ML20071L022
| ML20071L022 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/23/1982 |
| From: | Parker W DUKE POWER CO. |
| To: | Deyoung R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| EA-82-065, EA-82-65, NUDOCS 8208020303 | |
| Download: ML20071L022 (40) | |
Text
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f.tc 'd 7-23 DUKE POWER COMPANY Powen Ut;narsu 422 Sor;Tu Cutracia Srpeer, CHAhwTTE, N. C. ana42 WILLI AM O PAR K ER, J R.
Vice PpEUDENT T E t t Pwom c:AaEA 704 Stra Pmoov Tion 373-4383 July 23, 1982 Mr. Richard C. DeYoung Director, Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Re: Oconee Nuclear Station Docket No. 50-269 License No. DPR-38 Notice of Violation and Proposed Imposition of Civil Penalty
Dear Mr. DeYoung:
Duke Power Company (Duke) hereby files its answer, in accordance with 10 CFR 2.205(b), to the " Notice of Violation and Proposed Imposition of Civil Penalty" issued in this docket by the NRC's Region II on June 25, 1982.
As will be discussed in detail below, Duke does not believe that the incident which is the subject of the violation and the proposed civil penalty provides the necessary basis for imposition of a civil penalty.
Therefore, Duke is protesting the imposition of any civil penalty, and requests that, pursuant to 10 CFR 2.205, the NRC issue an order dismissing the proposed civil penalty.
The incident for which the civil penalty is proposed is discussed fully in our response to the Notice of Violation, which is Attachment 1 to this letter.
That incident involved an apparent failure to replace an instrument cap from the test tee for penetration WB-13.
While Duke's instrument calibration l
procedures required replacement of that cap as a final step in returning the l
instrument to service,
the cap apparently was not replaced following instrument calibration.
On March 23, 1982, the NRC's Resident Inspector discovered the missing instrument cap in a routine inspection of Oconee's Unit 1 Reactor Building. It appears that the instrument cap may have been missing from July 9, 1981 until March 23, 1982.
The NRC concluded that containment integrity was violated and the Reactor Building spray initiation system was degraded during certain periods in the July 9,
1p81 to March 23, 1982 interval. More specifically, two violations were alleged:
(1) " Containment integrity of the Unit I reactor building was not maintained for fifty-one days while RCS pressure was greater than 300 psig and temperature was greater than 200 F."
l (2) "For thirty-two days, one of three channels of Train A of reactor building spray initiation for Unit I was inoperable while the reactor was critical."
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As a result, the NRC Staff proposed a civil penalty in the amount of Forty-four Thousand Dollars ($44,000) "[t]o emphasize the need for [ Duke] to ensure that procedures affecting safe operation of the plant are meticulously follovnd. "
Because of the apparent duration of the event the base civil penalty ($40,000) was increased by $4,000.
Duke believes that the proposed civil penalty should be withdrawn.
In Duke's view there is no basis for imposing a penalty for this incident.
The incident from which the alleged violations appear to have arisen occurred when a step in the procedures for returning instruments to service after calibration was not followed.
Prior to the time that the missing instrument cap which is the subject of this incident was discovered, Duke had identified the potential for a problem to occur when instrument calibration work was being conducted.
Therefore, Duke had amended its procedures to require that compliance with those procedures be independently verified (by persons not from the crew doing the work) when instruments are returned to service after being calibrated.
Thus, to the extent that a deficiency in Duke's procedures led to the incident, Duke had discovered that deficiency and had taken effective corrective action to preclude an occurrence of the nature involved prior to discovery of the missing instrument cap.
( Attachment 1, pp.1, 4.)
- Moreover, Duke has recently completed a
comprehensive analysis (Attachment 2 to this letter) of the potential significance of the incident.
That analysis shows that, with respect to the Reactor Building spray systems, all three initiating channels were in fact operable within TechGcal Specification limits even with the instrument cap missing from the instrument tee.
Thus there was no violation of Technical Specification 3.5.1 relating to the Reactor Building spray trains and, of course, there is no basis for a civil penalty in this regard.
Moreover, even with the pressure cap missing, any doses to the public which could have resulted from an accident would be less than the accident doses set out in the FSAR.
Therefore, the incident itself was not of significance when measured against either the actual or potential impact on the health and safety of the public.
Duke would like to emphasize, however, that even though the analyses show that there was no threat to the public health and safety as a result of this incident, the Company both recognizes and appreciates the potential seriousness that might occur from incidents of this nature. Duke believes that its record in identifying and correcting potential problem areas in its operations is a good one, and will continue in the future to make every effort to assure that it continues in that fashion.
The Notice states that the civil penalty was proposed "to emphasize the need for [ Duke] to ensure that procedures affecting safe operation of the plant are meticulously followed.
Duke believes that the actions taken in regard to this incident (both before and after discovery of the missing instrument cap) demonstrate the Company's concern with safe operation of the Oconee plant.
Thus, Duke does not believe that a civil penalty is warranted.
In Duke's view the civil penalty is based upon an incorrect application of the Commission's Policy Statement to the incident and thus it should be withdrawn.
Moreover, other factors also warrant withdrawal of the proposed civil penalty.
These factors include the fact that Duke itself identified as a potential problem in its procedures (prior to discovery of the missing cap) the lack of an independent verificaticn for work of the nature done, and had taken appropriate corrective action.
Though 'the change in procedures came too late to prevent.this incident, it will, of course, prevent' the occurrence of similar incidents in the future.
In addition, Duke's enforcement history at the Oconee
- Station does not warrant _ imposition of a. civil penalty.
Each of these points is discussed in greater detail below:
The incident was of limited technical significance and posed no risk to the public health and safety.
Moreover, Duke's analysis shows that there was no violation of Technical Specification 3.5.1.
First, as the analysis demonstrates conclusively, thcugh there was in fact a technical. violation of Technical Specification'. 3.6.1, there was no actual impact to the public health. and safety from the missing instrument cap.
So far as any potential impact is1 concerned, the calculations performed show that, under design basis accident conditions and the Reactor Building leakage that existed at the time of the last Reactor Building leak rate test, even with the instrument cap missing from the line, the doses resulting from the design basis LOCA would be less than the doses calculated in the FSAR.
Thus, this incident did not involve a significant potential impact to the public health and safety, as there would have been no discernible increase in the potential radiological consequences of postulated accidents.
Second, the analysis demonstrates that the incident did not lead to a violation of Technical Specification ' 3.5.1,-
in that it did not result in
-inoperability of Channel 7 of the Reactor Building. spray system.. To the contrary, tests recently completed, described in Attachment 2, show that even with the. cap missing, the affected channel would have operated within' the Technical Specification limits if called upon.
2.
In light of this analysis, Duke believes that the proposed civil penalty was based on an improper application of the Commission's Policy Statement, in that the violation should be a-Level IV. violation rather than a Level III violation and thus a civil penalty is not warranted.
Application of the guidance set forth in the Commission's Policy Statement to ' the specific facts. resulting from the analysis in Attachment 2 demonstrates this.
