contribute to load drop accidents.
Such weaknesses could include insufficient training of personnel involved in applying the lifting and rigging procedures, procedures lacking in requirements for evaluating loads and for ensuring that the design limitations of the load-lifting system are not exceeded, insufficient inspection and preventive maintenance of cranes and liffingdevices, and inadequate review of loading capacities.
Also, the staff finds that because some licensees plan to move heavy loads at power, they may need to assess their capabilities to both mitigate and manage the adverse consequences of a heavy load drop.
Licensees should consider, among other things, possible plant shutdowns during the movement of heavy loads, limiting personnel exposure from required entry into contaminated plant areas following an accident, and recovering from the adverse conditions caused by an accident.
Accordingly, the staff is particularly interested in future evaluations of load drops and consequences associated with the load-handling operations of the licensees.
The staff also finds that several licensees have determined, after reviewing their licensing basis, that their load-handling operations may be inconsistent with their licensing basis.
Consequently, several licensees have undertaken actions to correct or resolve this condition, including reviewing the FSAR, TS requirements, and procedures governing the conduct of operations involving the movement of heavy loads. The staff will pursue enforcement actions for matters involving a noncompliance with regulatory requirements as appropriate.
On the basis of the preceding discussion, the staff willcontinue to review issues regarding the handling of heavy loads on a plant-specific basis as needed.
Generic issues regarding this subject will be addressed through an ongoing Task Action Plan (TAP) for Heavy Loads. Any additional information required for the completion of the TAP willbe obtained on a plant-specific basis.
Principal Contributor: Brian E. Thomas
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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003696694:1
Subject:
P,PR 04 350 Page 1
Distri97.txt Shearon Harris, Unit 0, Control Room Emergency Filtration System Technical Specificat ion Violation, Licensee Event Report Body:
ADAMS DISTRIBUTIONNOTIFICATION.
Electronic Recipients can RlGHT CLICKand OPEN the firstAttachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML003696694.
IE22 - 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.
Docket: 05000400 Page 2
Carolina Power S light Company Harris Nuclear Plant PO Box 165 New HillNC 27562 MAR ~S 2000 U.S. Nuclear Regulatory Commission ATTN:NRC Document Control Desk Washington, DC 20555 Serial: HNP-00-061 10CFR50.73 SHEARON HARRIS NUCLEARPOWER PLANTUNIT 1 DOCKET NO. 50-400 LICENSE NO. NPF-63 LICENSEE EVENT REPORT 2000-001-00 Sir or Madam:
In accordance with 10CFR50.73, the enclosed Licensee Event Report is submitted. This report describes a Technical Specification violation due to an inoperable Control Room Emergency Filtration System.
Si
- rely, J
uncan II General Manager Harris Plant MSElmse Enclosure c:
Mr. J. B. Brady (HNP Senior NRC Resident)
Mr. R. J. Laufer (NRC-NRR Project Manager)
Mr. L. A. Reyes (NRC Regional Administrator, Region II) 5413 Shearon Harris Road New HillNC '
NRC FORM 366 U.S. NUCLEAR REGULATOR COMMISSION (6-1998)
LtCENSEE EVENT REPORT (LER)
(See reverse for required number of digits/characters for each block)
APPROVED OMB NO. 3150%104 EXPIRES 06/30/2001 Estimated burden per response to comply with dis mandatory kIIOImatian collection request 50 hrL Reported lessons learned are ncorporated into the ircensng process and fed back to indusby.~ comments reganfrng burden estimate to the Informa5on and Records Management Branch (T4 F33), U.S. Nudear Regu)story Commiss)on, Washington, DC 205554NI, and to the Paperwork Reduction Pro(act (31500104), 05ce ol Management and Budget, Wssington. DC 20503.
If an intormation coaction does not display a currendy valid OMB control number, the NRC may not conduct or
- sponsor, and a person is not required to respond to, the information colecbon.
FACIUTYNAMEl1)
Harris Nuclear Plant, Unit 1 DOCKET NUMBER l2) 05000400 PAGE I3) 1 OF 2
TITLE te)
Control Room Emergency Filtration System Technical Specification violation.
MONTH DAY YEAR SEOUENTIAL REVISION NUMBER NUMBER MONTH OAY YEAR FACILITYNAME DOCKET NUMBER 02 22 2000 2000
- 001' 00 03 23 2000 FACIUTYNAME DOCKET NUMBER 05000 OPERATING MODE (9)
POWER LEVEL (10) 100 grrygprL
< Lr4",:+~at),:.
4W$@~
~'. t.""a.
