IR 05000302/1988009
| ML20155K181 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 06/06/1988 |
| From: | Julian C, Lawyer L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20155K168 | List: |
| References | |
| 50-302-88-09, 50-302-88-9, NUDOCS 8806210130 | |
| Download: ML20155K181 (27) | |
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@ MGt,q UNITED STATES
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, jo NUCLEAR REGULATORY COMMISSION 8%' / REGION il-I O D 0 101 MARIETTA STREET, ! ATLANTA, GEORGt A 30323
,,.~....j Licensee: Florida Power Corporation 3201 34th Street, South St. Petersburg, FL 33733 Docket No.: 50-302 License No.: DPR-/2 Facility Name: ,
Crystal River 3 Inspection Conducted: March 28-April 8, 1988 Inspection Team Leader: b Mb [+ d b L . LdW'ye r p D~a te' Si gned Inspection Team Members: M.' Archer M. DeGraff P. Kellogg W. Lyon T.-Stetka Approved by:
C. A. Julia6/ Chief
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hn Date Signed Operations Branch Division of Reactor Safety SUMMARY Scope: This special, anr.ouited inspection was conducted in the area of review of the adequacy of Emergency Operation Procedure Results: Although numerous technical and human factors deficiencies were identified, the Emergency Operating Procedures were found to be adequate for continued operation of the facilit The licensee committed to review the deficiencies and take prompt corrective action to resolve them. No violations or deviations were identifie i PDR ADOCK 05000302 Q DCD p
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REPORT DETAILS Persons Contacted Licensee Employees
- D. deMontfort, Nuclear Operations Engineer
- D. Green, Licensing Specialist
- D. Harper, Regulatory Specialist
- V, Hernandez, Supervisor, QA Surveillance
- B. Hickle, Manager, Nuclear Plant Operations and Maintenance
- L. Moffatt, Nuclear Safety Supervisor
- E. Renfro, Director, Nuclear Ops. Matl. & Cont *W. Rossfeld, Managec, Nuclear Compliance
- K. Vogel, Nuclear Operations Engineer
- Wilgus, Vice President, Nuclear Operations
- R. Wittman, Nuclear Operations Superintendent Other licensee employees contacted included engineers, technicians, operators and office personne NRR Attendees
- Regan, Chief Human Factors Assessment Branch, NRR
- T. Stetka, Senior Resident Inspector
- J. Tedrow, Resident Inspector
- Attended exit interview on April 8, 198 . Exit Interview The inspection scope and findings were summarized on April 8, 1988, with those persons indicated in paragraph 1. The inspectors described the areas inspected and discussed in detail the inspection findings listed below. Although proprietary material was reviewed during this inspection, no proprietary material is contained in this repor No dissenting comments were received from the license Note: A list of abbreviations used in this report is contained in l Appendix Item Number Status Description / Reference Paragraph IFI 302/88-09-01 Open Resolution of placekeeping deficiencies (paragraph 5).
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IFl 302/88-09-02 Open Licensee's implementation of an E0P cross reference document (paragraph 5).
IFI 302/88-09-03 Open Correction of technical discrepancies con-tained in E0Ps as outlined in Appendix IFI 302/88-09-04 Open Correction of human factors discrepancies contained in E0Ps as outlined in Appendix IFI 302/88-09-05 Open Correction of labeling discrepancies between E0Ps and panel indications as outlined in Appendix IFI 302/88-09-06 Open Licensee needs to re-perform E0P table top review and procedure walk-throughs to upgrade the V&V program (paragraph 6).
IFI 302/88-09-0 Open Licensee will review S0TA training and upgrade if necessary (paragraph 6).
IFI 302/88-09-08 Open Licensee needs to formalize the program for ongoing evaluation of E0Ps (paragraph 8).
IFI 302/88-09-09 Open Re-validation of the E0Ps when the plant specific simulator is operational (paragraph 8). Background Information Following the Three Mile Island (TMI) accident, the Office of Nuclear Reactor Regulation developed the "THI Action Plan" (NUREG-0660 and NUREG-0737) which required licensees of operating reactors to reanalyze transients and accidents and to upgrade emergency operating procedures (E0Ps) h tem I.C.1). The plan also required the NRC staff to develop a long-term plan that integrated and expanded efforts in the writing, reviewing, and monitoring of plant procedures (Item I.C.9). NUREG-0899,
"Guidelines for the Preparation of Emergency Operating Procedures,"
represents the NRC staff's long-term program for upgrading E0Ps, and describes the use of a "Procedures Generation Package" (PGP) to prepare E0Ps. The licensees formed four vendor type owner groups corresponding to the four major reactor types in the United States; Westinghouse, General Electric, Babcock & Wilcox, and Combustion Engineering. Working with the vendor company and the NRC, these owner groups developed Generic Technical Guidelines (GTGs) which are generic procedures that set forth the desired accident mitigation strategy. These GTGs were to be used by the licensee in developing their PGPs. Submittal of the PGP was made a requirement by i Confirmatory Order dated February 21, 1984. Generic letter 82-33,
"Supp'.ement 1 to NUREG-0737 - Requirements for Emergency Response Capability" requires each licensee to submit to the NRC a PGP which includes:
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(i) Plant-specific technical guidelines with justification for differences from the GTG (ii) A writer's guide (iii) A description of the program to be used for the validation and verification of E0Ps (iv) A description of the training program for the upgraded E0P From this PGP, plant specific E0Ps were to have been developed that would provide the operator with directions to mitigate the consequences of a broad range of accidents and multiple equipment failure Due to various circumstances, there were long delays in achieving NRC approval of many of. the PGPs. Nevertheless, the licensees have all implemented their E0Ps. To determine the success of the implementation, a series of NRC inspections are being performed to examine the final product of the program; the E0Ps. The objective is to perform table top reviera, simulator exercises where possible, and in-plant walk-throughs of the E0Ps with licensed operators to verify their adequacy. The E0Ps are considered to be adequate for use if they can be understood and performed successfully by the operators and they incorporate the accident mitigation strategy developed by the appropriate vendor specific owner grou This inspection report represents findings, observations, and conclusions regarding the adequacy of the E0Ps. It did not, as a matter of intent, review whether the E0Ps thus prepared conformed to the NRC staff's long-term program for upgrading E0Ps and whether those E0Ps had been properly prepared using a PG The success level of licensees in following the PGP submitted to NRC is a regulatory issue that will be dealt with on a case-by-case basi Although some licensee's E0Ps strayed far from their PGP, that issua is of secondary importance to this inspection effort. The purpose of this inspection is to verify adequacy of the E0Ps for continued safe operation of the facilit . E0P/GTG Comparison The inspectors performed a comparison of the Crystal River E0Ps against the Crystal River AT0G. From this comparison the inspectors determined that a significant change in procedural organization occurred between the Crystal River AT0G and the Crystal River E0Ps. For example, tha Crystal River AT0G contains the following major responses to a reactor trip or to the conditions which should have resulted in a reactor trip:
III A Lack of Adequate Subcooling Margin III B Lack of Heat Transfer III C Excessive Heat Transfer III D Steam Generator Tube Rupture