The fundamental criterion that the Commission intends be used in placing a violation in a' particular severity level is "the actual or potential impact on the health and safety of the public." 47 F.R. 9968.
For Reactor Operations, Severity Level III violations include one in which Technical Specification Limiting Controls for Operation are exceeded, which results in loss of a safety function, or one leading to a situation in which a system designed to prevent or ' mitigate a serious safety event is not able to : perform its intended funcation.
10 CFR Part 2, Appendix C, Supplement I C.
Severity Level IV violations include those that consist of a failure to meet regulatory requirements that have more than minor safety or environmental significance; '
that is, violations of requirements which, if left uncorrected, could lead to a more serious concern.
10 CFR Part 2, Appendix C, Section III; Supplement I D.
Clearly, under application of these criteria, this incident should not result in a Level III violation.
As noted, one of the two violations cited by the NRC resulting from this -incident (relating to the Reactor Building spray -
_ system) turns out, upon analysis, not to be a violation.
With respect to the violation relating to containment integrity, though there may have in fact been a technical violation of Technical Specification 3.6.1, in Duke's view this must be weighed against the fact that, as shown in Attachment 2, such violation posed no threat to the public health and safety.
Therefore, the incident did not involve exceeding a Technical Specification limit which resulted in loss of a safety function, nor did it involve a situation in which a system designed to mitigate consequences of or prevent a serious safety event was prevented from performing its intended function.
Indeed, Duke believes that the incident which led to the violations set forth in the Commission's June 25 Notice is a result of a problem in its procedures which Duke had identified in January of 1982.
That is, Duke determined that it was necessary to provide independent verification (by persons not from the crew doing the work) that an instrument is properly returned to service after being calibrated.
As Attachment 1 explains, upon identifying that problem, Duke changed its procedures to provide for such independent verification.
This change in procedures will preclude any such occurrence in the future.
Nevertheless, when the missing instrument cap was discovered in March, Duke reemphasized its change in procedures.
Thus, in Duke's view, the problem with the procedures which led to the incident is no more than a matter which has a "more than minor" safety significance.
That is, if Duke had not found the potential problem with its procedures, it could have led to a more serious concern.
Consequently, Duke believes that the incident should be classified as a Level IV violation, and no civil penalty should be imposed.
Duke would like to add one thought with respect to the violation relating to containment integrity.
It may be that the NRC has as a matter of policy determined that any breach of containment, regardless of whether it constitutes a potential impact to the public health and safety, is a Level III violation and warrants a civil penalty.
If that is the case, the facts of this matter clearly argue that the policy should sbe changed. Moreover, if the NRC has made such a determination, it should announce that policy to the industry and provide the technical basis for its judgment.
3.
The proposed civil penalty should be withdrawn.
The Commission's procedures provide that licensees, upon notification of a proposed civil penalty, will have the opportunity to raise mitigating circumstances unique to their particular cases.
These circumstances will be taken into account when the decision is made whether or not to order the imposition of a civil penalty for a particular incident.
Licensees have been assured by the Commission that
"[m]itigation or remission of civil penalties based on such..
responses is not uncommon when compelling arguments are presented." 47 F.R. 9988.
In Duke's view, complete remission of the proposed civil penalty is warranted in this case.
The primary reasons are, of course, the fact that the single violation involving the missing cap constituted no risk to the public health and safety and that Duke itself identified the potential problem with its procedures which could lead to incidents such as the missing instrument cap and took corrective action to preclude such incidents in the future.
However, in accordance with the Notice, and the Commission's Policy Statement, Duke hereby addresses the factors set out in Section IV B of 10 CFR Part 2, Appendix C.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _, Nr (i)
Prompt Identification and Reporting As noted above, it is Duke's view that the incident involving the missing instrt. ment cap which led to the violation resulted from the fact that the instrument calibration procedures did not provide for an independent verification that the instrument had been properly returned to service following calibration.
As discussed more fully in Attachment 1, in January of 1982 Duke identified that potential problem and promptly took appropriate corrective action. This potential problem was not reportable to the NRC under applicable requirements.
In light of the fact that the potential problem was identified by Duke, the proposed civil penalty should be mitigated.
(ii)
Corrective Action to Prevent Recurrence When Duke discovered the potential problem with its procedures, prompt corrective action was taken.
The corrective action is discussed fully in, but, briefly, Duke changed its procedures in January to require an independent verification (by persons not from the crew doing the work) that equipment is returned to normal after being taken out of service.
If the equipment is independently determined to be functional it must be documented by the individual performing the verification.
Thus, prompt corrective action was taken which will preclude occurrence of similar incidents in the future and therefore the proposed civil penalty should be mitigated.
(iii) Enforcement History Duke believes that its enforcement history at Oconee is a factor which should weigh in favor of mitigation of the proposed civil penalty.
In 1977, Duke was assessed a civil penalty for an incident at Oconee.
A review of the enforcement history since that time indicates that though violations have occurred, none has been of a seriousness sufficient to warrant imposition of a civil penalty.
Moreover, those violations are isolated in nature and not recurring.
Becat se of this, it can be inferred that Duke's corrective actions in response to these violations are effective, and that there is no serious programmatic deficiency.
Thus, this factor should mitigate the proposed civil penalty.
(iv)
Prior Notice of Similar Events Duke, upon learning of potential inadequacies in its procedures, promptly instituted corrective actions (see (ii) above) and thus this factor is not applicable.
(v)
Multiple Occurrences This factor does not appear to be applicable.
There have been no multiple examples of this incident.
As noted above, in Duke's view the subject incident at Oconee does not warrant the imposition of any civil penalty.
However, Duke wishes to comment on what it believes is a fundamental unfairness in the way in which the Commission's enforcement procedures, as set forth in its Policy Statement on Enforcement, are carried out.
Duke makes these comments in a constructive vein, recognizing that the Commission is monitoring activities under its Policy
- Statement on Enforcement, and has indicated its willingness to change the way it, and its Staff, carries out activities under the statement if circumstances warrant. 47 F.R. 9989.
To describe the process is to illustrate the problem.
Put briefly, when a violation exists which might lead to an enforcement action, it is not uncommon for an enforcement conference to be held.
At such a conference, the licensee discusses the incident, including the sequence of events which led to it, the actions taken to mitigate the event, and the corrective actions which have been and will be taken to prevent recurrence of similar events in the future.
Any questions which the NRC Staff might have are discussed.
In short, the enforcement conference serves as a useful forum for an exchange of information and viewpoints between the licensee and the NRC Staff relating to a specific incident which might lead to a notice of violation and proposed imposition of a civil penalty.
Following the enforcement conference, the NRC Staff then considers what further action to take.
If in the Staff's view a violation exists, it will issue a Notice of Violation and, if it believes it is warranted, a Notice of Proposed Civil Penalty.
Simultaneously with its issuance of the Notice of Proposed Civil Penalty, the NRC Staff issues a press release announcing the violation and the Notice of Proposed Civil Penalty.
A press release is issued only if notice is given of a proposed civil penalty.
