IR 05000413/1987013
| ML20236L635 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 07/31/1987 |
| From: | Shymlock M, Linda Watson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236L607 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 50-413-87-13, 50-414-87-13, GL-82-33, NUDOCS 8708100313 | |
| Download: ML20236L635 (28) | |
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@ Mou UNITED STATES
. o NUCLEAR REGULATORY COMMISSION I
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REGION ll n
5 s)
_ 2 ATLANTA, GEORGI A 30323 j
101 MA RIETTA STREET, N.W.
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Report Nos.:
50-413/87-13 and 50-414/87-13
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Licensee:
Duke Power Company
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422 South Church Street Charlotte, NC 28242 Docket Nos.-
50-413 and 50-414 License Nos.:
NPF-35 and NPF-52 Facility Name:
Catawba 1 and 2 Inspection Con uc e'd: Nay 18-22, 1987 Inspector: Yb
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b L'. J. Watson, Tpm Leader Date Signed Team Members:
C. Vanderniet B. Breslau M. Lewis J. Bongarra, NRR Accompanying Personn_ei:_ L.,Defferding, PNL
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W Barnes, Battelle Approved by:
1 N
Q M. B. Shymlock, CtMef Date Signed
/4 Operational Programs Section Division of Reactor Safety
SUMMARY Scope:
This special, announced inspection was conducted to review the emergency operating procedures required to implement the requirements of NUREG-0737 and Supplement No.1 to NUREG-0737 as stated in Generic Letter No. 82-33, Recommendations and Requ'raments for Emergency Response Capability.
Results:
One violation, failure to provide adequate training on calculation of the subcooling margin (paragraph 7.c.1); and one deviation, failure to implement the commitments of the Procedures Generation Package (paragraphs 7 a.1, 7.b.3, 7.b.4, and 7.b.5, respectively) were identified.
8708100313 070B06 PDR ADOCK 05000413 G
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
- J. Hampton, Station Manager
- H. Barron, Superintendent of Operations
- J. Cox, Superintendent of Technical Services
- C. Hartzell, Compliance Engineer
- C. Schiffley, Licensing Engineer
- R. Sharpe, General Office, Licensing
- D. Towers, Shift Operations Engineer
- D. Simpson, Station Emergency Planner
- R. Maynard, Operations Engineer
- W.
Barron, Director of Operator Training
- J. Knuti, Operating Engineer, Document Development
- S. Cooper, Assistant Operating Engineer, Document Development
- J. Roach, Security Coordinator
- G. Swindlehurst, Supervising Design Engineer H. Lee, Design Engineer Other licensee employees contacted included engineers, technicians, operators, mechanics, security force members, and office personnel.
NRC Inspectors
- K. VanDoorn, Senior Resident Inspector
- J. Kreh, Region II
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on May 22, 1987, with those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection findings listed below.
No dissenting comments were received from the' licensee.
(0 pen) Deviation 413, 414/87-13-01.
Failure to Meet the Commitments of the Approved Procedures Generation Package. (paragraphs 7.a.1, 7.b.3, 7.b.4 and 7.b.5)
(0 pen) Violation 413, 414/87-13-02.
Failure to Provide Adequate Training on Calculation of the Subcooling Margin. (paragraph 7.c.1)
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspection.
A list of acronyms used in this report is provided in paragraph 13.
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3.
Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspection.
4.
Unresolved Items Unresolved items were not identified during this inspection.
5.
Approval of the Procedures Generation Package (PGP)
The PGP was submitted for NRC approval by a letter dated February 28, 1983. Subsequently, this PGP was superseded by a revised PGP submitted on June 1, 1983. In a letter dated February 22, 1984, the licensee clarified the revised PGP by stating that the NRC-approved version of the Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERGS),
namely, Rev. O to the ERGS, served as the starting point for the development of the plant-specific technical guidelines.
The NRC staff's evaluation of these submittals was documented in Supplement 2 of the Safety Evaluation Report (SER) dated June 1984.
The staff concluded in Supplement 2 of the SER that three items needed to
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be resolved before the staff could determine that the licensee's program for developing procedures in accordance with TMI Task Action Plan Item I.C.1 was acceptable. Two of these items were described in Section 13.5.2 of Supplement 2 of the SER.
Item 1 of Section 13.5.2 required the licensee to identify the safety-significant differences in the plant-specific technical guidelines from the NRC-approved WOG generic technical guidelines and to provide justification for these deviations.
By letters dated June 18 and July 25, 1984, the licensee provided descriptions of deviations from the WOG generic guidelines, including deviations from both the basic version and Rev.1 of the guidelines.
Item 2 of Section 13.5.2 required the licensee to provide additional information and/or clarification of the Writer's Guide, the validation /
verification program, and the Emergency Procedure (EP) training program.
".he licensee provided additional information in each of these areas by letter dated August 29, 1984. The third item, which involved verification by Westinghouse of low power and power ascension testing procedures, was completed by the licensee as described in Section 13.5.3 of Supplement 3 j
of the SER, dated July 1984.
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The NRC staff concluded in Supplement 4 of the SER dated December 1984, j
that the licensee's submittals provided an acceptable method for meeting the objectives of NUREG-0899, Guidelines for the Preparation of Emergency Operating Procedures, and that the Writer's Guide, verification / validation program, and EP training program were adequately addressed. In addition, i
the staff found that the submittals provided an adequate technical basis j
for the EPs subject to confirmatory information on ECA-1.2, LOCA Outside
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Containment. This information was submitted in a letter dated October 29, j
1985. The NRC staff concluded in SER Supplement 6 dated May 1986 that the
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i informatic". was adequate.
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3 6.
Emergency Procedure / Generic Technical Guideline Comparison The licensee utilized the generic Westinghouse Emergency Response I
l Guidelines (ERGS) (Rev. O and Rev. I where appropriate) as the base i
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documents for the development of plant specific Emergency Procedure Guidelines (EPGs).
The conversion of the generic guidelines into plant-specific EPGs was accomplished by incorporating: (1) the design differences between the Westinghouse reference plant and the licensee's facility; (2) the plant specific setpoints, limits and precautions; and, (3) the plant specific transient response.
The EPGs and ERGS were l
l utilized as general guidelines for the plant specific Emergency Procedures
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(EPs).
In addition, plant specific details such as equipment identifica-
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l tion, plant specific steps and references to appropriate plant procedures were included. Major deviations from the ERGS were reviewed and approved
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by the NRC as discussed in paragraph 5.
l The inspector compared the EPs and EPGs to the ERGS to assure that the broad spectrum of accidents and equipment failures included in the ERGS were addressed. The EPGs and EPs sufficiently addressed the ERG suggested l
scenarios and were found to be in accordance with the deviations approved I
by the NRC with two exceptions.