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The licensee has not developed specific procedures corresponding to the first three major responses listed above. However, the licensee has incorporated the major actions of these responses into other Crystal River E0Ps. For example, comparable material can be found in procedures such as AP-580, Reactor Trip, and AP-380, Engineered Safeguards Actuatio The Oconee AT0G was submitted to the NRC by the BSW Owner's Group as the B&W generic model for the development of E0Ps. The licensee has in place documentation to support the development of the Crystal River AT0G from the approved Oconee AT0G. However, there is no documentation describing the development of the Crystal River E0Ps from the Crystal River AT0G. Based on the results of this inspection NRC observes that the basic elements of the Oconee AT0G have been incorporated in the E0P There were no violations or deviations noted in this are . Technical Adequacy Review of the E0Ps The inspectors determined by review of the procedures listed in Appendix A that generally the vendor recommenced step sequence is followed, even though this is not immediately evident when examining the E0Ps. Review of the procedures has established that the AT0G guidance is contained within each of the E0Ps as applicable. The general priority of treatment and order of steps are maintained at the expense of additional bulk in the procedure Placekeeping deficiencies were identified during control room walk-throughs of the E0Ps. Operators typically use loose sheets of paper or their fingers as placekeeping aids. Additionally, when questioned on the problem of placekeeping, the operators indicated that they would remove the individual procedures from the notebooks and place them on the des This is undesirable, particularly when one considers that the E0Ps are not stapled and can easily become intermixed, separated, or lost. This is an indication of a placekeeping deficiency. The licensee has committed to resolve these placekeeping deficiencies. Resolution of this issue will be identified as IFI 302/88-09-0 The inspectors verified that entry conditions into the procedures were
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clearly identified and could be easily followed by operations personnel.
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The scenarios postulated during the procedure walk-throughs resulted in multiple transfers and cases of simultaneous use of several different A0Ps and E0Ps. Although this is a complicated method of operatior., no examples of significant performance error were identified. The licensee's use of notes and cautions within the E0Ps is generally clear, appropriate, and
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placed in the correct location. The inspectors verified that the priority l
of accident mitigation appears to be maintained in the licensee's E0Ps i even though the organization is quite different froni the AT0G.
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The licensee has not developed documentation to identify major deviations between the Crystal River E0Ps and the AT0G. No review was performed to ensure that any identified deviations have adequate technical justification nor could it be determined that any safety significant deviations were documente Currently, the licensee has no document in place to cross reference operator action points for plant parameters to where they occur in procedure The . licensee has committed to implement an E0P cross reference docur.ent. Resolution of this issue will be identified as IFI 302/88-09-0 There were no violations or deviations noted in this are . Review of the E0Ps by In-Plant and Control Room Walk-throughs In-plant and control room walk-throughs of the emergency, abnormal and verification procedures listed in Appendix A were conducted to ensure that:
Procedural guidance L clear enough that operator confusion and/or error can be avoide * Actions required by the procedures, either locally or in the control room, can be accomplished using existing available equipment, instru-mentation and control There are two sets of emergency and abnormal procedures maintained in the control room at all time These procedures were verified to be of the latest revision and free of any handwritten change As a result of these walk-throughs no violations or ' deviations were identifie However, many discrepancies in the' areas of technical content, writer's guide adherence and human factors were note Technical discrepancies are identified in Appendix B, while writer's guide and human factors discrepancies are noted in Appendix The licensee has comitted to correct the discrepancies identified in the aforementioned appendice Appendix B discrepancies will be identified as ,
IFI 302/88-09-03 and Appendix C discrepancies will be identified as IFI 302/88-09-0 Generally, there are a large number of inconsistencies (listed in Appen-dix D) between the instrumentation and control labeling on the control board and the nomenclature used in the procedures. The licensee has committed to perform a complete nomenclature review as part of the current work in progress to change control board labelin Appendix D discrepancies will be identified as IFI 302/88-09-0 _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _. . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ __ _ _ ___
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Most of the problems identified by the NRC are inconsistencies between equipment label designation and the nomenclature used within the proce-dure There were also minor problems in sequencing of steps and discrepancies with the Writer's Guide. While individually, most of the specific problems were relatively minor, the large number of these problems indicates that E0P verification has not been adequately completed. This finding of the NRC was supported by operator interview All the operators agreed that while the procedures were basically sound, there were still many minor flaw These problems appear to be due to deficiencies in the validation and verification (V&V) program. While the Crystal River V&V program appro-priately consisted of table-top reviews, control room walk-throughs, and scenario based simulations using the B&W simulator, the control room walk-throughs were largely performed by the author of most of the procedures. This is a departure from good V&V practice, which calls for-V&V activities to be performed by different personnel, preferably working for different management. The extensive familiarity of the author with his work makes it difficult for hin. to identify the types of discrepancies uncovered in this inspection. TM licensee needs to repeat the table-top and control room walk-throughs using different personnel. These personnel should be familiar with the Writer's Guide and generally familiar with plant design. However, they need not be licensed operators, whose expertise might prevent them from identifying these types of problem The licensee has committed to re-perform the table-top reviews and procedure walk-through Resolution of this issue will be identified as IFI 302/88-09-0 The scenario postulated by the NRC during the walk-throughs of AP-380 required local operator action to establish long term heat removal with DH Under the scenario used, following a massive core damage accident, the reactor coolant would be highly contaminated. If this coolant were circulating in the MU lines, the resultant high radiation levels in the auxiliary building may prohibit access to the locked breaker for valve DHV-3 at the 95 ft elevation. This breaker must be actuated to open the DHR drop line and permit operation in the DHR mod This problem can be eliminated if procedural guidance directs the operators to close the breaker for the valve prior to changeover from the BWST suction to the RB sump. The licensee committed to make this procedure change. This was the only deficiency of this type found during the inspectio The licensee should postulate and review additional examples of core damage accidents
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l There was a strong indication that certain Shift Operation Technical L Advisors (50TAs) lacked sufficient training to adequately perform their function. Two out of four 50TAs, who demonstrated the use of VP-540 and l
VP-580 during walk-throughs, exhibited an unfamiliarity with the l procedure. Some S0Tas made incorrect assessments of the proposed symptoms l and were unaware of various plant instrument indications. Examples of this include not knowing that computer group 59 indicated the current plant heat balance, not knowing that the indicated RCS code safety valve
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position was by acoustical means vice tailpipe temperature measurement, lack of . knowledge concerning the emergency bus ' configuration (i.e. ,
control board ES breaker alignment) and not being able to calculate subcooling margi The licensee committed to promptly review the training of the S0TAs and upgrade it if found necessary. This matter will be- reviewed during a future inspection (IFI 302/88-09-07).