20.2201(b) 20.2203(a)(1) 20.2203(a)(2) ti) 20.2203(a)(2)(iil 20.2203(a)(2) (iii) 20.2203(a)(2) (iv) 20.2203(a)(2)(v) 20.2203(a)(3) II) 20.2203(a) l3)(iil 20.2203(a)(4) 50.36(c) l1) 50.36(c)(2) x 50.73(a)(2)(I) 50.73(a)(2)(ii) 50.73(a)(2)(iii) 50.73(a) (2)(ivl 50.73(a) (2)(v) 50.73(a)(2)(vii) 50.73(a) (2)(viii) 50.73(a) l2)tx) 73.71 OTHER Specify in Abstract below or in NRC Form 366A Mark Ellington, Senior Analyst - Licensing TELEPHONE NUMBER tlnclude Area Code)
(919) 362-2057 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE To EPIX CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE To EPIX YES (lf yes, complete EXPECTED SUBMISSION DATE).
x NO EXPECTED MONTH DAY YEAR On February 22, 2000, with the Harris Nuclear Plant (HNP) at 100% reactor power, a duct access panel for the Control Room Emergency Filtration System (CREFS) ventilation boundary was removed to facilitate surveillanc'e testing of the charcoal in the filtration unit. This access panel is located at the common suction line for the two redundant CREFS units, R-2A and R-2B. During the time (approximately five minutes) that the access panel was removed and the test panel installed, the CREFS could not achieve and maintain a positive pressure as required by Technical Specifications (TS) surveillance requirement 4.7.6.d.3. This condition caused both of the CREFS units to become inoperable. Harris Nuclear Plant (HNP) TS do not provide an action when both CREFS are inoperable.
Additionally, there is no action, in TS 3.7.6, for an inoperable'CREFS ventilation boundary. Thus, opening the CREFS duct work during performance of the surveil)ance test required entry into TS 3.0.3.
Cause of this event: Inadequate review of plant procedures which affect control room ventilation boundaries.
Corrective actions include: (1) Reviewing p)ant procedures to ensure that passive barriers which may be opened do not violate CREFS pressure boundary requirements.
(2) Install signs on applicable ventilation duct access panels instructing plant personnel to contact the main control room prioi to removal of the associated duct access panel.
NRC FORM 366 I6 1996)
O.
NRC FORM 366A (6-96)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION FACIUTYNAMEIl)
Harris Nuclear Plant, Unit 1 DOCKET 05000400 LER NUMBER IB)
YEAR SEOUENT)AL REVISION NUMBER NUMBER 2000
001
00 PAGE )3) 2 OF 2
TEXTIffmare spsce is eeafrerf, use erfdfoonsl copIes ofNRC Form 366AI I1 7)
I.
DESCRIPTION OF EVENT On February 22, 2000, with the Harris Nuclear Plant (HNP) at 100% reactor power, a duct access panel for the Control Room Emergency Filtration System (CREFS) ventilation boundary was removed to facilitate surveillance testing of the charcoal in the filtration unit. This access panel is located at the common suction line for the two redundant CREFS units, R-2A and R-2B. During the time (approximately five minutes) that the access panel was removed and the test panel installed, the CREFS could not achieve and maintain a positive pressure as required by Technical Specifications (TS) surveillance requirement 4.7.6.d.3. This condition caused both of the CREFS units to become inoperable. Harris Nuclear Plant (HNP) TS do not provide an action when both CREFS are inoperable. Additionally, there is no action, in TS 3.7.6, for an inoperable CREFS ventilation boundary. Thus, opening the CREFS duct work during performance of the surveillance test required entry into TS 3.0.3.
The CREFS consists of two 100 percent capacity redundant fan and filter subsystems.
Normally, the CREFS is in a standby alignment. During an accident, the normal outside air intake for the CREFS isolates and both emergency recirculation fans automatically start. Following verification of isolation of control room ventilation, an operator places one of the two emergency filtration units in standby. Next, the operator selects and opens one emergency outside air intake to pressurize the control room to 1/8 inwg. at less than or equal to 315 cfm flow.
With the applicable access panel removed, operators would have been unable to pressurize the control room as required.
II.
CAUSE OF EVENT Inadequate review of plant procedures which affect control room ventilation boundaries.
III. SAFETY SIGNIFICANCE There were no actual safety consequences as a result of this event. The individual performing the test could have replaced the access panel in a timely manner should the need arise. This report is being submitted pursuant to the criteria of 10CFR50.73(a)(2)(i) for Technical Specification Prohibited Operation or Condition.
IV. CORRECTIVE ACTIONS (1) Reviewing plant procedures to ensure that passive barriers which may be opened do not violate CREFS pressure boundary requirements.
(2) Install signs on applicable ventilation duct access panels instructing plant personnel to contact the main control room prior to removal of the associated duct access panel.
SIMILAR EVENTS HNP submitted LER 1999-08-00 to document that for during four refueling outages, main control room doors were being blocked open to facilitate testing. The corrective actions for that investigation did not consider breaching of ventilation boundaries other than doors. As a result, this issue was not identified during that investigation.