The Commission views the publicity attending a civil penalty as an important part of its enforcement procedures.
10 CFR Part 2, Appendix C, SIV. 47 F.R. 9990.
The Commission's procedures provide a mechanism for a licensee to demonstrate why it believes a civil penalty is not warranted, or why one of a lesser amount is justified.
See, e.g., 47 F.R. 9988; 10 CFR Part 2, Appendix C, SIV B 1, 2.
Ilowever, only after the Notice of Proposed Civil Penalty and the press release have been issued is a licensee permitted to make its case in that regard.
Thus, to the extent factors exist which would eliminate or mitigate the proposed civil pen. ilty,
licensees are foreclosed from the opportunity of being heard before the NRC takes action and the public documents and press release which the NRC issues cannot reflect licensee's position.
This problem is particularly acute when dealing with Level III violations, where a significant amount of judgment is involved and a violation does not always result in a notice of proposed civil penalty and attendant press release.
In Duke's view, as a matter of basic fairness, licensees should be given an opportunity to respond (as provided in the Policy Statement) to a proposed civil penalty, putting forth the factors which it believes justify cancellation or mitigation before public announcement of such a penalty is made.
In light of the fact that the Commission views the press release mechanism as a punitive measure, it should take particular pains to ensure that, at least in the case of Level III violations, no public announcements are made of civil penalties until it has considered carefully all factors involved and such a penalty is actually assessed.
The instant matter perfectly illustrates the problems inherent in this process.
An enforcement conference was held on May 21, 1982.
At that time, the subject incident was discussed.
That discussion included the fact that Duke had identified the potential problem with its procedures and had taken
7_
appropriate corrective action.
The incident of the missing instrument cap was also discussed in detail.
Thus, following that conference, both the NRC Staff and Duke had an understanding of the facts surrounding that incident.
Specifically, a full discussion was held on the potential problem with Duke's procedures, the fact that Duke had identified that problem, the corrective action taken, and the missing cap itself.
However, the analyses presented in had not been prepared and were not discussed. The question of a civil penalty was not discussed, beyond an indication by the NRC Staff that the incident appeared to them to be a Level III violation which might warrant a civil penalty.
On June 25, the NRC Staff issued its " Notice of Violation and Proposed Imposition of Civil Penalty" and press release concerning the incident.
As a result of that issuance, a substantial amount of publicity resulted, both local and national.
However, because of the nature of the process, the NRC had not--and could not have--considered any of the factors which Duke has raised in this letter and its attachments which warrant withdrawal of the proposed civil penalty.
And it naturally follows that, because of the timing of the public announcement by the NRC, none of these factors could be reflected therein, thus neither the notice nor the press release reflected in any way Duke's position on the matter.
Of particular concern is the fact that, based on the timing of the NRC's release, it could not clearly state that the incident had no potential to affect adversely the public health and safety and therefore could not put this incident in its proper perspective.
Therefore, in light of all these circumstances, Duke is compelled to protest this procedure.
Duke believes that it has been made to suffer unneeded and unjustified harm through advarse publicity caused by the premature issuance of the Commission's press release, when the facts clearly demonstrate no potential danger to the public health and safety existed and that no civil penalty is warranted.
Duke would like to make a second comment. The NRC, before imposing a civil penalty for an incident at a facility, should take into account the enforcement history of the licensee at that facility.
The Commission's Policy Statement indicates that the enforcement history will be considered, but only in the context of determining whether the base civil penalty to be assessed for an incident should be increased.
10 CFR Part 2, Appendix C, Section IV B 3.
Duke believes that, when a licensee's enforcement history warrants, credit should be given (in combination with other factors set out in Section IV B of Appendix C) in considering whether a civil penalty should be either mitigated or imposed at all.
In light of the above discussion, and the analyses and explanation set forth in Attachments 1 and 2, Duke does not believe that imposition of any civil penalty is warranted in this matter.
Therefore, Duke requests that the NRC issue an order which withdraws the civil penalty proposed in Mr. O'Reilley's letter of June 25, 1982, and further requests that the order be accompanied by a press release which announces that the NRC is withdrawing its proposed civil penalty and explains the reasons for its withdrawal.
Ve y truly yours, o_. - _ tj.
William O. Parker, Jr.
I, William O.
Parker,-Jr., hereby affirm that I have read the foregoing document and that it is true and correct to the best of my knowledge and belief.
_ d.
/
O m.
c-Mr. Richard C.'DeYoung July 23, 1982 Page 9 cc:
Mr. James P. O'Reilly, Regional Administrator.
U. S. Nuclear Regulatory Commission Region II
.101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Mr. W. T. Orders NRC Resident Inspector Oconee Nuclear Station Mr. Philip C. Wagner Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 L_
ATTACEMENT 1 RESPONSE TO NOTICE OF VIOLATION On July 9, 1981 pressure switch IPS22, one of the three providing input to Channel 7 of Engineered Safeguards which initiates the Reactor Building Spray System was calibrated in accordance with Technical Specification 4.1.
Unit 1 was in.the refueling mode and this test was scheduled for. completion during the outage. The instrument was required to be returned to-service following calibra-tion by the following Step 10.2.10 of Procedure IP/0/A/310/5D.
10.2.10 Decrease pressure to zero; disconnect pneumatic calibrator; replace cap on tee; open isolation valve and return switch to service.
(See Attachment 4.)
The procedure also required redundant verification and usually involved two persons from the same crew.
Apparently, the test cap on the k inch calibration line was left off and was not discovered until March 23, 1982.
The missing cap in the Penetration Room could have allowed flow from the Reactor Building into the Penetration Room.
This constituted a violation of Technical Specification 3.6.1.
(See the response to the Notice of Violation contained in this attachment.)
On January 1 and 2, 1982, two unit trips occurred as a result of certain secondary side non-safety related instrumentation being inoperable because their isolation valves were closed.
The valves were not returned to the normal position after testing, as a result of personnel error, which identified a need for a more reliable method of verification of return to normal after work is completed.
The instrument procedure, at the time, _ required redundant verification of restora-tion of the tested instrument to normal.
This involved two persons from the same crew who performed the calibration and verified that.the instruments were properly returned to normal.
On January 15, 1982, the Oconee I6E Engineer issued.a letter (Attachment 3) requiring independent verification, in an effort to ensure that for all future tests, equipment would be returned to normal and proper documentation would exist. This ind nendent verification requires that an individual unconnected with the crew perform the testing and calibration confirm that all the required procedural steps are complete and that the system is properly returned to an operable condition.
On March 23, 1982, the NRC Resident Inspector discovered during an inspection t.our that a test tee cap to the instrument line for pressure switch 1PS22 was missing.
A cap.was.immediately installed and all units were checked for similar situations.
No other abnormalities were found. A verbal report was provided to the NRC that day and Licensee Event Report number R0-269/82-08 was submitted on April 6, 1982 (revised July 23, 1982). On March 26, the Oconee Station Manager issued a letter emphasizing the need for an improved return-to-normal system that would include signing off each step of the test procedure required to return the instrument to normal.