The first exception involves a procedure which was deleted from the Unit 1
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EP set and not included in the Unit 2 EP set due to the upgrading of the pressurizer power operated relief valves to safety grade components.
A 10 CFR 50.59 evaluation was conducted and the deletion was determined not to involve an unreviewed safety question.
The second exception involves the creation of a unique procedure, EP/1/A/5000/1E6, Steam Generator Tube Rupture Cooldown Using ND.
EP/1/A/5000/IE6 is a hybrid procedure compiling similar information from other procedures. A safety evaluation performed by the licensee concluded that the implementation of EP/1/A/5000/1E6 did not involve an unreviewed safety question.
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No violations or deviations we'e identified.
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Verification and Validation Program Review
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Verification Program
The approved Procedures Generation Package (PGP) included the
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procedure, Verification Process for Emergency Procedures, Rev.
1.
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The verification process consisted of two phases: the written
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correctness review; and, the technical adequacy review.
The purpose of the written correctness review was to ensure that the Emergency Procedures (EPs) conformed to the format and other principles as specified in the Writer's Guide.
The technical adequacy review was to ensure that the plant-specific EPs were technically accurate, t
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consistent with the plant-specific Emergency Procedure Guidelines (EPGs) and Emergency Response Guidelines (ERGS), ana included 'all appropriate licensing commitments.
Both phases of this review were required to be conducted prior to implementation of all EPs and changes to EPs.
1.
Compliance with Writer's Guide and Human Factors Review The inspectors reviewed the compliance of portions of twelve EPs to the approved Writer's Guide for Emergency and. Abnormal Procedures. The inspector utilized Rev. 6 of the Writer's Guide dated September 19, 1984 for this rev.iew.
During the review of.the selected procedures, numerous deviations from the Writer's Guide were noted.
The inspector discussed these examples in detail with the licensee and advised the licensee that similar deviations exist in many of the EPs.
The inspectors determined that, although each specific deviation identified had minor safety significance when taken as a single item, sufficient examples of deviations from the Writer's Guide were found to question the adequacy of the written correctness review described in the PGP, Verification Process for Emergency Procedures, Rev.
1.
In addition, during interviews with licensed operators, these deviations from the Writer's Guide affected the useability of the procedures.
The failure to
implement the requirements of the approved Writer's Guide is I
identified as an example of deviation 413, 414/87-13-01.
i Examples of these deviations are listed below.
Note that the
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abbreviation "RN0" indicates that the deviation is in the Response Not Obtained column of the EP.
a.
Logic Statements The structure of logic statements did not follow the guidance in Writer's Guide Sections 2.2.3.3 and 3.10.
Sections 2.2.3.3 and 3.10 describe the structure of i
conditional steps and indicate that if dependencies exist in actions, then "IF...THEN" and "IF NOT....THEN" are to be used. The inspectors noted the following discrepancies in the writing of EP logic statements.
(1) The following discrepancies were noted where the licensee used the phrase "Otherwise" rather than
"IF NOT".
The use of the term "Otherwise" is not
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I described in the Writer's Guide and could result in an
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unclear description of conditions.
EP/1/A/5000/1C5, Loss of Emergency Coolant l
Recirculation p.13, Step 13.a and 13.b RNO.
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EP/1/A/5000/1C3, Transfer to Cold Leg Recirculation p.7, Step 10.b RNO.
EP/1/A/5000/1E, Steam Generator Tube Rupture p.4, Step 4.b.
(2) The following logic statements were in an incorrect format.
Format errors included examples where the consequence preceded the antecedent, the conditional statement was obscured by the action step format, the step contained an "IF" without a "THEN" statement, or inappropriate conditional phrases were used:
EP/1/A/5000/IE, Steam Generator Tube Rupture p.2, Step 3.b RNO (last entry on page)
p.4, note preceding step 4.b p.4, Step 4.b p.7, Caution p.7, Steps 7.a RNO and 7.c RNO
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p.8, Step 9.a.1 RNO p.11, Step 11 p.12, Step 14.b RNO EP/2/A/5000/2A2, Loss of Core Shutdown p.1, Step 1.b RNO p.3, Step 7 EP/2/A/5000/1B, S/I Termination Following Spurious S/I p.23, Step 31 RNO EP/1/A/5000/10, High Energy Line Break Inside Containment p.4, Step 7.b (3)
In the following cases,
"IF" clauses were used without indicating the action for the opposite of circumstances, i.e., the "IF NOT" action:
EP/1/A/5000/1C6, LOCA Outside Containment p.3, step 5.c RNO, states that if feed and bleed of the reactor coolant system is required, then refer to EP/1/A/5000/1E4. The procedure does not specify the next action in the case where feed and bleed is not required.
EP/1/A/5000/1A, Reactor Trip Response p.4, bullet, 1).
There was no indication of action to take if Main Steam Isolation actuate _ - - _ - _ _ _ - _ _. _ _ _ _ _ _ _ _ _ _
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i (4) The following three logic statements were reviewed with a Senior Reactor Operator (SRO) during a control room walk-through conducted by the inspector, i
EP/1/A/5000/1C5, Loss of Emergency Coolant Recirculation p. 13, RNO, Step 13.a EP/1/A/5000/1A, Reactor Trip Response p.4, Step with the bullet, Substeps one and two.
EP/1/A/5000/1C3, Transfer to Cold Leg Recirculation p.7, RNO Step.10.b.
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was different from the manner in which the SRO stated he would perform the steps.
In the third sequence, the SR0 had a great deal of difficulty in determining l
which actions he should perform.
This indicates that I
unclear direction is being provided by the logic steps and, therefore, review of the logic steps is necessary.
b.
References to Other Procedures i
The inspectors noted the following discrepancies in regard to references in the EPs to other documents.
Writer's
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Guide Section 2.8.3 states that referring between procedures should be kept to a minimum and that it is better to reproduce small sections than to reference them.
The Writer's Guide also describes the use of "go -to" and
" refer to", and states that if possible the case number, sectinn number, and specific step to enter the document is to be referenced. The majority of the EPs reviewed did not include the step number of the referenced procedure.
Operators and staff members who were interviewed provided different philosophies as to where to start in a referenced procedure when the step number was not provided.
Some individuals indicated that the procedure was to be performed from the beginning.
Others stated that the procedure should be entered at the applicable procedure step.