There were no violations or deviations noted in this are . E0P User Interviews Ten interviews were conducted by the NRC inspection team. The personnel interviewed consisted of four Nuclear Operators (three Reactor Operators and one Senior Reactor Operator), one Assistant Nuclear Shift Supervisor, two Nuclear Shift Supervisors, two Operations Technical Advisors, and one Chief Nuclear Operator. The purpose of these interviews was to determine if the current E0Ps satisfy the needs of the operational personne Personnel were questioned on the adequacy of the E0Ps in the following areas:
Adequate staffing levels for performance of the E0P Problems in physically using the E0Ps from personal experience or observation, or from discussions with other Knowledge of technical discrepancie Adequacy of training on E0P The results of these interviews can be summarized as follows:
The operators felt the level of detail contained in the E0Ps is adequate and compatible with their level of knowledge.
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The operators felt the E0Ps are relatively easy to use.
l, Placekeeping aids are felt to be sufficient to allow the use of several procedures simultaneously. (Note: This opinion expressed by the operators was not substantiated during the actual walk-throughs where difficulty was encountered in place keeping between several procedures.)
Communications during use of E0Ps both within the control room and with other areas of the plant are adequat Operators felt there is adequate staffing to perform the E0P Operators felt the current E0Ps represented a significant improvement over previous version .
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Almost all pecole interviewed incorrectly believed the words "ensure" and "verify," to have the same meaning contrary to the Writer's Guid Additionally, the operators did not understand that a conditional statement preceded by WHEN is to be considered a holding point, unless otherwise stipulated with a continue statement. These inconsistencies indicate a need for further operator training in the conventions and definitions contained in the Writer's Guid The operators felt that current - procedures are free of major technical errors, but that they do contain a fairly large number of small discrepancies which need to be correcte They felt that inconsistencies in nomenclature between the procedures
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and equipment designation are common. However, they stated that the E0P nomenclature is consistent with operator usag Du ring walk-throughs no operator confusion was observed as a result of these inconsistencie In conclusion, the operations staff was confident that the E0Ps would function effectively during an actual even There were no violations or deviations noted in this are . Ongoing Evaluation of the E0Ps Administrative controls were reviewed to determine if the licensee has an acceptable program in place for a continuing evaluation of E0Ps. The licensee's controls on revising procedures based on changes to plant equipment, operator feedback for improvement, and revisions to the vendor GTGs were reviewe The original E0Ps were reviewed in accordance with the licensee's V&V program that is detailed in their E0P Writer's Guide. Other than the two year periodic ruiew of procedures that is required by the STS and implemented by by a procedure) procedure program AI-400,evaluation for continuing the licensee has no of E0P Theformal two year(i.e., covered periodic review program is essentially a "paper" review of the procedures and does not require a re-validation of the procedures by the use of walk-throughs or use on a simulator. The licensee does have informal methods for feedback to the E0P These methods include:
When operator training either on the simulator or during plant walk-throughs is conducted and discrepancies are identified, these discrepancies are fed back to the appropriate section for procedure correctio During analysis of a plant event, if discrepancies with the E0Ps are identified, the discrepancies are fed back to the appropriate section for procedure correctio The licensee has recently issued a new procedure Al-402A, Writer's Guide for Abnormal, Verification, and Emergency Operating Procedures, which encompasses the original Writer's Guide. This procedure requires a
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procedure validation to be performed (which is the licensee's V&V program)
for any initial procedure issued but does not provide for an on-going revie To provide for an on-going E0P evaluation, the licensee has committed to develop a formal evaluation program (possibly in Al-402A).
In addition, when the licensee's plant specific simulator becomes opera-tional (presently scheduled for September 1989), the licensee has committed to re-validate the E0Ps on this simulator. Resolution of these commitments will be identified as IFIs 302/88-09-08 and 302/88-09-09, respectivel . Writer's Guide For ops The plant staff has initiated and developed a Writer's Guide for station operating procedures. This Writer's Guide was reviewed. Lack of such a document has been a human factors concern at nearly all plants due to the format discontinuity between Ps and referenced ops. Development of this document and further improvement as suggested below should add to the efficiency of E0P usage by maintaining the formats of the ops and E0Ps very simila However, many format differences are allowed by the Writer's Guide for the ops and E0P Some differences are required due to the different uses of these procedures, but they should be as consistent as possible. For example, notes and cautions are formatted differently according to the two Writer's Guides. This could result in a negative transfer of training with the ops degrading performance of the E0Ps. The licensee acknowledged the inspector's comments and agreed to consider the matter.