Independent verification of these steps is also to be conducted in accor-dance with the procedural-changes accomplished in January. Refer to Attachment 5 for a revised version of IP/0/A/310/5D.
On May 21, 1982, Duke met with the NRC and discussed the incident and the improve-bi ments in the verification system initiated prior to and following the incident.
It was emphasized that procedural changes had been initiated prior to the incident discovery in the area of independent verification.
In addition, Duke stated that in its view, the health and safety of the public were not threatened. A detailed evaluation of the potential impact on the health and safety of the public is con-tained in Attachment 2 of this submittal.
The Notice of Violation states the reason for imposing this civil penalty on Duke Power Company is "to emphasize the need for the licensee to ensure that procedures affecting safe operation of the plant are meticulously followed..."
Duke considers that the actions taken prior to and following this incident demon-strate that the company is vitally conce ned with the safe operation of its facility; thus, the proposed action by the NRC is not warranted to accomplish this stated goal.
Following is the response to the Notice of Violation.
Violation Technical Specification 3.6.1 requires that containment integrity be maintained whenever reactor coolant system (RCS) gressure is greater than 300 psig and temperature is greater than 200 F.
Technical Specification 3.5.1 requires that all three channels of both trains of reactor building spray initiation be operable when the reactor is critical.
Technical Specification 6.4.1 requires that the plant be maintained in accordance with approved procedures. Procedure IP/0/A/310/5D was established and approved to implement 6.4.1.
Step 10.2.10 of the procedure requires replacement of the cap on the k inch calibration line connected to the inch sensing line for reactor building pressure switch 1PS-22.
Contrary to the above, on July 9, 1981, the licensee failed to follow step 10.2.10 of procedure IP/0/A/310/5D. As a result of the failure the following conditions existed between July 9, 1981 and March 23, 1982.
1.
Containment integrity of the Unit 1 reactor building was not maintained for fifty-one days while RCS pressure was greater than 300 psig and temperature was greater than 200 F.
2.
For thirty-two days, one of three channels of Train A of reactor building spray initiation for Unit 1 was inoperable while the reactor was critical.
This is a Severity Level III violation (Supplement I).
(Civil Penalty - $44,000)
Response
1.
Admission or denial of the alleged violation:
Duke Power admits the violation outlined in~ Item 1; however, the company does not. agree with the severity level assigned to the violation ncr the imposition of a civil penalty. Technical bases justifying a lesser severity level and a withdrawal of the proposed civil penalty are included in Attach-ment 2.
Duke Power denies the' violation contained in Item 2.
Technical bases for denial are also included in Attachment 2.
2.
Reasons for the violation:
I&E tcchnicians performing the instrument calibration apparently failed to-replace the calibration tee cap following completion of the calibration procedure.
Prior to January 15, 1982, there was not a procedure which required inde-pendent verification (by persons not from the crew doing the work) of return to service of equipment by station 16E personnel.
3.
Corrective actions taken and results:
The immediate corrective action was to replace the missing cap for the test tee and to check all units for similar situatione. No other abnormalities were found.
All safety related Instrument Procedures have been revised to incorporate a section for specifically identifying all equipment removed from service as well as an appropriate independent verification of the equipment restora-tion. The independent verification will require a second person (other than those doing the work) to verify the specific items are restored to service or verify operability by diverse means (Control Room indications, string check, etc.).
4.
Corrective actions to be taken to avoid further violations:
All corrective actions have been taken as discussed in the preceding para-graphs.
5.
Date when full compliance will be achieved:
Full compliance has been achieved.
ATTACHMENT 2-TECHNICAL SIGNIFICANCE OF THE INCIDENT The safety significance of an operation occurrence can be assessed in terms of (1) actual impact on the health and safety of the public as a result of the incident.and (2) potential impact on the health and safety of the public via a significant degradation in the plant's design basis safety level.
Since this incident did not involve any release in radioactive material, there was no actual impact on the health and safety of the public.
EVALUATION OF POTENTIAL DIPACT ON.THE HEALTH AND SAFETY OF THE PUBLIC An operational occurrence might constitute an increased risk to the' health and safety of the public if the incident resulted in a significant reduction in the plant's design basis safety level. To determine whether a particular operational occurrence' constituted a significant reduction in the plant's design basis safety level, the following questions must be answered:
(a) Did the event involve a significant adverse impact on the radiological consequence of design basis accidents?
(b) Did the event result in a significant degradation of safety systems designed to mitigate accidents?
(c) Did the event result in a significant increase in the like-lihood of accidents?
The following evaluation addresses each of these questions with respect to the specific incident under consideration.
Evaluation of Radiological Impact The reactor building is the major barrier designed to contain radioactivity following accidents involving core damage and release into the containment. The design basis criterion is to limit the RB leakage to within 0.25 weight percent /
day under the conditions of the maximum hypothetical accident as described in Section 14 of the FSAR.
During this incident a leakage path existed from the RB'to the penetration room through the h" RB pressure sensing line and out through the \\" calibration tube with an effective flow path I.D. of 0.19".
Assuming the RB is at its design pressure of 59 psig continuously, the resulting leakage into the penetration room would have been 0.36 weight percent / day.
How-ever, such an RB pressure behavior is rather impossible to be manifested during the design basis accidents. Using a conservative envelope of the RB pressure response for the limiting LOCA (59 psig for the first hour and 5 psig for the subsequent 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />) as calculated in the FSAR (Figure 14-631), a leakage rate of.093 weight percent / day is calculated to occur.
This is considerably smaller than the RB leakage limit of 0.25 weight percent / day. Furthermore, the most recent (February 11, 1980) integrated leak rate test of the RB indicated a leakage rate of 0.168 weight percent per day (normalized value at 59 psig from the test pressure of 29.5 psig) with adequate margin to accomodate some unanticipated leakage.
The most significant aspect of this incident is the potential impact of the leak path on the offsite accident doses. The penetration room is designed to filter cs out the most significant species of radioisotopes (iodine and particulates) f rom l
i..
O the leakage out of the RB by passing through several filter assemblies prior to discharge into the unit vent.
In order to assess the impact of this incident on the radiological consequences of potential accidents, it is necessary to examine the contribution of the RB leakage through the penetration room to the offsite doses'following the design basis accidents.
The FSAR offsite dose analysis assumed 50 percent of the RB design basis leakage to pass through the penetration room and the remaining 50 percent to bypass the penetration room.
Further, the RB leakage through the penetration room results in an elevated release through the unit vent.
Taking these features into account, the impact of this incident on the radiological consequences of the design basis LOCA--the limiting mechanistic accident in the RB--can be assessed as follows:
(a) Using the FSAR assumption of 0.25 percent RB leakage and considering the additional leakage of 0.36 percent during the first hour and 0.08 percent after the first hour, the 2-hour thyroid dose at the exclusion distance in the event of a design basis LOCA would have been 5 percent higher than the FSAR value.
Since the FSAR value is only 1.5 percent of the 10 CFR 100 limit, this increase represents only a 0.077 percent of the 10 CFR 100 limit.