(1)
Failure to use "go to" or " refer to":
EP/1/A/5000/1E, Steam Generator Tube Rupture p.8, Step 9.a RNO p.8, Step 9.a.2 RNO EP/1/A/5000/1C3, Transfer to Cold Leg Recirculation p.8, Step 12,a RNO
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7 EP/1/A/5000/1B, S/I Termination Following Spurious S/I l
p.22, Step 31 RNO 1st bullet.a
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EP/1/A/5000/1E1, Post Steam Generator Tutte Rupture Cooldown and Depressurization p.1, Step 1.1 l
EP/1/A/5000/01, Reactor Trip or S/I p.3, 1st bullet, RNO p.3, 2nd bullet (2) Failure to include step numbers in references to other procedures:
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EP/1/A/5000/IE, Steam Generator Tube Rupture p.5, Step 5 bullet RNO p.8, Step 9.a RNO p.8, Step 9.a.2 RNO p.10, table p.11, Step 10.f RNO p.11, Step 12 RNO EP/1/A5000/1C, High Energy Line Break Inside Containment p.1, Step 1.a. and c.
p.4, Step 5 RNO p.4, Step 7.d RNO p.6, Step 10 RNO p.7, Step 11 RNO p.8, Step 16.C RNO p.8, Step 17 RNO c.
Action Steps
The inspectors reviewed the format of action steps against l
the requirements of Writer's Guide Section 3.3 and noted the following problems in the presentation of action steps.
Action steps were not limited in some cases to one main thought.
Complex sentences should have been rewritten as several simpler sentences to assure that actions were not overlooked. Action steps were inappropriately included in notes although Writer's Guide Section 2.5 indicates that action steps are to be avoided in cautions and notes.
Examples included:
EP/1/A/5000/1B, S/I Termination Following Spurious S/I p.5, Step 5.d RNO p.22, Step 31 1st bullet and Step 31 1st bul.lel RNO
EP/2/A/5000/2D3, High PZR Press p.7, Step 16 RNO
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EP/1/A/5000/1C5, Loss of Emergency Coolant Recirculation p.1, Step 1.c RNO p.1, Step 1.d RNO p.7, Step 7.a RNO EP/1/A/5000/1A, Reactor Trip Response p.10, Step 12, 2nd bullet RNO EP/1/A/5000/1C, High Energy Line Break Inside Containment p.7, Step 14 d.
Emphasis Techniques The inspectors noted inconsistent use of emphasis techniques (e.g., bolding, capitalization, underlining, and various combinations) which detracted from the readability of steps and eliminated the effectiveness of any one technique.
(1) The requirements of Writer's Guide Section 2.7 for emphasis techniques were not met in the following procedures:
EP/1/A/5000/1E, Steam Generator Tube Rupture p.2, "Not" in caution at top of page should be underlined and in caps j
p.5, "No" in 1st bullet under Step 5 should be i
underlined and in caps
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p.5, Step 5, "ALL" does not need to be in caps or j
underlined p.9, RNO, 1st bullet, "no" should be underlined and in caps (2) Examples of inconsistent use included:
EP/1/A/5000/1E, Steam Generator Tube Rupture l
p.1, Step 1 RNO "WHEN" p.2, "0 PEN" in Steps 3.a.1 and 3.a.1 RNO (bolding)
p.2, "CLOSE" in Steps 3.a.2 and 3.a.2 RNO (bolding)
p.2, Step 3.b RNO, last sentence "when" p.4, "0N" in Note preceding Step 4.b and "0N" in Step 4.b (bolding)
p.7, Step 7.b RNO "CANNOT" p.14, Step 16.a RNO "can NOT" EP/1/A/5000/2C1 - Loss of Secondary Heat Sink Step 3.b.
The "AND" in Feed AND bleed is not a logic statement and should not be capitalize l
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Punctuation The Writer's Guide does not describe the use of a colon to introduce the condition of a component as found in the following examples:
i EP/1/A/5000/IE, Steam Generator Tube Rupture p.6, Step 6.a "N/R Level: >5%"
p.8, Step 9.a "At least two (2) NC Pump:
RUNNING."
p.14, Step 17.a "NC pressure:
STABLE OR INCREASING."
f.
Use of bullets Sections 3.11.3 and 3.14 of the Writer's Guide describe the
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use of bullets to indicate non-sequential steps or to identify equally acceptable steps. The following uses of j
bullets were not consistent with the Writer's Guide:
EP/1/A/5000/1E, Steam Generator Tube Rupture p.1, bullets are used incorrectly prior to the first caution and note.
p.5, bullets are used in the RNO column when the steps are intended to be performed in sequence.
g.
Spacing Appendix 6, Section 6.3 of the Writer's Guide requires the use of triple-spacing in a number of instances such as before and after notes and cautions, and between major steps.
The inspector found no instances in which these j
requirements were followed.
The following additional discrepancy was noted-I
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EP/1/A/5000/2C1 - Loss of Secondary Heat Sink Step 12, RNO. Logic statement is not a subset of 1st bullet and should not be indented, h.
Adverbs Section 3.1.4 of the Writer's Guide requires that ambiguous instructions be avoided. However, vague adverbs such as
" sl owl y" or rapidly" are often used in EPs.
Examples include the following:
EP/1/A/5000/1E, Steam Generator Tube Rupture p. 2, " slowly" in Steps 3.a.3 and 3.a.3 RNO p. 10, " rapidly" in steps 10.c and 10.c RNO i
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Writer's Guide Comments The inspectors noted that in some instances correction of Writer's Guide requirements rather than revision of the
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l procedures may be appropriate.
The licensee was provided
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I the following list of comments on the Writer's Guide:
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(1) The Writer's Guide does not define cases or subprocedures.
The Writer's Guide states that the purpose and the symptoms for all cases should be given at the beginning of the procedure.
The tinspector noted that the EPGs contain a cover page with useful information such as the procedure purpose, plant symptoms, and procedure entry points.
Not all EPs provide these types of information.
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(2) Page numbering is incorrect in the Table of Contents.
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(3) Overuse of capitalization, bolding, and underlining was described.
j (4) Guidance for conditional statements is inadequate.
(5) Enclosure 2, Appendix 1 of the Writer's Guide violates the Writer's Guide requirements in Section 2.9 by failing to have becn prepared in accordance with standard technical graphics practices.
(6) Color-coding of procedures for different units should be addressed in the Writer's Guide.
(7) Guidance for preparing tables, figures, and Critical Safety Function Status Trees should be substantially expanded.
(8) Action steps (even in passive voice) do not belong in notes or cautions.
(9) The same acronym or abbreviation should not be used to indicate different meanings, conversely, more than one acronym or abbreviation should not be used to indicate the same meaning.
(10) The Writer's Guide coversheet example is out of date with the coversheet currently being used.
2.
Technical Verification Review The inspectors compared the Westinghouse Emergency Response Guidelines (ERGS) and Emergency Procedure Guidelines (EPGs) to
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the Emergency Procedures (EPs) to determine that proper technical verification of the ERGS and EPGs had been completed.
The inspectors observed that in many cases the EPs did not correspond to the EPGs, however, except for minor items identified below, the EPs were more conservative than the EPGs.
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No~significant technical deficiencies were identified in the EPs during this review.
The following discrepancies and comments were discussed with the licensee:
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a.