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APPENDIX A LIST OF EMERGENCY OPERATING PROCEDURES E0P TITLE AP-250 Radiation Monitor Actuation, Revision 0 AP-330 Loss of Nuclear Services Water, Revision 3 AP-360 . Loss of Decay Heat Removal, Revision 1 AP-380 Engineered Safeguards Actuation, Revision 8 AP-450 Emergency Feedwater Actuation, Revision 10
.AP-460 Steam Generator Isolation Actuation, Revision 5 AP-513' Toxic Gas, Revision 4 AP-525 Continuous Control Rod Motion, Revision 0 AP-530 Natural Circulation, Revision 6 AP-545 Plant Runback, Revision 0 AP-580 Reactor Trip, Revision 8 AP-660 Turbine Trip, Revision 4 AP-770 Emergency Diesel Generator Actuation, Revision 8
'AP-961 Earthquake, Revision 2 AP-990 Shutdown from Outside Control Room, Revision 2 AP-1075 Violent Weather, Revision 9 EP-140 Emergency Reactivity Control, Revision 4 EP-220 Pressurized Thermal Shock, Revision 3 EP-290 Inadequate Core Cooling, Revision 6 EP-390 Steam Generator Tube Leak, Revision 5 VP-540 Runback Verification Procedure, Revision 1 VP-580 Plant Safety Verification Procedure, Revision 8
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APPENDIX B TECHNICAL COMMENTS This appendix contains technical comments, observations and suggestions for E0P improvements made by the NRC inspectors. Unless specifically stated, these comments are not regulatory requirements. The licensee agreed to evaluate the comments and take appropriate action. These items will be reviewed during a future NRC inspectio . AP-360, Loss of Decay Heat Removal, Revision 1 Step 3.2; Containment integrity should be established under any condition that involves the potential for a significant release of radioactive material from the fuel and not just for the leak into the
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R Step 3.3; The licensee should consider isolation of the RCS, if a leak is determined to exist, as the initial step in leak location thereby preventing a further inventory loss from the RCS. Connec-tions to the RCS can then be reestablished one at a time, while maintaining inventory and cooling, Step 3.4; For entry into this procedure, the reactor coolant system would have to be less than 280 F with pressure less than 230 psi This would provide for little OTSG cooling to La uvailable. The most likely cooling method would be via the spent fuei cooling system as discussed in step 3.5. Therefore, the step 3.4 method of cooldown, which would apparently only apply after attempts to establish other methods for cooldown have failed and a plant heatup is occurring, should be placed later in the procedure, Step 3.4; RCPs will only be available over a narrow range of potential applicatien of procedure AP-360. Under some conditions, it is not necessary for the RCS to be filled and vented for the OTSGs to be useful. Under two phase conditions, the OTSGs can provide a valuable temporary cooling function without feedwater being available due to the heat capacity of the contained inventor Step 3.7; This step direct. the closure of reactor building sump valves DHV-42 and DHV-43. Con /idering the conditions for entry into this procedure, there appears to he no reason why these valves were open (neither the procedure nor plant conditions require it).
Therefore this step appears to be unnecessary.
l This step also directs LPI cooling by injecting from the BWST into i the RCS. As presently directed by the procedure, it appears that the
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LPI will become deadheaded if there is no outlet from the RC Therefore the procedure should direct operators to provide a l
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Appendix B 2 discharge path from the RCS if necessary (e.g., manually opening the PORV to assure cooling water flow through the RCS), Step 3.9; This caution applies to LPI cooling with the RB sump in use. It is not applicable to water from the BWS Step 3.9; The intent of this step is not clear. It apparently assumes that LPI is not available from the BWST and that HPI is supplying the flooding water via the BWST. If HPI and BWST are not available, there appears to be no source for flooding the RB. By the time this step is entered, it appears to assume that the BWST is unavailable; therefore, if this is correct, it cannot be a source of water, Step 3.10; This step should be preceded by a caution warning opera-tors to be aware of and watch for indication of LPI pump cavitations due to low RB sump leve . AP-380, Engineered Safeguards Actuation, Revision 8 PPA investigations consistently identify loss of injection capability during transfer to the recirculation mode as a significant contribu-tor to core melt probabilit AP-380 step 3.29 addresses switching of HPI suctions from the BWST to the DHR pump discharge side. If for some reason the DHR pump (s) was (were) not developing head, immediate MVP pump damage could result. Because this operation is critical, the licensee should revise the procedure to include a caution statement prior to step 3.29 to warn the operators to ensure tufficient reactor building sump level prior to DHR recirculation initiation and proper DHR operation prior to individually switching HPI suctions. The same comment also holds for EP-290, Inadequate Core Cooling, steps 3.5 and 3.6 and any other location in the procedures where the same conditions exis Conflicting instructions should be resolved and correcte For example, AP-380 steps 3.39 and 3.14 are similar in that the PORV is to be opened, yet 3.14 addresses only the RCS pressure limit of 2300 psig. Step 3.39 instructs the operator to open the PORV before exceeding 2300 psig whereas 3.14 uses a pressure greater than 2300 psig. VP-580 step 2.1, requires that RCS pressure be greater than or equal to 2300 psig, and uses the PORY or high point vents for pressure control. See also EP-290 step Step 3.10; The licensee should consider replacing "IF PORV is NOT open, THEN close RCV-11" by "IF PORV is closed, THEN close and reopen RCV-11". The negative stateJnent is not consistent with the Writer's Guide, Step 3.16; High point vent operation should be addressed if PORV i operation is not obtained or does not provide the desired results.