NRC FORM 366 (6.96)
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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003676422
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Distri28.txt SHEARON HARRIS NUCLEAR POWER PLANT UNIT 0 LICENSEE EVENT REPORT 1999-009-00 Body:
PDR ADOCK05000400 S Docket: 05000400, Notes: Application for permit renewal filed.
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Carolina Power & light Company Harris Nuclear Plant PO Box 165 New HillNC 27562 Jw 13 2000 U.S. Nuclear Regulatory Commission ATTN:NRC Document Control Desk Washington, DC 20555 Serial: HNP-00-004 10CFR50.73 SHEARON HARRIS NUCLEARPOWER PLANTUNIT 1 DOCKET NO. 50-400 LICENSE NO. NPF-63 LICENSEE EVENT REPORT 1999-009-00 Sir or Madam:
In accordance with 10CFR50.73, the enclosed Licensee Event Report is submitted. This report describes a condition which resulted in a manual reactor trip and auxiliary feedwater system actuation.
Sincerely, B. H. Clark General Manager Harris Plant MSE/mse Enclosure c:
Mr. J. B. Brady (HNP Senior NRC Resident)
Mr. R. J. Laufer (NRC-NRR Project Manager)
Mr. L. A. Reyes (NRC Regional Administrator, Region II) 5di 3 Sh r
HarreisaRoooad New Hill NC rs iw" '7Q/p g L<'~~9 cD-
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (6-1999)
LICENSEE EVENT REPORT (LER)
(See veverse ror required number of digits/characters for each block)
APPROVED MB NO. 3150-0104 EXPIRES 06/30/2001 Estimated burden per response to compty with INs mandatory Infonnatkxt request 50 hrs.
Reported lessons learned are incorporated Into the licensing process and fed back to industry. Forward comments regarding burden estimate to the Information and Records Management Branch (TQ F33), U.S. Nudear Regulatory Commission, Washington, DC 205554001, and to the Paperwork Reduction Proied (31500104), Ofrice of Management and Budget, Washington, DC 20503.
II an information collection does not display a currently valid OMB control number, the NRC may not conduct or
- sponsor, and a person is not required to respond to, the InforrnaUon collection, FACILITYNAMEl1l TITLE tel Harris Nuclear Plant, Unit 1 DOCKET NUMBER l2l 05000400 PAGE I3) 1 OF 3
Reactor Trip and Auxiliary Feedwater Actuation MONTH DAY YEAR YEAR ssoUENTIAL REYIBIoN NUMBER IIVMBER MONTH DAY YEAR FACILITYNAME DOCKET NUMBFR 12 14 1999 1999
- 009' 00 01 13 2000 FACILITYNAME DOCKET NUMBER 05000 OPERATING MODE (B)
POWER LEVEL (10) 100 P~~)g>fgi&@q~~jk.,"atra I 20.2201(b) 20.2203(a)(1) 20.2203(a)(2) (i) 20.2203(a)(2)(ii) 20.2203(a)(2)(iii) 20.2203(a)(2)(iv) 20.2203(a) (2) (v) 20.2203(a) (3) II) 20.2203(a) (3)(ii) 20.2203(a) (4) 50.36(c)(1) 50.36(c)(2) 50.73(a) (2) Q 50.73(a)(2)(ii) 50.73(a)(2)(iii) x 50.73(a)(2)(iv) 50.73(a)(2)(v) 50.73(a)(2)(vii) 50.73(a)(2)(viii) 50.73(a)(2)(x) 73.71 OTHER pecify in Abstract below or ln NRC Form 3BGA NAME Mark Ellington, Senior Analyst - Licensing TELEPHONE NUMBER Irnclvde Area Cade)
(919) 362-2057 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE SYSTEM SD COMPONENT MANUFACTURER Siemens-Allis REPORTABLE To EPIX CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE To EPIX YES (Ifyes, complete EXPECTED SUBMISSION DATE).
NO EXPECTED MONTH DAY YEAR On December 14, 1999 at 0353, the Harris Nuclear Plant (HNP) manually tripped the reactor from 60% power. At 0352 on December 14, 1999 the "A"Condensate Pump tripped due to a ground fault on the associated motor. The trip of the "A"Condensate Pump resulted in subsequent trips of the "A" Condensate Booster Pump and the "A" Main Feedwater Pump. The main turbine toad control circuitry sensed the trip of'the "A"Main Feedwater Pump and automatically reduced turbine power to approximately 60% (turbine runback). The reactor trip coupled with the loss of the "A" Condensate and Feedwater train resulted in a reduction of steam generator water levels. The main control room staff manually started the three auxiliary feedwater pumps in accordance with the applicable abnormal operating procedure. The steam generator levels continued to lower until the main control room staff manually tripped the reactor at approximately 40% steam generator level prior to reaching the steam generator low-Iow level automatic trip setpoint of 38.4%. Additionally, following the trip an auxiliary feedwater {AFW) actuation signal was generated due to the low-low steam generator level. However, as stated previously, the auxiliary feedwater pumps had been manually started during the transient therefore, the AFW actuation signal did not start any additional components.