EP/1/A/5000/1A1, Natural Circulation Cooldown, Retype 2 l
EPG, ES-0.2, Natural Circulation Cooldown (1) Step 1 of the EP is not contained in the EPG. The step instructs the operator to refer to OP/1/A/6100/02, Controlling Procedure for Unit Shutdown, and to perform all applicable steps concurrently with the EP.
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(2) A note at the beginning of the EPG states that at any time in the procedure a reactor cuolant pump (NC)
should be started if possible.
The note does not exist in the EP, instead, EP Step 2 states "If at any time a NC pump can be restarted, then..." The note appears to be a more effective way to inform the operator.
(3) The EPG does not contain a note that is in the EP af ter Step 2 stating the preferred starting order of reactor coolant pumps for pressurizer spray flow considerations. The technical importance of this note should not have been missed in the drafting of the EPGs.
(4) The pressurizer liquid space is added as a sampling point for a boron sample in the EP, Step 4, and is not found in the EPG.
(5) A caution added to the EP before Step 7, informing the operator about a lag in the cooldown response to any adjustments made in the rate of steam dumping, is not present in the EPG.
(6) A caution added immediately after >tep 7 of the EP, informing the operator to observe only wide range coolant temperatures, is not present in the EPG.
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(7) The method of dumping steam to the condenser is included in the EP but not in the EPG.
(8) RNO, Step 6.c, in the EPG instructs the operator to control feed flow as necessary. The direction is not included in the EP.
(9) Step 16 has been added to the EP telling the operator to continue cooldown.
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(10) Step 24 of the EP has been added to instruct the operator to verify reactor coolant system low-range pressure is on scale. This does not appear to be a necessary step.
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EP/1/A/5000/101, SI Termination Following Steam Line Break, Retype 4 EPG, ES 2.1, SI Termination Following Excessive Cooldown Step D.24.a. RNO, specifies maintaining containment spray until containment pressure is < 0.35 psig.
EPG, ES 2.1, Rev. 2, specifies < 0.3 psig.
The licensee stated that this change was an enhancement to the EP and was not a safety significant deviation. The licensee, however, does not have documentation to show thst an evaluation was conducted to make thi s determination.
This lack of documentation does not afford an auditable trail for future audits of verification for change in EP setpoints.
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EP/1/A/5000/10, High Energy Line Break Inside Containment, l
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EPG, E-1, Rev. 1, Loss of Reactor or Secondary Coolant.
l (1) No safety significant problems were identified; J
however, it is important to note that the EPG was written with a mitigating strategy different from that of the corresponding ERG. The change in strategy was
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approved by the NRC as a deviation. Thus, a step by
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step comparison was not appropriate. A step by step i
technical review was conducted and no safety f
significant problems were identified.
J (2) Step 5 of the ERG requires the operator to verify that pressurizer pressure is below the power operated relief valve actuation setpoint before manually closing its isolation valves.
This verification
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ensures that the operator does not dr.' eat a properly functioning power operated relief valve.
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corresponding EPG and EP do not require this l
verification. Discussions with the licensee indicated l
that the EP will be revised to include the additional check of pressurizer pressure.
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(3) EPG, Step 4.c, specified that, if the containment sump is not increasing, "if necessary, perform ECA-1.2, LOCA Outside Containment."
The EF only refers the operator to EP/1/A/5000/1C6, LOCA Outside Containment.
Containment sump level not increasing is an excellent indication that the LOCA is outside containment; and, therefore, the appropriate procedure should not be just a reference but should be entered.
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(4) Step 7.d of the EP refers to EP/1/A/5000/2C3 on high S/G 1evel; however, the corresponding EPG step does n6t refer the operator to a procedure.
(5) EPG Step 8.b, directs the operator to perform steps on containment checks and plant status evaluation (Steps 16 and 17) if safety injection (S/I) can not be terminated. The corresponding EP step, Step 9.d, also allows switchover to cold leg recirculation. Although the foldout page will allow cold leg recirculation at any time the criteria at e met, the two procedures should, where possible, be consistent in the evolution of operator actions.
(6) Step 9 of the EP directs the operator to check if S/I can be terminated.
If the termination criteria are met the operator should continte on with the procedure.
Step 9 does not tell the operator to actually terminate S/I.
The action step to terminate S/I is Step 13.
The separation of Step 9 and Step 13 introduces confusion.
(7) EP Step 12 is provided for information only and includes a note that describes the step as diagnostic.
Yet it is still confusing as to what actions the operator performs following completion of Step 12 if a feedwater line break is indicated.
The step should include a bullet to proceed with the procedure in the RNO section.
d.
EP/1/A/5000/2C1, Loss of Secondary Heat Sink, Retype 6 EPG, FR-H.1, Loss of Secondary Heat Sink (1) Step 1 of the EP replaced the caution with an action J
step that checks the entry condition and returns the operators to the entry procedure.
The inspectors noted that this change improved the procedure.
(2) The caution before Step 3 of the EP should be an action step.
The Steam Generator (S/G) wide range level was different from the EPG.
(<3% level versus
<7% in EPG.)
(3) hep 3.a of the EP indicates a value for S/G wide range level that is different than the EPG (same as in the caution discussed above).
(4) Steps 4.a and b of the EP indicates a S/G wide range level value different from the EPG. (10% in EP versus 15% in the EPG.)
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(5) Step 10.c of the EP did not contain a list of corresponding bypass control valves for operator reference.
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(6) A caution was added to step 14 of the EP which is good information for the operator, but would be more appropriate as a note.
(7) Step 15 of the EP is unclear. The value listed is the setpoint for main steam isolation rather than a rate.
j (8) The licensee stated that the EPs will be revised to reflect the new S/G wide range level values listed in l
the EPG.
b.
Validation Program The licensee's Procedure Generation Package (PGP) included a validation program for the Emergency Procedures (EPs). Validation of Emergency Procedures, Rev.1, required an evaluation to assure that the EPs provided sufficient and understandable information to the operator and were operationally correct.
The validation program required initial validation of the procedures and a formal on going
validation by review of comments and problems identified during the real or simulated use of the EPs.
The initial validation consisted of the completion of a Step-by-Step Walk-Through performed by a cold-certified operator as described in Section -6.1 of the validation procedure.
The Step-by-Step Walk-Through included a review of the procedure for adequate j
guidance, control room compatibility, and administrative adequacy.
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In addition, a Real-Time Walk-Through, performed by one or more
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cold-certified operators and a minimum of two observers, was performed as described in Section 6.2 of the validation procedure.
During the Real-Time Walk-Through, an opei stor exercised the EP by performing actions required by a scenario designed to test the usability of the EP. The observers were to identify deficiencies in procedure action steps, control room compatibility, procedure flow, personnel compatibility, and adherence to administrative requirements.