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Appendix B 3 Step 3.18; The comment regarding preference of RCP-18 should be a note prior to the instruction to start one RC Steps 3.20 and 3.24; Step 3.14 could have resulted in opening of the pressurizer vent if the PORV is not available. Steps 3.20 and 3.24 should address this possibilit Step 3.30; This step should be performed earlier in the procedure so that chemistry results will be available prior to the use of the sump, Step 3.33; the action taken as a result of step 3.33 should be completed after the requirements of step 3.34 have been satisfie Step 3.34; The procedure should ensure that high point vents are close Step 3.35; A step should be added to deal with the possibility of insufficient cooling as a result of actione taken in steps 3.34 a nd/or . 3. 3 Step 3.36; A step should be added to deal with containment pressure increasing following termination of spra . Step 3.37; The direction under this step is unclear in that when the SS0D is notified that VP-580 is completed, AP-380 requires a transfer to OP-209, Plant Cooldown. This exit point may be inappropriate in that all the actions required under AP-380 may not have been complete Step 3.39; The caution prior to step J.38 states HPI cooling must be established prior to any opening of the PORV. Yet step 3.39 requires opening the PORV prior to exceeding any of several conditions. The conflict should be resolve . AP-450, Emergency Feedwater Actuation, Revision 10 Step 3.20 is based on knowing hot well levels. The wide range hot well level gage in the control room is out of service. Control room operators indicated it had been inoperative for "a long time" and there were no immediate plans to return it to service. The licensee indicated that readings from local indicators of hot well levels could be obtained in less than five minutes and, at this stage of the procedures, this would not be a highly time critical step. Either the defective instrument should be repaired, or the procedure should explicitly indicate that hot well level should be determined locall . AP-513, Toxic Gas, Revision 4 Step 2.2; This step requires that the operator ensure that dampers, ,
including AHD-2 and AHD-99, are closed. Operators are not sure
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Appendix B 4 whether a blue status light will illuminate if both AHD-2 and AHD-99 close or if only AHD-2 closes. This understanding is aggravated by the incorrect labeling on the ligh Step 3.9; This step directs operation of a potentiometer, however, there are no . instructions on which way to tu'n the potentiometer to achieve the desired result. Since the potentiometer is labeled with numbers, the procedure should provide information (e.g., turning the potentiometer toward 10 will increase flow) that would tell the operators the effect each direction of the potentiometer would have on flo . AP-525, Continuous Control Rod Motion, Revision 0 Step.3.8; The procedure reference to technical specification 3.1.1.6 for safety rods and 3.1.3.5 for regulating rods is incorrect. Safety rods are discussed under 3.1.3.5 and regulating rods are discussed under 3.1. . AP-530, Natural Circulation, Revision 6 Step 3.24; The reference to Enclosure 2 for the natural circulation cooldown curve is incorrect. The correct curve is Enclosure . AP-580, Reactor Trip, Revision 8 Step 2.3; The operator is instructed to initiate emergency boration by starting CAP-1A or CAP-1B, opening CAV-60, and establishing maximum letdown. This may provide a slow response. The licensee should consider a more rapid boration if needed, such as by use of HPI from the BWS Step 2.11; This step instructs the operator to close the block orifice bypass valve. This is incorrect if emergency boration is underwa . AP-660, Turbine Trip, Revision 4 Step 2.3; This step directs closure of the MSIVs. Closure of more than one MSIV requires a mandatory reactor trip. Therefore this step should direct operators to trip the reactor and refer them to the reactor trip procedure (AP-580). AP-990, Shutdown from outside the Control Room, Revision 2 There were no calibration stickers on some of the instruments on the RS If these instruments are not in proper calibration there is a significant possibility for confusing and misleading the operator at l the RS The inspection team found no calibration stickers on RC-5B-TI4-2 or RC-48-TI4-2, and could not find documentation that j
these had been recently calibrated. The resident inspectors will follow up on this item.
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Appendix B 5 The pressurizer level instrument on the RSP is not temperature com-pensated. In the control room there are two instruments, one compen-sated and one non-compensate The compensated instrument is used unless the unit is in cold shutdown. If the unit is not in cold shutdown this would lead to a significant difference between the compensated instrument ir the control room and the non-compensated instrument on the RSP. There is nothing in the procedures, or in the RSP labeling that warns the operator of this potential difference. A note to this effect should be included in the procedure and a-label added to the RS Step 3.6; The procedure requires isolation of letdown from outside of the control room to be performed at the RSP. However, transfer of control for these valves from the main control board to the RSP does not occur until step 3.1 The procedure should be revised to provide operations personnel the necessary information to isolate letdown if they are at the remote shutdown panel prior to step 3.1 Step 3.21; The procedure states that if letdown cannot be estab-lished then decrease make-up flow. The details column states this can be accomplished by minimizing or isolating seal injectio During procedure walk-throughs, operations personnel indicated they would use MVV-31 to accomplish this action. The procedure should be clarified as to the preferred method for decreasing make-up flo . EP-140, Emergency Reactivity Control, Revision 4 Step 3.5; This step states "IF RB is. occupied, THEN evacuate RB."
Walk-throughs of the procedure indicated some confusion among the operators as to how to determine if the RB Evacuation Alarm should be sounde This should be resolved, and the procedure changed to clearly reflect the required actio Step 3.6; This step directs the operator to stop all deborations.
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l The licensee should consider closing all connections to the RCS except those connected with boration that is underway. Then the operator can selectively open connections and determine the source of the dilutio . EP-290, Inadequate Core Cooling, Revision 6 Steps 3.8 and 3.9; References to clad temperature are of no use to the operator and should be removed. It is sufficient that the operator be instructed to reference the proper ICC region and react accordingl Step 3.14; The RCP start permissives should be provided here to be consistent with other procedure .
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Appendix B 6 Step 3.15; The licensee should consider reproducing this caution on the following facing pages because of the generality of the instructio Step 3.17; Operators have indicated they do not perform EM-308 and have no need for information from that procedur The licensee should consider deleting this reference in EP-29 Step 3.18; Guidance should be provided that the best indication is the one listed last in the ste . EP-390, Steam Generator Tube Leak, Revision 5 There are several steps within the procedure which require the operators to monitor and maintain parameters based on current plant conditions. These parameters include subcooling margin, fuel pin compression limits, 0TSG levels and steaming requirements, and emergency cooldown limits. These items have been included within the procedure as enclosures or tables on the facing pages. To be consistent throughout the procedure when a reference to these parameters is made the appropriate table or enclosure should be annotated within the ste Examples of this deficiency can be found in steps 3.15, 3.18, 3.2 Step 3.7; The procedure requires the operator to open one or more HPI valves to maintain pressurizer level. The procedure should be revised to include the use of MUV-24 fi rs t , thus reducing the possibility of thermal shoc Step 3.19; The procedure requires the operator, if RCPs are not operating, to maintain RCS pressure above the natural circulation curve and increase cooldown to less than or equal to 50 degrees per hou The wording of the cooldown requirements appears confusing and should be revise Step 3.19; The procedure requires the operators to refer to AP-530, Natural Circulation, Enclosure To reduce the number of procedures the operator would be required to be in at once, a copy of Enclosure 1 from AP-530 should be included in the procedur Step 3.36; During procedure walk-throughs, operations personnel were unsure as to what the nonnal steaming requirements for an OTSG with both a tube leak and steam leak would be. The procedure should be revised to include the steaming requirements for an OTSG in this condition, Step 3.36; The procedure states that if a steam leak is identified in the same OTSG that has a tube leak, and the steam leak is in the reactor building then allow the OTSG to steam to the reactor buildin A statement should be included in the procedure to inform
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Appendix B 7 operators that due to the steam leak, localized temperature increases could cause' instrument error . VP-540, Runback Verification Procedure, Revision 1 Step 1.3; These details are too general. . The first detail, refer-ring to STS 3.1.3.1, is essentially repeated in step 1.4 which provides specific and useful guidance for control rod alignment.