Cause of this event: A ground fault on the "A"Condensate Pump motor caused a toss of the "A"Condensate and Feedwater train that resulted in low steam generator water levels and a subsequent manual reactor trip and AFW actuation.
Corrective actions include: The "A" Condensate Pump motor was replaced on December 16, 1999.
NRC FORM 366 (6.1996)
NRC FORM 366A (6-96)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION FACILITYNAME I1)
Harris Nuclear Plant, Unit 1 DOCKET 05000400 LER NUMBER IB)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 1999 009
00 PAGE I3) 2 OF 3
TEXT (Ifmore space fs reeofred, use adrenal copfes ofIVRCForm 366AJ I17)
I.
DESCRIPTION OF EVENT On December 14, 1999 at 0353, the Harris Nuclear Plant (HNP) manually tripped the reactor from 60% power.
At 0352 on December 14, 1999 the "A"Condensate Pump (EIIS SD-P) tripped due to a ground fault on the associated motor. The trip of the "A"Condensate Pump resulted in subsequent trips of the "A"Condensate Booster Pump'and the "A" Main Feedwater Pump. The main turbine load control circuitry sensed the trip of the "A Main Feedwater Pump and automatically reduced turbine power to approximately 60% (turbine runback).
The reactor trip coupled with the loss of the "A",Condensate and Feedwater train resulted in a reduction of steam generator water levels. The main control room staff manually started the three auxiliary feedwater pumps in accordance with the applicable abnormal operating procedure. The steam generator levels continued to lower until the main control room staff manually tripped the reactor at approximately 40% steam generator level prior to reaching the steam generator low-low level automatic trip setpoint of 38A%. Additionally, following the trip an auxiliary feedwater (AFW) actuation signal was generated due to the low-low steam generator level ~ However, as stated previously, the auxiliary feedwater pumps had been manually started during the transient therefore, the AFW actuation signal did not start any additional components.
After the reactor trip, the main control room staff stabilized plant conditions and recovered inventory in all three steam generators.
All safety related systems responded as expected during this event.
The HNP Condensate and Feedwater design includes two redundant trains each with a condensate pump, a condensate booster pump, and a main feedwater pump. The condensate pumps take suction from the main condenser hotwell. The discharge from both condensate pumps combine and flow through the condensate polishers to the suction of both condensate booster pumps. The discharge of the condensate booster pumps flow through a series of feedwater heaters and combine with the discharge of the heater drain pumps to provide a suction to the two main feedwater pumps. The main feedwater pumps discharge flow through two additional feedwater heaters and then is separated into three lines to provide inventory to the three steam generators.
HNP condensate and feedwater system design causes a trip of the associated condensate booster pump and main feedwater pump when a condensate pump trips.
At HNP, steam generator levels lower (shrink) during a rapid load reduction such as a turbine runback or reactor trip, The shrink coupled with the loss of the condensate and feedwater train resulted in steam generator levels becoming unacceptably low and caused the main control room staff to manually trip the reactor to avoid challenging the automatic steam generator low-low level reactor trip.
II.
CAUSE OF EVENT A ground fault on the "A" Condensate Pump motor caused a loss of the "A"Condensate and Feedwater train that resulted in low steam generator water levels and a subsequent manual reactor trip and AFW actuation.
IIL SAFETY SIGNIFICANCE There were no actual safety consequences as a result of this event. The plant was manually tripped as required for plant conditions and all safety related systems functioned as required during the transient. This report is being submitted pursuant to the criteria of 10CFR50.73(a)(2)(iv) for the reactor trip and the automatic actuation of AFW (Engineered Safety Feature signal).
NRC FORM 366 (6.9BI
NRC FORM 366A (6-98)
LiCENSEE EVENT REPORT (LER)
TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION FACIUTYNAME I1)
Harris Nuclear Plant, Unit 1
DOCKET 05000400 LER NUMBER I6)
YEAR SEQUENTIAL REVISION NUMBER 1999
009
00 PAGE I3) 3 OF 3
TEXT (ifmore spece Is required, use edditfonsf copies ofNRC Form 388AJ l17)
IV. CORRECTIVE ACTIONS The "A"Condensate Pump motor was replaced on December 16, 1999.
V.
SIMILAR EVENTS There have been no previous reactor trips due to a ground fault on a condensate pump motor.
l.
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