The on going evaluation, described in Sections 7.0 and' 8.0 of 'the I
validation procedure, was to include validation of the EPs through a feedback program to be conducted by Shift Supervisors and senior training instructors.
Comments from the actual performance of the EPs in the plant or on the simulator were to be forwarded to the Operating Engineer, Document Development, for resolution.
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l The inspectors reviewed eight EPs in detail and conducted control room walk-downs and interviews with licensed operators to determine if the procedures were technically adequate and operationally correct.
Portions of additional procedures were also reviewed. The inspectors identified the following discrepancies in the implementation of the validation program:
1.
Real-Time Walk-Throughs were not conducted for the following Unit 2 EPs as required by Section 6.2 of the Validation of Emergency Procedures, Rev. 1, in the approved PGP:
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EP/2/A/5000/1A1 Natural Circulation Cooldown
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EP/2/A/5000/2F3 Void in Reactor Vessel l
EP/2/A/5000/1C High Energy Line Break Inside Containment l
EP/2/A/5000/01 Reactor Trip or Safety Injection l
EP/2/A/5000/1D1 S/I Termination Following Steam Line Break l
EP/2/A/5000/2C1 Loss of Secondary Heat Sink EP/2/1/5000/38 Loss of All AC Power Recovery With S/I Required The approved validation program does not indicate that the Real-Time Walk-Throt gns were required to be performed only on Unit 1.
The licensee stated that Unit I and Unit 2 are essentially identical and that the Unit 1 Real-Time Walk-Through validation process would have identified problems with both l
the Unit I and Unit 2 procedures.
The inspectors reviewed differences between the units and concluded that Unit 2 design differences affecting the EPs could be validated by the Unit 2 Step-by-Step Walk-Through and incorporation of comments frcm the Unit 1 Real-Time Walk-Throughs.
l 2.
Members of the groups of operators and observers performing the
Real-Time Walk-Throughs of the following procedures were not I
documented l
l EP/1/A/5000/1A1 Natural Circulation Cooldown EP/1/A/5000/2F3 Void in Reactor Vessel EP/1/A/5000/1C High Energy Line Break Inside Containment l
EP/1/A/5000/01 Reactor Trip or Safety Injection i
EP/1/A/5000/1D1 S/I Termination Following Steam Line Break EP/1/A/5000/2C1 Loss of Secondary Heat Sink EP/1/A/5000/3B Loss of All AC Power Recovery With S/I Required Due to the lack of documentation, the inspector: could not verify that these Real-Time Walk-Ti roughs were performed in accordance with the approved validation program in regards to the numbers and qualifications of individuals.
The licensee stated that the review criteria had been met, but that the
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l documentation had been discarded.
The inspector interviewed
individuals involved in the Real-Time Walk-Throughs and did not identify any c.ses where individuals did not meet the qualifications.
3.
Sections 7.0 and 8.0 of the Validation of Emeroency Procedures, Rev.
1, require that after implementation, feedback on the useability of the emergency procedures during a real or simulated emergency and input from classroom discussion and in plant walk-throughs of the EPs be provided.
Feedback was required to be documented on Attachment 2, Comment Resolution Form, of the Validation of Emergency Procedures, Rev.
1.
No feedback from actual plant or simulator performance was evident in the documentation reviewed for the following procedures:
EP/1/A/5000/1A1 Natural Circubition Cooldown EP/1/A/5000/2F3 Void in Reactur Vessel J
The inspectors interviewed operators who indicated that training i
on the abeve procedures had taken place but the feedback forms j
were - not available.
The inspectors determined that while J
I feedback from plant or simulator performance may have been conducted informally, t, formal program was not conducted to assure appropriate feedback in accordance with the validation program.
Failure to document feedback of actual plant or simuistor performance of the EPs as required by Sections 7.0 and 8.0 of the validation procedure, Validation of Emergency I
Procedures, Rev.1, is a deviation from the commitments of the approved PGP and is an example of deviation 413, 414/87-13-01.
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4.
Real-Time Walk-Throughs require the completion of Attachment 4 of Validation of Emergency Procedures, Rev.1.
The purpose of Attachment 4, Validation Observation Form, was to provide a mechanism to record operator errors of omission, commission, and uncertainty. No copies of Attachment 4, Validation Observation Form, were found in the documentation for the'following EPs:
EP/1/A/5000/1A1 Natural Circulation Cooldown EP/1/A/5000/2F3 Void in Reactor Vessel EP/1/A/5000/1C High Energy Line Break Inside Containment EP/1/A/5000/01 Reactor Trip or Safety Injection EP/1/A/5000/101 S/I Termination Following Steam Line Break EP/1/A/5000/2C1 Loss of Secondary Heat Sink EP/1/A/5000/38 Less of All AC Power Recovery With S/I Required l
The licensee r.tated that Attachment 4 was not adequate and, I
therefore, not used. Failure to use Attachment 4 constituted a change to the validation progr.:m that was never documented. The failure to utilize Attachment 4 to assure that performance
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deficiencies were linked to the appropriate evaluation criteria j
is another example of deviation 413, 414/87-13-01. It was noted i
by the inspectors that the proper use of the attachment might have assisted the licensee in identifying the numan factor
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discrepancies discussed in paragraph 7 of this report.
5.
Real-Time Walk-Throughs require the completion of Attachment 2
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of Validation of Emergency Procedures, Rev. 1, for each deficiency identified during the walk-down.
Attachment 2, Comment Resolution Form, provided a - mechanism to document discrepancies; forward discrepancies to the Operating Engineer Document Development for resolution; and document the resolution of the discrepancy.
The copy of the procedure used in the walk-down contained hand writter, notes identifying the following discrepancies that were not listed on the Attachment 2 sheets I
attached to the walk-down EP.
Therefore, no record existed of I
the resolution of these discrepancies.
l EP/1/A/5000/1A1, Natural Circulation Cooldown
"G0 T0" was added to the first caution in the procedure.
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Operating Procedure number identified as missing from
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Step 2.
"plus 100 ppm" added to step 2.
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Comment stated that name of procedure needed to be added to
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the caution on page 2.
Valve name needed in step 8.
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Comment stated that Step 16 Response Not Obtained (RNO)
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Substep 3 needed to be moved.
The note preceding step 17 was identified as being unclear.
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Step 21.c RNO questioied how an action is to be
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accomplished.
Comment stated that Step 28.a required setpoint change.
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EP/1/A/5000/1C, High Energy Line Break Inside Containment The need for Step 5 was questioned.
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In Step 22 the breaker number was added.
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The need for Step 24 was questioned.
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Enclosure ID commented on adding a setpoint.
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Enclosure IF questioned references.
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EP/1/A/5000/1C2, Post-LOCA Cooldown and Depressurization Step 3.e An auto stop caution was questioned.
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Step 13.b.3 The need to indicate caution was questioned.