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Therefore this detail should be deleted in step 1.3. The remaining details dealing with the STS limit for RCP operation need to be clarified such that the person performing the verification knows what needs to be verified in each of the STS sections liste For example, STS. 3.3.1.1 addresses the operability of the RP The intent of this step (i.e., whether the verifier should be checking all RPS instruments or specific instruments) is not clea Step 2.2; This step requires reference to STS 3.3.1.1. The reason for reference to this STS is not clear. The step needs to be clari-fied to specify what should be verified. The same comment applies to the DETAIL section of this ste The verifier is referred to Computer Group 59, however, there is no guidance as to what in Computer Group 59 is to be verified, Step 3.1; This step directs the observation of radiation monitors for trends. This step should direct the observation of the radiation monitor recorder since trends are not easily determined on a monito Step 4.2; This step refers to AI-500, Step 2.4 as a means of deter-mining the reporting requirements. Step 2.4 applies to the documen-tation of a reactor trip or shutdown and therefore does not appear to apply to a plant runbac . VP-580, Plant Safety Verification Procedure, Revision 8 There appears to be no specific termination or exit criteria delineated within the procedur The licensee should revise the procedure to include these item Step 1.7; The licensee should examine this instruction for accuracy particularly with respect to the inequality sign and operator instructions for SGT Step 2.8; The licensee should consider the following wording for the last item to better reflect expected response: "WHEN OTSG PRESS is lowered, THEN verify Tc, incore TEMPS, and Th lower." Step 2.11; Recording of P-T data should be more often during transi-ent conditions. (One of the 50TAs indicated plotting should not be initiated until the plant has stabilized - an incorrect decision since the information is most needed when the plant is in a transient condition.) longer term plotting should also be considered since the
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Appendix B a information is useful in following plant state during- the entire proces , i ,
e.- Step 3.5; This' step appears-inconsistent, and is something that ordinarily would be done by the operating personnel. The licensee should examine this step to determine if the S0TA is expected 't'o '
perform this ste Step 4.3; The hotwell level instrumentation provided .on the main control board reads in inches and the E0P references fee Page 9; This figure should be improved by showing acceptable regions and by providing contrast between the plotted information and the
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APPENDIX C n s WRITER'S GUIDE AND HUMAN FACTOR ISCREPANCIES
> The following are short descriptions of discrepanciis (betwcen the Writer's Guide and E0Ps or of discrepancies identified in the E0Fi.: The licensee agreed to avaluate these comments and take appropriate actiork \ These items will be reviewed during a future NRC inspectio , GIhralProblems j j/ 2 i , A>1arne number of labs, ling inconsistencies betw3en the procedures and control room instrumen:ation and controls were identified. These are listed in Appecpx s There are onf choihs 'of the E0Ps and APs in the control roo )
An additional copy cduld;be provided for use when multiple procedures
,, , are being performed, for use by the S0TA when performing VP's, and as
' p hick-up for the current copie Instructions to perform the same actions appear at a number of locations within the E0Ps. Frequently these instructions are worded differentl Furthermore, sometimes the actual steps to be taken are inappropriately differen The licensee should examine all procedures and ensure consistency. An example of this is given below
, p 6er AP-380. Additional examples may be found in AP-580 step m d'EP-290 step )
d, f E0Ps often inappropriately reference other pro'cedures as information
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sources without indicating the specific step or pages to be refer-ence The step or page location should be specified whenever <
practical. Some examples are given in AP-380 belo Graphs and figures often dc not ccn't in grid lines, are sometimes unclear, and occasionally contain extraneous information not needed by the operator. These difficulties should be resolved and the procedures correcte Examples are AP-380, page 30, which has no grid lines, EP-290, which has no grid lines and 'contains references ,
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to clad temperature which are of no immediate use to the operator, and EP-220, Enclosure 1, which contains handwritten infornntion and has n.o grid line . EP-290, Inadequate Co*e Cooling, Revision 6 Etep 0.8; The caution located before this step should be clarifie , AP-380, Engineered Safeguards Actuation, Revision 8
) tStep 3.15; This step references AP-530 but does not indicate what hteps in the AP are applicable. The specific steps should be 4
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Appendix C 2 Step 3.19; The intent of this action is that AP-530 should be used for subcooled natural circulation and AP-380 should be continued for inadequate subcooling margin. The wording should reflect the inten Similar actions are indicated in -different steps, but are not, consistent with each othe Fot' cumple:
,i (1) Step 3.3; This step contains instructions to "Start full HPl" and step 3.38 is to "Establish fell HPI". The actions of several of the steps are identical, although the wording is differen i (2) Other actions diffe For example, "Ensure greater than or equal to 2 MUPs and their cooling water pumps are running" versus "Start second MVP and its cooling water pumps and Ensure HPI flow is greater than 500 gpm."