(b) Using the realistic value of the RB leakage as indicated by the RB integrated leak rate test and considering the additional leakage of 0.36 during the first hour and 0.08 percent after the first hour, the 2-hour thyroid dose at the exclusion distance in the event of a design basis LOCA would have been less than the value calculated in the FSAR.
(c) Using the FSAR assumption of 0.25 percent leakage _and considering the additional leakage of 0.36 percent during the first hour and 0.08 percent after the first hour, the 2-hour whole body dose at t'he exclusion distance in the event of a design basis LOCA would have been 40 percent higher than the FSAR value.
Since the FSAR value for the 2-hour LOCA whole body dose is only 0.04 percent of the 10 CFR 100 limit, this increase represents only a 0.016 percent of the 10 CFR 100 limit.
(d) Using the realistic value of the RB leakage as indicated by the RB integrated leak rate test and considering the additional leakage of 0.36 during the first hour and 0.08 percent after the first hour, the 2-hour whole body dose at the exclusion distance in the event of a design basis LOCA would have been less than the value calculated in the FSAR.
Therefore, this incident did not involve a significant impact on the radiological consequences of potential accidents, if they occurred.
Evaluation of the impact on Safety Systems The subject incident affected the performance of one of three independent channels of one of the two redundant RB spray trains. It did not affect the performance or operation of any other instrumentation or safety system. The RB spray system is a safety system designed to limit the RB pressure to within the design limit during accidents involving blowdown of mass and energy into the RB (LOCA and accidents involving secondary system breakawithin the RB).
It is also considered to be use-ful in scrubbing the radioactivity from the RB atmosphere during the post-accident
E -
i phase of-a LOCA. With respect to limiting the RB pressure, the RB cooling system, which is independent of the spray system, is fully capable of limiting the RB 1
pressure to within the design limit, irrespective of the availability of the spray system.
Each of the two redundant RB spray trains is automatically actuated upon-tripping i
of two of three redundant channels.
The nominal trip setpoint for each channel is 10 psig with a required trip actuation (as required by Tech Spec and safety analysis assumptions) of 30 psig.
The subject incident caused Pressure Switch 1PS-22 not to trip at the nominal setpoint of 10 psig. However, it has now been determined that the channel would have indeed tripped at a pressure of approximately 22 psig, well within the required setpoint of 30 psig. This determination is based on recent tests conducted on the same channel of a similar unit simulating the as-found-condition and applying a gradually increasing pressure on the sensing line. During this test the channel repeatedly tripped at approximately 22'psig;-
therefore, this incident did not degrade the-RB spray system nor did it signifi-cantly degrade the function of the channel.
It is also pointed out that even if the channel were inoperable, the RB spray system can still accommodate a single failure and achieve system design function.
Evaluation of Impact on the Likelihood of Accidents This incident did not, in any manner, affect the initiation of or occurrence of any accident.
CONCLUSION The foregoing evaluation has demonstrated that this incident did not have any actual impact on the health and safety of the_public.
It also did not have any significant potential impact on the health and safety of the public by virtue.
of the fact that (1) it did not result in any increase in the likelihood of accidents, (2) it did not result'in any significant degradation of safety systems designed to mitigate accidents, and (3) it did not involve a significant adverse impact on the potential radiological consequences of design basis accidents.
+
u
January 13, 1982 INTRASTATION LETTER OCONEE NUCLEAR STATION To:
I&E Coordinators & Supervisors SU3 JECT:
Oconee Nuclear Station Independent Verification As you are aware, we experienced two (2) instances of instrumentation being valved out during start-up after Unit 1 refueling outage.
We experienced similar problems after Unit 3 refueling outage.
The pro-blems on Unit 1 caused three (3) unit trips.
While these instances represent only a minute fraction of the work performed they do indicate a trend and we feel we should make adjustments to our practices in an attemp t to reduce this type of incident.
In the future we will require an independent (different from person (s) doing work) verification that equipment is returned to normal if it can not be confirmed by diverse means.
(Control room indication, string checks, etc.)
This will require independent verification of all instru-mentation connected to system piping when the system is not in service and in some cases when the system is in service.
If equipment is deter-mined to be functional by diverse means it shall be so documented by the individual (s) performing the work.
Please insure that this requirement is presented to and understood by all of your e=ployees.
We realize that these changes will be costly in =an-power but believe them necessary to insure proper equipment status for safe efficient operation of the plant.
We appreciate the effort and performance of everyone in the section and hope that these changes will enable us to do an even better job in the future.
D? " MA j/R.-C. Adams I&E. Engineer RCA/pc cc:
J. E. Smith Gerald Vaughn Joe Davis
0 x,..
MASTER FILE P _ uD-1002-1 DUKE POWER COMPANY (1)
ID No:IP/0/A/310/5D PROCEDURE PREPARATION Change (s) 3 to PROCESS RECORD 1
Incorporated (2)
STATION:
Oconee E
- "E"#
I" **
8 (3)
PROCEDURE TITLE:
R.B. Pressure Switch Calibration and Pressure Switch Contact Buffer Tests m-DATE:
[ ~/h d/
(4)
PREPARED BY:
f [ DATE: f - / [" [ /
(5)
REVIEWED BY:
i fj Cross-Disciplinary Review By:
N/R:
(6)
TEMPORARY APPROVAL (IF NECESSARY):
By:
(SRO)
Date:
By:
Date:
(7)
APPROVED BY:
~7h ham Date:
((f /f?/
(8)
MISCELLtNEOUS :
Revt.i d/%. :. l Byff*6MY' h,
// Date: f/b pfM R
l Reviewed / Approved By:
Date:
l 1
I b
FORM SPD-1001-2 DUKE POWER COMPANY NUCLEAR SAFETY EVALUATION CHECK LIST Oconee (1) STATION:
UNIT: 1 2
3 OTHER:
(2) CHECK LIST APPLICABLE TO:
IP/0/A/310/5D (3) SAFETY EVALUATION - PART A The item to which this evaluation is applicable represents:
Yes No X
A change to the station or procedures as described in the FSAR or a test or experiment not described in the FSAR?
If the answer to the above is "Yes", attach a detailed description of the i tem being evaluated and an identification of the affected section(s) of the FSt,R.
(4) SAFETY EVALUATION - PART B Yes No I
Will this item require a change to the station Technical Specifications?
If the answer to the above is 'Tes," ' identify the specification (s) affected and/or attach the applicable pages(s) with the change (s) indicated.
(5) SAFETY EVALUATION - PART C As a result of the item to which this evaluation is applicable:
Yes No I
Will the probability of an accidene previously evaluated in the FSAR be increased?
Yes No I
Will the consequences of an accident previously evaluated in the FSAR be increased?
Yes No I
May the possibility of an accident which is different chan any already evaluated in the FSAR be created?
Yes No I
Will the probability of a malfunction of equipment to safety previously evaluated in the FSAR important be increased?
Yes No I
Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
Yes No, I May the possibility of malfunction of equipment important to safety different than any already evaluated in the FSAR be created?