Step 15.b Reorganizing the step was recommended.
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Step 30 The need for a new action statement was indicated.
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i The inspector noted during the review that Attachment 2 was completed and indicated resolution for similar ioentified discrepancies in other procedures. A sample of the discrepan-f cies was reviewed against the EP and interviews were conducted to determine if the discrepancies had been resolved informally.
The inspector did not identify any cases where a safety signifi-cant comment had not been resolved, however, the significance of some comments was not apparent from the hand written notes.
The failure to track and document the resolution of discrepancies identified during the Real-Time Walk ~Ihrough on Attachment 2 as committed in the approved PGP is ar example of deviation 413, 414/87-13-01.
c.
Control Room Walk-downs The inspectors conducted walk-downs of the Emergency Procedures (EPs)
in the control room utilizing facility staff to evaluate the validation and verification program used by the facility.
The following discrepancies were noted during the walk-downs:
1.
Subcooling Margin Setpoints The inspector noted that a number of the EPs utilized 0 F as the minimum margin to ensure reactor coolant system subcooling and to verify that upper head injection' system injection and cold leg accumulator injection were not required.. The licensee stated that a formula for the error band was utilized in the subcooling margin monitor and in the Safety Parameter Display System (SPDS) to account for instrument error and support the i
l use of a 0 F subcooling margin setpoint.
The instrument error was calculated to be about 12 F at operating pressure and >50*F at low pressures.
Examples of procedures in which the 0*F subcooling setpoint is used included:
EP/1/A/5000/1E4 S/GTR With Continuous NC System Leakage:
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Saturated Recovery EP/1/A/5000/1El
~ Post - S/GTR Cooldown and Depressurization EP/1/A/5000/1A1 Natural Circulation Cooldown EP/1/A/5000/1E3 S/GTR With Continuous NC System Leakage:
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Subcooled Recovery EP/1/A/5000/1E6 S/GTR Cooldown Using ND EP/1/A/5000/IC1 S/I Termination Following High Energy Line Break Inside Containment J
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During interviews conducted in the control room with a Senior
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Reactor Operator (SRO), the. SRO indicated that the subcooling I
margin could be determined using standard steam tables.
The inspector interviewed a second SRO who also stated that, in addition to other methods such as the SPDS and subcooling monitor, the standard steam tables could be used to calculate
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subcooling margin.
The use of the steam tables alone does not allow for instrument error to be reflected in the calculation t
of subcooling margin. Therefore, with a calculated indication of 0 F sLbcooling from the steam tables there could be a 12 F to greater than 50 F error in the calculation depending on reactor coolant system pressure. The licensee indicated that a corrected curve for manual calculation of subcooling margin was available ir the control room and the SR0s should have utilized this curve. Tre training program for licensed reactor operators did not address the potential problem with the use of standard steam tables in the determination of degrees of
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subcooling.
The fai'ure to provide adequate training to assure the imple-mentation of emergency procedures in regard to the calculation of subcooling margin is identified as an apparent violation (413, 414/87-13-02).
In accordance with the Commission Policy Statement on Training and Qualification of Nuclear Power Plant Personnel (50 Federal Register 11147 (March 20, 1985)), a Notice of Violation will not be issued at this time. The Policy states that licensees who are making reasonable efforts in develuping and implementing the INP0/NUMAtC programs described in the Policy statement will generally not be cited for violations related to these programs, provided the violations, whether or not identified by NRC, are appropriately corrected in a timely manner.
The licensee's corrective actions will be examined during future inspections.
2.
Labeling and Identification of Equipment Several of the aumerical values required to be read by operators were too specific to be read from some gauges (e.g., Wide Range T Hot).
However, these values are obtainable from computer displays in the detail called for in the procedures.
If the computer is unavailable, some steps in the procedures would be difficult to perform or would require the operator to interpo-late values. This could cause added error in the actual values.
Steps should be taken to ensure gauges are clearly marked or readable at the values specified in the EPs.
Some component labe'.s or location information in the procedures
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are not consistent with labels in the control room. For example
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EP/1/A/500/1E, Steam Generator Tube Rupture, p.16 refers to EMFs 26 (27, 28, 29) and also refers to a "STMLINE Hi Rad" annunciator.
The EMFs are not labelled in this manner in the
control room and the annunciator is labeled a s "HI RAD. "
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j Components identified in all EPs should contain the same labeling as found on the main control board to eliminate any confusion
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Examples of incorrect labeling or missing identification were identified in the following procedures:
EP/1/A/5000/1A1, Natural Circulation Cooldown.
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Step 5.b.
Step should have read "NC Makeup Control Mode
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Select Jwitch".
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Step 8.
Specific instrumentation names were not given to operator to verify specific parameters.
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Step 10.c.
The labels of the four specific annunciators to
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be.-bserved should have been included.
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Step 10.b.
This step should have read " BLOCK" not l
" BLOCKING".
Step 25. The valve number should have been added to the
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step.
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Step 29.b.
Incorrect valve labels were listed.
l EP/1/A/5000/1C, High Energy Line Break Inside Containment.
Steps 2.c and 6.a.
The steps should contain the label "ND
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Flow to C-Legs" to provide greater operator guidance.
Step 11.
The step should contain the label " Cold Leg
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EP/1/A/5000/01, Reactor Trip er Safety Injection Step C.S.
The location of the SI Injection Actuated status
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light was incorrectly identified as IMC-1. The status light j
should be labelled as 151-13.
I RNO Step C.S.a. The step reads "S/G Pressure (s):
<725
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PSIG." Panel IMC-2 S/G instruments are marked indicating 710 psig and are not graduated to accurately read 725 psig.
RNO, Step D.3.g.2.
The nomenclature of the third bullet
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which reads "M-2,
-3, -10, -11" was confusing to operators.
This description was for status light grid coordinates and was inconsistent with similar information used in the other EPs.
Three reactor operators were consulted before it was determined that the nomenclature referred to the status lignts.
RNO, Step D.6.a.
Valve noun names used in the EP were i
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different from the control board labels for the following
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valves:
ICA-11A, 7A, 9B; ICA-64, 52, 48; 1CA-50A and 54B.
RNO, Step D.9.a.
The step reads, "If PZR press <2315 psig
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then:".
The instrument scale is graduated in 10 psig l
increments; therefore, the value of 2315 psig stated in the EP must be estimated.
Step D.15.a.
The step reads, "PZR level: >5%."
This value
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must be estimated because instrumentation is graduated in 0.2% increments.
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The inspectors ' discussed these findings with the licensee and emphasized that i ncon si ster," and missing identification of specific components could lead to operator di f ficulty and confusion during the use of the EPs. The licensee acknowledged these concerns and stated that the items would be reviewed, i
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3.
For the following procedures, the Unit 1 and Unit 2 EPs were not consistent.