(3) Step 3.3 is followed by 3.6 which has the operator balancing flow in the four injection lines. Step 3.38 has no correspond-ing actio Step S.21 refers to EP-390 but does not indicate which steps in the EP are applicable. The specific steps should be included in the referenc f AP-450, Emergency Feedwater Actuation, Revision 10 Step 3.8; The sub-step saying "G0 TO AP-380" is located before the sub-step starting HPI. This would prevent HPI from starting for an unknown period of time, Step 3.12 refers to OP-605 Section There is no Section 9.0 in OP-60 This section is referred to in a number of other steps in this and other procedure Step 3.14 contains two logically separate steps, with some of the details referring to one step and some the other. This is not
consistent wi h the Writer's Guid . AP-460, Steam Generator Isolation Actuation, Revision 5 Step 3.6; This step contains two separate actions. The first action requires response if both emergency feedwater and main feedwater are not available. The second action require, response if emergency feedwater is not available. The Writer's Guide states that only one l idea should be presented in en action step. The step should be t
l revised to be consistent with the requirements of the Writer's Guid Throughout the procedures the Once-Through Steam Generators are referred to as OTSGs. On the main control room boards these are
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referred to as Steam ' enerators (STM GEN). On the RSP they are referred to as OTSGs, Step ~ 3.19 -indicates that MSV-55 is located on DPDP-8 It is actually located on DPDP-8 , Step 3.27 instructs the operator to "trickle feed 0TSG." No
quantitative definition is given to tell the operator what flow would constitute a reasonable "trickle." When questioned about this, the operators indicated that a flow of less than 100 gpm would be reasonable. The procedure should be changed to define a "trickle" quantitativel . AP-513, Toxic Gas, Revision 4 In step 3.12 the operator is referred to AH-35-FR to verify the proper flow. This recorder is labeled AH0-32-FIR on the back pane . AP-530, Natural Circulation, Revision 6 Step 3.3; The logic statement when reproduced as a recurring step on the facing pages was not capitaligd and underlined, Enclosures 1, 2, and 3; The graphs do not contain grid lines and contain handwritten informatio . AP-990, Shutdown from Outside Control Room, Revision 2 The "B" "RELAYS ENERGIZED" light on the RSP is covered with a green lens cap, while the "A" light has no lens cap. Since plant color
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conventions call for a red indicator to ind'eate energization, this
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discrepancy should be correctea, No steam tables we.re aveliable in the RSP roo / .
' AP-1075,; Violent Weather, Revision 9 Step 1 refers the operator to Enclosure 1 for defin < > ions of entry
, conditions. This list is relatively short and should be included on I the entry condition pag In several steps (e.g. , 3.3, 3.4, 3.5, and 3.6) the operator is instructed to perform an action (such as ensure SF Pool Missile
- Shields are in place). The procedure does not indicate who should be l contacted and/or responsible for performing these task In Step 3.4 the procedure instructs the operator to perform pre-start-checks on each EGD It does not instruct him to do this task l concurrently, so the Writer's Guide would indicate that the remainder l
of the procedure would not be completed until the EGDG pre-start l
checks are completed. This is clearly in error. Most of the l
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App.ndix C 4 i
subsequent steps should be initiated immediately and performed
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concurrently with completion of the rest of the procedure. The
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procedure does _ not indicate that any of these steps should be performed concurrently, which would extend the time required to complete this procedure.
4 Enclosure 2 lists members of _ the violent weather preparation
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APPENDIX D N0MENCLATURE DISCREPANCIES IDENTIFIED BY NRC E0P INSPECTION TEAM f
Step or _
Procedure Page Procedure Nomenclature Label on Equipment AP-330 WDT-5A and WDT-5B DW Transfer Pumps WTP6A, 6B AP-330 SW Surge Tank Nuc. Serv. Clg. Water Surge Tk. Level AP-360 3.7, 3.10 LPI Suctions from RB DHP-1A, RB Sump, DHP-18,
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Sump RB Sump AP-360 LPI Control Valves DHHE-1B Dis, DHHE-1A Dis AP-360 LPI Suctions from BWST DHP-1A BWST Suct., OHP-18 BWST Suc AP-360 3.7, 3.10 LPI Discharge to RCS DHP-1A LP Inj., DHP-1B LP 1
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In AP-360 3.8, 3.11 HPI Suctions from BWST BWST to MVP-AP-360 3.11 LPI Discharges to ifPI DHV-1A to MUPS, DHV-1B to Suction MUPS AP-360 3.13 CFT Outlets CFT-1A Outlet Iso AP-380 MVP suction valves from BWST MUV-58 MUV-58 Hi Press. Suc MVV-73 MUV-73 BWST to MVP
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MUV-23 MUV-23 HP Inj. Loop A MUV-24 MUV-24 HPI Loop A
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MUV-25 MUV-15 HPI Loop B MUV-26 MUV-26 HPI Loop B
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AP-380 A and B HPI Channels RC1 HP 1 RC1 RC2 HP 1 RC2
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RC5 HP 1 RC5
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< Appendix 0 2 Step or Procedure Page Procedure Nomenclature Label on Equipment AP-380 A and B RBI Channels RBI RB ISO RBI et al., for six references total AP-380 BSV-3 BS HDR Inlet Is BSV-4 BS HDR Inlet Is AP-380 3.10 P.CY-11 (not identified) RCT-1 to RCV-10 RCS-13 PZR Spray Block RCT-1-to RCV-10 Iso, to MUHE-1A
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MUV-38 A Letdown Cooler Inlet Isolation MUV-39 B Letdown Cooler Iso. to MUHE-1B Inlet Isolation MUV-498 C Letdown Cooler Iso to MUHE-1C Inlet Isolation MVV-49 Letdown Isolation High Temp Bypass DHV-3 (not identified)
AP-380 3.14 RCV-11 PORY Block See 3.10 PORY DPDP 4B
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RCT-1 Relief RCV 10 .
PZR Vent (not identified) -
RCV-11 PORY Block Valve See aboya
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AP-380 3.17 MVP Recire MVP's Recir MVP Recire. Valves AP-380 3.23 CFV-5 CFT-1A Outlet Iso.