Yes No I
Will the margin of safety as defined in the bases to any Technical Specification be reduced?
If the answer to any of the preceding is "Yes", an unreviewed safety.
question is involved.
Justify the conclusion that an unreviewed safety question is or is not inv ved.
Attach additional pages as necessary.
(6) PREPARED BY:
-/
JTE:
[ ~ l 9 O/
p DATE:
["/9- ((
(7) REVIEWED BY:
(8) Page 1 of
~
m e
Form SPD-1002-1 4
DUKE POWER COMPANY (1)
ID No: IP/0/A/310/5D PROCEDURE PREPARATION Change (s) 3 to PROCESS RECORD 3
Incorporated (2)
STATION:
Oconee (3)
PROCEDURE TITLE: Engineered Safeguards System Analog Channel C R.B. Pressure Switch Calibration and Pressure Switch (4)
PREPARED BY:
DATE:
3
/3-8 /
(5)
REVIEWED BY:
x DATE: M/9 O r
Cross-Disciplinary Review By:
N/R:
5 (6)
TEMPORARY APPROVAL (IF NECESSARY):
By:
(SRO)
Date:
/,
By:
Date:
'//
!, k (7)
APPROVED BY:
.4-M M f-(
Date:
(8)
MISCELLANEOUS:
d/Appre ::d By3///f_ '//JIp-/] [ Date: _ T,*[2 4,/8/
Revi Reviewed / Approved By:
Date:
n.
e
--e,.----
- ~ -. - -, -. - - - -.. -, - -
DUKE 70WER COMPMPt (1) It No. IP/0/A/310/5D COMPL.UED PROCEDURE
~~
PROCESS RECORD (2)
STATION:
oconee Engineered Safeguards Syste::1 Analog Channel Cl R.B. Pressure (3)
?RCCIDU?2 U TLE.
Switch Calibration and Pressure Switch Contact Buffer Tests (4)
DATI(S) PERFORMED:
(5)
?ROCZDURE COMPLt uCN VERa.-LCATION:
TIS N/A Check lists and/or blanks p;cperly initialed, signed, dated or filled in N/A or N/R, as app cpriace?
TES N/A Listed enclosures attached?
TIS N/A Data sheets at: ached, cc=pleted, dated and sig=ed?
TIS N/A Charts, graphs, ecc. attached and properly dated, identified and marked?
TIS.
N/A Acceptance criteria =e:?
Verified By:
DATI:
(6)
PRCCIDU?2 Cw % 'IION AF? ROVED:
DATI:
(7) 7 WARKS :
(8) Page 1 ef
IP/0/A/310/5D DUKE POWER COMPANY OCONEE NUCLEAR STATION ENGINEERED SAFEGUARDS SYSTEM ANALOG CHANNEL C R.B. PRESSURE SWITCH CALIBRATION AND PRESSURE SWITCH CONTACT BUFFER TESTS 1.0 Purpose 1.1 To furnish a procedure for calibration of Reactor Building pressure switch instrumentation.
1.2 To calibrate digital computer inputs.
1.3 To perform the functional and operational tests.
2.0 References (Use Current Copy) 2.1 Vol. 1 & 3 BMC. NI & RPS & ESS inst.
2.2 881 System Checkout Procedure 2.3 FSAR, Section 7.1.3.
2.4 Technical Specifications, Section 4.1.1.
3.0 Test Equipment Required 3.1 Pneumatic Calibrator, W/T series FA 145 Range 0-30 psi, or equivalent.
4.0 Prerequisites Sign-off(s) on Enclosure 11.1 4.1 Supervisor has reviewed and initialed all portions of this procedure which are not applicable to the activity being performed.
The supervisor's review is not required if the procedure specifies sections to be omitted except as needed r;
for abnormal conditions.
4.2 Verify all changes on the control copy are incorporated on the working copy.
4.3 Computer in operation.
4.4 Verify that.other ES Channels are not in trip conditions.
5.0-Limits and Precautions Use proper precautions while working with components that have HIGH VOLTAGE, HIGH PRESSURE or HIGH TEMPERATURE present.
6.0 Unit Status Sign-off(s) on Enclosure 11.1 6.1 N/A 7.0 General Description This instrumentation is used to sense high building pressure which actuates the ES Building Spray System.
8.0 Major Components Component Description Reference BS4-PS5 (PS-22)
Mercoid Snap Switch Control Bailey Ref.
BS4-PS6 (PS-23)
Mercoid Snap Switch Control Bailey Ref.
9.0 Equipment Specifications 9.1 Operating Range Instrumentation Designation Input Output BS4-PS5 (PS-22) 10 psig Close BS4-PS6 (PS-23) 10 psig Close 9.2 Computer Inputs From BS4-PS5 (PS-22)
D1957 BS4-PS6 (PS-23)
D1958 10.0~ Procedure CAUTION:
If any component calibration is out of tolerance by 2%,
proceed to Maximum Tolerance Exceeded Sheet.
10.1 Pressure Switch ~ Contact Buffer Tests In this case there is a pressure switch (N/0 Contact) applied to cach relay circuit of the contact buffer.
Therefore, in normal operation both lamps will be "0N".
h
I 10.1.1 Depress S1 (Switch at top).
DS1 should go "0FF" (lamp at top).
10.1.2 While holding S1, check auxiliary relay for proper indication.(DS1 and DS2 on auxiliary relay will be bright).
The building spray trip lamp on the analog indicating panel should be bright.
10.1.3 Release SI.
All of the lamps in steps 1 and 2 should return to their initial states.
Check computer input D-1957 for proper indication.
Also check for proper statalarm indication.
10.1.4 Depress S2 (Switch at bottom).
DS2 should go "0FF".
10.1.5 While holding S2, check auxiliary relay for proper indication (DS3 and DS4 on auxiliary relay will be bright).
The building spray trip lamp should once j
again be bright.
10.1.6 Release S2.
All of the lamps mentioned in Steps 4 and 5 should return to their initial states.
Check computer input D-1958 for proper indication.
Also check.for proper statalarm indication.
10.2 Pressure Switch Setting l
Complete Enclosures 11.2a and 11.2b using the following procedure to calibrate:
10.2.1 Close isolation valve and connect Pneumatic Calibrator to Tee.
10.2.2 Slowly increase pressure until respective contact buffer light goes "out".
Record pressure on data sheet under "As Found".
10.2.3 If switch makes within the pressure tolerance specified, decrease pressure to zero, disconnect pneumatic calibrator, put cap on Tee, open valve, and return switch to service.
10.2.4 If setting is in error or needs to be changed, continue with Step 5.
10.2.5 Remove cover from switch to expose the switch adjust-l ment.
l 10.2.6 Set input pressure to that value listed on data sheet at which you want the switch to make.
10.2.7 Adjust switch setpoint until switch makes.
10.2.8 If switch makes on pressure increase, reduce the pressure until switch resets.
Slowly increase pressure until switch just makes.
If it makes at desired setpoint, record data on data sheet under "AS LEFT".