EP/1/A/5000/1A1, Natural Circulation Cooldown l
EP/2/A/5000/1A1, Natural Circulation Cooldown Enclosure 1, Part A, of the Unit I and Unit 2 procedures were different. The Unit 1 procedure required the operator to go to EP/1/A/5000/01, Reactor Trip or Safety Injection, for S/I actuation at >1955 psig and to go to AP/1/A/5500/05, ECCS Actuation During Plant Shutdown, for S/I actuation at <1955 psig.
The Unit 2 procedure only referenced the nperator to EP/2/A/5000/01, Reactor Trip or Safety Injection.
After discussion with the licensee, the Unit 1 procedure was determined to be correct.
The licensee indicated that the Unit 2 procedure would be corrected.
d.
Additional Observations l
1.
The inspector determined that there was a clear distinction between the Unit I and Unit 2 EPs. Each set of EPs were stored in separate drawers in the control room and were color coded and labelled clearly.
2.
The EPs in the centrol room included a master copy of each EP and working master copies of each EP.
The EPs appeared to be l
positively controlled, easily accessible, and clearly identified I
as EPs.
The inspector reviewed the corrective action for a finding identified in licensee Surveillance Report dated May 13, 1986, which indicated that the control room copy of i
EP/1/A/5000/1C4, Transfer to Hot Leg Recirculation, was retype
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Retype #2 was current at that time.
The inspector verified that Retype #2 had been placed in the control room.
5.
The licensee has instituted a 24-hour telephone " Hot Line" that is available to facility personnel to call in problems that may be encountered with the use of the EPs.
The Hot Line problems are documented and are evaluated to determine the proper course of corrective action.
Information concerning the final resolution of the problem is forwarded to the originator for his
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information.
This appears to be an effective way to promote prompt input from operators or others who identify problems with an EP.
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8.
Technical Competency of Responsible Emergency Procedure Reviewers T,:e inspector reviewed Station Directive 4.2.1, Rev. 23, Development, Approval and Use of Station Procedures, which expands upon information presented in the licensee's Administrative Policy Manual (APM) Sections 4.1, 4.2. and 4.3.
Also reviewed were the requirements specified in Reactor Safety (RS) procedure RS-003, Technical Verification of Nuclear
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Station Emergency Procedures and Guidelines, Rev. O.
Ac ytionally, the
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inspector reviewed the education and experience background of the I
individuals who developed, wrote and technically verified the Emergency
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l Procedure Guidelines, i
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concept was used in the development and implementation of the Emergency l
Pr3cedures (EPs) and that the developing writers and reviewers were l
cognizant of the requ;rements for EP format and technical content.
No violations or deviations were identified.
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9.
Training a.
Lesson Plans
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The inspectors reviewed lesson plans and handouts developed for operator training on Emergency Procedures (EPs).
Specifically, the
inspectors reviewed training material associated with EP/1/A/5000/1C, l
High Energy Line Break Inside Containment. The lesson plans did not l
specifically indicate that the EPs were provided to the operators during training, however, the licensee stated that the EPs were utilized in the training classes.
The initial lesson plan for EP/1/A/5000/1C, Rev. O, dated June 26, I
1985, was found to be general in nature, providing a very brief I
explanation or justification for only the " major steps" of the procedure.
The lesson-plan failed to address such items as exit conditions, foldout pages, cautions, and contingency actions (" response not obtained").
In addition, the lesson plans failed to reflect the stated obje:tives of listing equipment required to be locally operated and discussing the equipment, switches, and indications used to implement the procedure. A general review of other lesson plans developed during the same time period indicated similar deficiencies. The student handouts were merely a photocopy of the lesson plan contents, providing very little additional guidance
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The inspector also reviewed a more current lesson plan and associated handouts for the same procedure.
The more recent revision of the handouts, Rev.1, dated April 16, 1987, had improved substantially over the training material discussed above. The handouts included a more detailed step by step explanation of operator actions, providing technical bases for manipulations where appropriate.
The inspector observed that the handouts were essentially a reiteration from the EPG Reference Document, Section 3.
The inspector observed that the EPG Reference Document had not been updated and did not accurately reflect the EPs.
Consequently, the use of this document as a training tool for EPs could introduce inconsistencies. The inspector believes that the concept of providing operators with explanations of procedure manipulations enhances training, however, the inspector was (
also concerned that the handouts may have introduced confusion due to
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the inconsistencies with the EPs.
The handouts should be corrected l
to accurately reflect the EPs.
No violations or deviations were identified.
b.
Simulator Scenarios The inspector reviewed simulater scenarios utilized for initial demonstration of the adequacy of the EPs and for the requalification training program for liter. sed operators.
The simulator training program for initial exercises demonstrates single failures for each EP.
The program also includes a number of multiple failures to exercise the more complex procedures.
Each simulator scenario worksheet includes forms that are used to evaluate the operator's performance during the scenario.
The requalification training program includes a list of malfunctions to which each operator is exposed during training. This list includes annual and biennial control manipr'ation requirements.
These simulator scenarios appear to adequately train and evaluate the operators on the use of EPs.
The inspector interviewed a reactor operator who stated that FP/1/A/5000/01, Reactor Trip or Safety Injection, was not usually exercised completely on the simulator. The inspector discussed with the licensee the concern that the EPs were not fully exercised on the simulator.
The licensee stated that the original scenarios were developed to exercise the majority of the paths of the EPs and that subsequent training was conducted on selected scenarios to meet the
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requirements of Technical Specification 6.3, Training.
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No violations or deviations were identified.
10.
Plant Specific Values The inspector determined that from June 1984 through January 1986, the licensee did not have a formal Emergency Procedure Guideline (EPG)
setpoint document.
Control of setpoints, although informal, had continuity through this period since setpoints were _ established and revised through the Reactor Safety group of the Nuclear Production Department, the same organization responsible for the technical accuracy review of the EPGs. This organization has since been p1&ced in the Design Engineering Department.
Formal control of setpoint documents under the Quality Assurance Program was established in May 1986.
The inspector determined that two different setpoint documents were utilized as sources of the setpoints in the Emergency Procedures (EPs).
Reactor Safety maintained a setpoint document for values associated with the EPGs.
In addition, other setpoints for site specific instrumentation are taken from the site mechanical instrumentation computerized database.
The database is a controlled document, and changes to the database are controlled by the station procedure change process.
The inspector determined that Master File maintained revisions to the document.
The database and changes to the data base are subject to QA audits; however, the inspector noted that QA sign-off was not required for changes.
Setpoint changes to EPs were processed in accordance with the nuclear station modification process. This process requires the performance of a safety evaluation before setpoint changes are made to the EPs.
Reactor Safety reviews these changes and either performs a technical verification of the changes when the setpoint change affects basic assumptions and accident mitigation techniques reflected in the EPGs; or, confirms that the change does not affect the EPGs.