DFV-6 CFT-1B Outlet Is AP-380 3.29 LPI Suction DHV-34 DHP-!A BWST Suc DHV-35 DHP-1B BWST Suc DHP-1A DH Removal Pump A DHP-1B DH Removal Pump B
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LPI discharge to HPI Suctions <
DHV-11 DHP-1A to MUPS HDV-12 DHP-1B to MVPS HPI Suctions from BWST MUV-58 Hi Press. Suc MUV-73 BWST to MVP
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-Appendix 0 3-Step or Procedure Page Procedure Nomenclature Label on Equipment
.AP-380 3.32 LPI Suctions from BWST DHV-34 DHP-1A BWST Suc DHV-35 DHP-1B BWST Suc AP-380 3.35 E FW-56 E FV-56 EFW-58 EFV-58 EFW-55 EFV-55 EFW-57 EFV-57 AP-450 p OTSG Level Stm. Gen. Ly .12 FWV-398 Startup Control FWV-39B SU FW Yl pg.13 MFW Flow FW to Stm. Ge .15 EFW EFV CDHE-3 Inlet Disch. Is .23 EFW Control Valves EFV Control Valves AP-460 p OTSG Press. . Stm. Gen. A/B Pres p Subcooling Margin Saturation Margin 3.19 MSV-55, DPDP-8A MSV-55, DPDP-8B 3.21 SU Control Valves Stm. Gen. B SU FW Vi .23 OTSG Level Chart Stm. Ge .26 - EFT-2 Level EF Tank Level 3.29 A SV Block A SU FW Block B SU Block B SU FW Block 3,30 ASV-5/204 ASV-5 and ASV-204 EFW Control Valves EFV-55 through 58 AP-513 Heating & Ventilation Control Complex HVAC Control Panel ,
AP-513 ARD-2 D2 CC Rel. Air Damper AHD-99 Closed AP-513 "CC Damp Override" Damp Override AP-513 "AH-193-FC" Cntrl. Complex Recir Damper AP-513 3.12 AH-35-FR Top of Recorder:
Control Ccmplex supply Air RB & AB Air Sy Bottom of Rec 1rder:
Supply & Exh. Air Monitoring AH-032-FIR
' A-Control Complex Air l Supply l
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Appendix 0 4 Step or Procedure Page ' Procedure Nomenclature Label on Equipment AP-513 "CC Damp Override" Damper Override AP-513 3.12 AH-35-FR AH0-32-FIR AP-530 % Level 50%/30" AP-530 HPI Valves HPI Loop A,8 HP Inj. Loop A AP-530 3.26 MFW Block Main FW Block AP-530 3.26 LL Block Lo Load FW Block AP-530 3.26 Startup Control SU FW VLV e AP-530 3.26 Cross-Tie FW Disch. Crosstie
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AP-530 3.26 SU Block SU FW Block AP-770 ES 480V 480V ES Bus A,B AP-770 ES-MCC-3AB ES-MCC-3a2 ,
AP-770 Seal Injection Control RC Pumr Total Seal Inlet Valve Flow AP-770 Seal Injection Block RC PP Seal Supply Valve AP-770 SW Raw Water Pres Nuc. Serv. Sea Wtr. Pump Disch. Pressure AP-990 CRD Bkr. A Feeder No. I CRD Bkr. B Feeder No. 2 Letdewn Isolation Valve Letdown Cir. Iso.
, RCV-11 PORV Block Valve RCT-1 Bicek Valve
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3.12 "AB" and 7,co-safety" Transfer switch "AB" and i controis Transfer SW non-safety 3.14 "Voltage Adj" VP Adjust 3.16 Mur Suction Valves BWST HP Suct and BWST l to MVP AP-990 3.16 HPI Valves HP In .19 MUV-53 MVPP Recirc. 53 MUV-257 MVPP PP Recir .23 RCP Seal Return RCP Bleed Is Seal Isolation RB Bleed off Iso.
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. Appendix Step or Procedure 'Page Procedure Nomenclature Label on Equipment EP-390 3.3 page MUV Block Orifice MUV-51'also MV-3-MIC no 5 of 35 Bypass mention of Block Orific Bypass EP-390 3.5 page Comment same as above Coment same as above not 5 of 35 labeled Block Orifice Bypass EP-390 3.7 page- Comment same as above Comment same as above not 7 of 35- labeled Block Orifice Bypass EP-390 3.7 page MVP Suction Valves 58 - Hi Press. Suc of 35 MUV-58, MUV-73 73 - BWST to MVP None None MVV 23 Label differs MUV 23 indicates HP In from MUV 24, 25, & 26 Loop A other are HPI Loop A-B EP-390 Various Reference to OTSG* Steam Generator on Labels EP-390' 315 Per Spray PORV RCT-1 SPR Cntrl. RCV-14 RCT-1 Relief RCV-10 EP-390 Various 65% Level Level Select Pushbutton 50%/30" EP-390 3.25 MSV-55, MSV-56 EFP RCSG - Should have been Supply Deleted EP-390 3.27 Same Comment as Step Same Comment as 3.15 3.15 VP-540 MS Radiation Monitors 1) No labels on recorders 2) Recorder scale in linear. 0-100, meter face 1 9 logrithmic, 0.1-10 MR/h, no correlation between the tw ) Monitors labeled as:
A-1 RMG 25 (ADV MSV-25)
A-2 RMG 27 B-1 RMG 26
+0TSG is on Control Board for L Chan. EFIC Act. Bypass.
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Appendix D 6 Step or Procedure Page Procedure Nomenclature Label on Equipment B-2 RMG 28 (ADY MSV-26)
VP-540 RCP Seals and Dumpsters 1) RC-19A, PR-1(A)
RC-198,PR-1(C)
RC-19A,PR-2(B)
RC-198,PR-2(D)
2) No labeling on dumpster integrators 3) Recorder labeled:
RC Pump Seal Leakage RC-134-FIR (Dumpster Clics)
VP-540 RCDT Level RC Drn. Tnk. Level VP-540 MUT Level MU Tank Level VP-540 RB Sump Level RB Sump A Level RB Sump B Level Relief Valve Tailpipe R205 Press. Relief Vi Temp RCV-8 out Tem R206 Press. Relief Viv.-9 out Tem R207 Press. Relief Vi RCV-10 out Tem Note: This labelinc is not in agreement with the computer point VP-580 RML-1 RM-L1 VP-580 EFT-2 EF Tank VP-580 EFT-2 EF Tank l
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APPENDIX E LIST OF ABBREVIATIONS AP- Abnormal Procedure
'AT0G Abnormal Transient Operating Guidelines BWST Borated Water Storage. Tank CR Control Room-DHR Decay Heat Removal EGDG Emergency Diesel Generator E0P Emergency Operating Procedure ES Engineered Safeguards -
GTG Generic Technical Guidelines HED Human Engineering Deficiencies H7I High Pressure Injection ICC Inadequate Core Cooling -
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LPI Low Pressure Injection MSIV Main Steam Isolation Valve MVP Makeup Pump NRC Nuclear Regulatory Commission OP Operating Procedure OTSG Once Through Steam Generator PORY Power Operated Relief Valve PRA -Probabilistic Risk Assessment RB Reactor Building RCP Reactor Coolant Pump-RCS Reactor Coolant System RPS Reactor Protection System RSP Remote Shutdown Panel
SOTA Shift Operations Technical Advisor SS0D Shift Supervisor On Duty STM GEN Steam. Generator STS Standardized Technical Specifications VP Verification Procedure V&V Validation & Verification
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