If still not within tolerance specified, repeat Steps 5, 6, and 7, until it is within tolerance specified.
10.2.9 Replace cover on switch and recheck calibration to insure switch is still within tolerance.
10.2.10 Decrease pressure to zero; disconnect pneumatic calibrator; replace cap on tee; open isolation valve and return switch to service.
10.2.11 Insure that Building Pressure Analog Channel C is reset and not in the tripped state.
11.0 Enclosures 11.1 Sign-Off Sheet 11.2 Calibration Data Sheets 11.2.a PS-22.( BS4-PS5) 11.2.b PS-23 (BS4-PS 6) l 11.3 Tolerance Limit Exceeded Sheet l
t e
4 f
ENCLOSURE 11.1 Sign-off(s)
Unit Status Date Began Prerequisites 6.1 Date Completed 4.1 WR#
4.2 Unit 4.3 4.4 Data Sheets 11.2.a 11.2.b PERFORMED BY REMARKS 9
o O
l I
l
DUKE POWER COMPANY OCONEE NUCLEAR STATION UNIT #
ENCLOSURE 11.2.a CALIBRATION DATA SHEET IP/0/A/310/5D 4
' ITEM Pressure Switch TEST EQUIPMENT USED MFG.
Mercoid ITEM SN TYPE APW-7041-153 Satpoint Tolerance t.5 % of Span t.095 PSIG SYSTEM ES Building Spray INSTRUMENT NO.
PS-22 (BS4-PSS)
PRODUCT INSTRUCTION SPAN 19 PSIG Switch to close on pressure increase.
SWITCH LOC.
Input Desired As As Left Error
)
to Actuation Found Setpoint PSI Switch Point (PSIG)
PSI PSI PSI 10 10 All isolation valves left open - Verified By MAILWM ERROR IN PSI 4
PERFORMED BY DATE i
.__ --.,_ _ ~ -
r DUKE POWER COMPANY OCONEE NUCLEAR STATION UNIT #
ENCLOSURE 11.2.b CALIBRATION DATA SHEET IP/0/A/310/5D ITEM Pressure Switch TEST EQUIPMENT USED MFG.
Mercoid ITEM SN TYPE APW-7041-153 Satpoint Tolerance
.5 % of Span
.095 PSIG SYSTEM ES Building-Spray INSTRUMENT NO.
PS-23 (BS4-PS6)
PRODUCT INSTRUCTION SPAN 19 PSIG Switch to close on pressure increase.
I SUITCH LOC.
i Input Desired As As Left Error to Actuation Found Setpoint' PSI Switch Point (PSIG)
PSI PSI PSIG 10 10 All icolation valves left open - Verified By MAXIMUM ERROR IN PSI PERFORMED BY DATE L
ENCLOSURE 11'.3 IP/0/A/310/5D MAiCIMUM TOLERANCE LIMIT EXCEEDED SHEET Initial & Date (A).
Notified Instrument Supervisor that a tolerance Tech of 2% was exceeded on the following components:
4 s
T Initial & Date (B).
An evaluation was made on the above problem (s)
Inst. Sup.
and the following corrective action was taken:
O t
I i
?
4
(
Attachm nt'5 MASTER 771.p Form SPD-1002-1 DtTKE POWER COMPANY (1)
ID No: IP/0/A/310/5D PROCEDURE PREPARATICN PROCESS RECORD Change (s)_ Y to
_4-Incorporated 3
(2)
STATION:nr m.
(3)
PROCEDURE TITLE: Engineered Safeguards System Analog Channel C R.B.
Pressure Switch Calibration and Pressure Switch Contact Buffer Tests (4)
PREPARED BY:
No_
DATE:
Y - $ -67 (5)
REVIEWED BY:
h DATE:
/d [d Cross-Disciplinary Review By:
N/R:
(6)
TEMPORARY APPROVAL (IF NECESSARY):
By:
(SRO)
Date:
By:
Date:
(7) APPROVED BY:
[.
Date:
Y-28"82 (8)
MISCELLANEOUS:
Revieve pp;c.cl 3y: [ d Date:
!/
b Reviewed / Approved By:
Date:
d
e FORM SPD-1001-2 DUKE POWER COMPANY NUCLEAR SAFETY FVALUATION CHECK LIST (1) STATION:
Oconee UNIT: 1 2
3 OTHER:
(2) CHECK LIST APPLICABIE TO:
IP/0/A/310/Su (3) SAFETY EVALUATION - PART A The item to which this evaluation is applicable represents:
Yes No A change to the station or procedures as described in the FSAR, or a test or experiment not described in the FSAR?
O If the answer to the above is "Yes", attach a detailed description of the itea being evaluated and an identification of the affected section(s) of the FSAR.
(4) SAFETY EVALUATION - PART B Yes No Will this item require a change to the station Technical Specifications?
If the answer to the above is "Yes," identify the specification (s) affected and/or attach the applicable pages(s) with the change (s) indicated.
(5) SAFETY EVALUATION - PART C As a result of the item to which this evaluation is applicable:
/
Yes No Will the probability of an accident previously evaluated in the FSAR be increased?
Yes No
, M ill the consequences of an accident previously evaluated Jn the FSAR be increased?
Yes No
/May the possibility of an accident which is different than any already evaluated in the FSAR be created?
Yes No g lli che probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
No M ll the consequences of a malfunction of equipment Yes important to safety previously evaluated in the FSAR increased?
Yes No May the possibility of, malfunction of equipment important to safety different than any already evaluated i the FSAR be created?
Yes No Will the margin of safety as defined in the bases to any Technical Specification be reduced?
If the answer to any of the preceding is "Yes", an unreviewed safety question is involved.
Juptify the conclusion that an-unreviewed safety
"* 4 o ved.
Attach additional pages as necessary.
question is or is (6) PREPARED BY:
j DATE:
/
M (7) REVIEWED BY:
/
[ --
DATE:
/ 2.
7t
~
fjylc,jdckhy
IP/0/A/310/5D DUKE POWER COMPANY OCOMEZ NUEE.E :IIATION ENGINEERED SAFEGUARDS SYSTEM ANALOG CHANNEL C R.B. PRESSURE SWITCH CALIBRATION AND PRESSURE SWITCH CONTACT BUFFER TESTS 1.0 Purpose 1.1 To furnish a procedure for calibration of Reactor Building pressure switch instrumentation.
1.2 To calibrate digital computer inputs.
1.3 To perform the functional and operational tests.
2.0 References (Use Current Copy) 2.1 Vol. 1 & 3 BMC. NI & RPS & ESS inst.
2.2 881 System Checkout Procedure 2.3 FSAR, Section 7.1.3.
2.4 Technical Specifications. Section 4.1.1.
3.0 Test Equipment Required 3.1 Pneumatic Calibrator, W/T series FA 145 Range 0-30 psi, or equivalent.
4.0 Prerequisites Sign-off(s) on Enclosure 11.1 4.1 Supervisor has reviewed and initialed all portions of this procedure which are not applicable to the activity being perfo rmed.