The inspectors reviewed a sample of the plant specific values in the EPs against the values in the setpoint documents and did not identify any discrepancies in the EPs.
i No violations or deviations were identified.
I 11. Quality Assurance (QA) Measures for Emergency Procedures (EPs)
a.
Corporate QA Program for EP Review i
The inspector reviewed audit documentation and conducted interviews
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to determine the corporate quality assurance measures taken to assure
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that the emergency ;nocedures were adequate and met the requirements of the Procedure Generation Package (PGP).
The following documents (
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were reviewed:
Departmental Audit NP-86-20(CN), dated January 7, 1987 Audit Plan for Departmental Audit NP-86-20(CN)
Departmental Audit Procedure Q/A-120, Rev. 20 The co porate QA organization conducts audits in accordance with QA Departmental Audit Procedure QA-210.
Audit NP-86-20(CN) was conducted to fulfill the annual review requirement for 1986.
From the description of the audit process and discussions with QA personnel, the inspector determined that only one aspect of the EP Upgrade Program, i.e.,
the consistency between the EPs and the Writer's Guide, was evaluated during the 1986 audit.
The inspector reviewed the following four EPs to determine if the QA audits of these procedures against the plant specific Writer's Guide represented a comprehensive review of the procedures:
EP/2/A/5000/1B, SI Termination Following Spurious SI
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EP/2/A/5000/203, High Pressurizer Pressure EP/1/A/5000/2F2, Low System Inventory EP/1/A/5000/2E3, High Containment Radiation Level Review of the EPs indicated that overall the EPs did conform to the Writer's Guide, however, the inspector found the following discrepancies. In three of the four :Ps, the cover sheet used in the EPs was more recent than the cover sheet described in the Writer's Guide, indicating that the Writer's Guide had not been updated appropriately.
The inspector also noted that EP/2/A/5000/203, Retype 0, contained an action step, step 7.a RNO, that did not begin with a verb. The step used the word establishing versus. establish.
Also EP/2/A/5000/1B contained ambiguous logic statements in step 10.b, step 20 RNO 3,
1st bullet, and step 31 RNO.
These discrepancies had not been identified in the QA audit. The inspector brought these items to the attention of the licensee for correction.
The QA audit also stated that Real-Time Walk-Through doct mentation for the Unit 2 EPs samoled were " verified to be contained i-the Master File copy of the Emergency and Abnormal procedures.'
The inspector did not have an opportunity to review the Master
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records to determine which documentation was reviewed, however,
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inspectors had determined during this inspection that Real-Time Walk-Throughs were not conducted for Unit 2.
Instead, the licensee took credit for the Unit 1 Real-Time Walk-Throughs as discussed in paragraph 7.b.1 of this report.
The licensee indicated that the audit statement was in error and should have indicated that the corresponding Unit 1 Real-Time Walk-Through documentation had been
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audited for the Unit 2 EPs sampled.
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l From the review of the 1986 corporate QA audit, the inspector was
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concerned that the audit covered only one element of the EP Upgrade j
l Program and did not have provisions to review the program to
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determine if the concerns identified by the NRC in IE Information
'i Notice 86-64, Deficiencies in Upgrade Programs for Emergency Operating Procedures, applied to the facility.
The scope of the audit also did not address review of the commitments in the approved PGP.
The inspector discussed these concerns with the licensee.
The licensee stated that additional corporate level audits of the EP Upgrade Program had been conducted, however, due to limited time at the end of the inspection period, the licensee was not able to l
provide all audits for review.
Further review of corporate audits will be the subject of future inspections.
The inspector did
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emphasize to the licenne that, if not already developed, specific audit criteria addressing the commitments made in the PGP for developing, implementing and maintaining the EPs should be developed since the 1986 audit contained criteria which appeared administrative in nature and did not ensure total ccmpliance with PGP commitments and the requirements of NUREG-0737, Supplement 1, Clarification of TMI Action Plan Requirements - Requirements for Emergency Response Capabilities.
No violations or deviations were identified.
b.
Site Quality Assurance (QA) Program for Review of EPs The site QA organization does " program level" surveillance on a routine basis in accordance with QA Manual Procedure QA-500. Site QA does not conduct a surveillance that concentrates specifically on the EPs, however, EPs may be chosen by an auditor to be reviewed as part of scheduled site procedure reviews.
The inspector reviewed a surveillance report dated May 18, 1986, which included verification that EPs were administered and changed per the guidelines set forth in station administrative procedures; that completed EPs were maintained in thu Master File; and that the current revision of the EPs were on file in Master Files, properly maintained and controlled, and agreed with the control room copy. There is no indication, from
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either the surveillance report, QA documentation for conducting the surveillance, or interviews with site QA personnel, that EP support documents (i.e., Emergency Procedure Guidelines, Writer's Guide, or the validation and verification program documentation) are being reviewed by site QA.
Although content and format of the EPs in accordance with APM-4.2 appears to have been verified, there appears
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to be no indication that criteria specified in the Writer's Guide portion of PGP are reviewed by QA to insure that gui_ dance is being used consistently or at all. Review of additional site QA audits of the emergency procedures will be the subject of future inspection.
I No violations or deviations were identified.
12.
Evaluation of Natural Circulation Cooldown Capabilities (TI 2515/86)
The inspector reviewed documentation provided by the licensee to verify the implementation of programs for the control of natural circulation j
cooldown were in accordance with the basic requirements of IE Temporary i
Instruction 2515/86.
The training of operations personnel regarding natural circulation had
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been previously evaluated by the facility Resident Inspectors through a l
review of training records and was closed out in inspection report l
50-413/84-102 and 50-414/84-47.
The inspector conducted interviews with j
three operators at the facility.
During the interviews, operations
personnel stated that train %ng on EP/1/A/5000/1A1, Natural Circulation Cooldown, had taken place in both the classroom and on the simulator. The completion of the training was also evident during a walk-through of EP/1/A/5000/1A1 in the control room as operators had little difficulty completing the EP.
Based on this information, the area of training conducted on Natural Circulation Cooldown is considered closed.
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13. Acronyms l
APM Administrative Policy Manual l
CNS Catawba Nuclear Station l
EP Emergency Procedures EPG Emergency Procedure Guidelines ERG Emergency Response Guidelines NC Reactor Coolant System ND Residual Heat Removal System NRC Nuclear Regulatory Commission NSSS Nuclear Steam Suppiv System
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OP Operating Procedure l
PGP Procedure Generation Package l
PORV Power Operated Relief Valve
RNO Response Not Obtained R0 Reactor Operator SER Safety Evaluation Report S/G Steam Generator S/GTR Steam Generator Tube O pture S/I Safety Injection SR0 Senior Reactor Operator WOG Westinghouse Owners Group