The supervisor's review is not required if the procedure specifies.<ections to be omitted except as needed io e-'10w M'a3 M d * ' i :11t.
4.2 Verify all change = we the control copy are incorporst :d on the working copy.
4.3 Computer in operation.
2-4.4 Verify that other ES Channels are not in trip conditions.
5.0 Limits and Precautions Use proper precautions while working with components that have HIGH 170LTAGE,. HIE:NLGI or HIGH TEMPEFL~URE present.
_ 6.0 Unit Status Sign-off(s) on Enclosure 11.1 6.1 N/A 7.0 General Description This instrumentation is used to sense high building pressure which actuates the ES Building Spray System.
8.0 Major Components Component Description Reference BS4-PSS (PS-22)
Mercoid Snap Switen Control Bailey Ref.
BS4-PS6 (PS-23)
Mercoid Snap Switch Control Bailey Ref.
9.0 Equipment Specifications 9.1 Operating Range Instrumentation Designation Input Output BS4-PS5 (PS-22) 10 psig Close BS4-PS6 (PS-23) 10 psig Close 9.2 Computer Inputs From BS4-PSS (PS-22)
D1957 BS4-PS6 (PS-23)
D1958 10.0 Procedure CAUTION:
If any component calibration is out of tolerance by 2%,
proceed to Maximum Tole:nace Exceeded Sheet.
' 0.1 P re:r.s.:.=:.Sc * --' Ceutaet Baus."-* ".
In this case there is a pressure switch (N/0 Contact) applied to each relay circuit of the contact buffer.
Iherefore, in normal operation both lamps will be "0N".
3-10.1.1 Depress S1 (Switch at top).
DS1 should go "0FF" (lamp at top).
10.1.2 While holding S1, check auxiliary relay for proper indication (DS1 and DS2 on auxiliary relay will be N b:Eght).
The bui!r'-i spray trip lamp on the analog indicating panel should be bright.
10.1.3 Release SI.
All of the lamps in steps 1 and 2 should return to their initial states.
Check computer input D-1957 for-proper indication.
Also check for proper statalarm indication.
10.1.4 Depress S2 (Switch at bottom).
DS2 should go "0FF".
10.1.5 While holding S2, check auxiliary relay for proper indication (DS3 and DS4 on auxiliary relay will be bright).
The building spray trip lamp should once again be bright.
10.1.6 Release S2.
All of the lamps mentioned in Steps 4 and 5 should return to their initial states.
Check computer input D-1958 for proper indication.
Also check for proper statalarm indication.
10.2 Pressure Switch Setting Ccmplete Enclosures 11.2a and 11.2b using the following procedure to calibrate:
10.2.1 Close isolation valve, remove test tee cap, and connect Pneumatic Calibrator to Tee.
Sign off on data sheet.
10.2.2 Slowly increase pressure until respective contact buffer light goes "out".
Record pressure on data sheet under "As..Found".
NOTE:
DS1 is for PS-22 and DS2 is for PS-23.
10.2.3 If switch makes within the pressure tolerance specified, decrease pressure to zero, disconnect pneumatic calibrator, put cap on Tee, open valve, and return switch to service.
10.2.4 If setting is in error or needs to be changed, continue with Sue M..
1.1.1.'P. '4e:sove m t_ __.w. :n to expose the switch adj us t-ment.
10.2.6 Set input pressure to that value listed on data sheet at which you want the switch to make.
10.2.7 Adjust switch setpoint until switch makes.
4-10.2.8 If switch ma kes on pressure increase, - reduce the pressure until switch resets.
Slowly increase pressure until switch just makes.
If it makes at desired setpoint, record data on data sheet under "AS LEFT".
If still not within tolerance specified, repeat Steps 5, 6, and 7,
> until it is within.;tlerance specified.
10.2.9 Replace cover on switch and recheck calibration to insure switch is still within tolerance.
10.2.10 Decrease pressure to zero; disconnect pneumatic calibrator; replace cap on tee; open isolation valve and return switch to service.
Sign off on data sheet.
10.2.11 Insure that Building Pressure Analog Channel C is reset and not in the tripped state after both pressure switches are complete.
11.0 Enclosures 11.1 Sign-Off Sheet 11.2 Calibration Data Sheets 11.2.a PS-22 (BS4-PSS) 11.2.b PS-23 (BS4-PS6) 11.3 Tolerance Limit Exceeded Sheet I
4
ENCLOSURE 11.1 Sign-off(s)
Unit Status Date Began Prerequisites 6.1 Date Completed 4.1 WR#
4.2 Unit 4.3 4.4 Data Sheets 11.2.a 11.2.b PERFORMED BY 9
l i
l l
I l
l P
me o
t a
i
s DUKE POWER COMPANY OCONEE NUCLEAR STATION
- Ei!T d ENCLOSURE 11.2.a CALIBRATION DATA SHEET IP/0/A/310/5D ITEM Pressure Switch TEST EQUIPMENT USED
'!E G.
Mercoid ITEM SN
. TYPE APW-7041-153 Satpoint Tolerance t.5 7. of Span t.095 PSIG SYSTEM ES Building Spray INSTRUMENT NO.
PS-22 (BS4-PS5)
PRODUCT INSTRUCTION SPAN 19 PSIG Switch to cl'ose on pressure increase.
SWITCH LOC.
Step 10.2.1 Removal from service l
f Input Desired As As Left Error DS1 to Actuation Found Setpoint PSI Light Goes Switch Point (PSIG)
PSI PSI Out PSI' YES l NO 10 10 Test Tee Cap Replaced - Verified By All isolation valves-left.vgen - V "i-C 2v MAXIMU': ERROR IN PSI PERFORMED BY DATE
.O DUKE POWER COMPANY OCONEE NUCLEAR STATION
- WIT ;;.
ENCLOSURE 11.2.b CALIBRATION DATA SHEET i-IP/0/A/310/5D ITEM Pressure Switch TEST EQUIPMENT USED MFG.
Mercoid ITEM SN TYPE APW-7041-153 Setpoint Tolerance
.5 % of Span t.095 PSIG SYSTEM ES Building Spray INSTRUMENT NO.
PS-23 (BS4-PS6)
PRODUCT INSTRUCTION SPAN 19 PSIG Switch to close on pressure increase.
SWITCH LOC.
Step 10.2.1 Removal from service Input Desired As As Left Error DS2 to Actuation Found Setpoint PSI Light Goes Switch Point (PSIG)
PSI PSI Out PSIG YES NO 10 10 Test-Tee Cap Replaced - Verified By
. All isolation ralk s9 eft opma - Vedded eq ^
MAXIMUM ERROR IN PSI PERFORMED BY DATE a
p ENCLOSURE 11.3 IP/0/A/310/SD MAXIMUM TOLERANCE LIMIT EXCEEDED SHEET Initial & Date (A).
Notified Instrument Supervisor that a tolerance Tech of 2% was exceeded on the following components:
Initial & Date (B).
An evaluation was made on the above problem (s)
Inst. Sup.
and the following corrective action was taken:
I l-
.a~..>~.
I
,__- _