ML19261D847
| ML19261D847 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/13/1979 |
| From: | Jennifer Davis, Harold Denton NRC OFFICE OF INSPECTION & ENFORCEMENT (IE), Office of Nuclear Reactor Regulation |
| To: | Ahearne J NRC COMMISSION (OCM) |
| Shared Package | |
| ML19261D848 | List: |
| References | |
| FOIA-79-98 NUDOCS 7906260519 | |
| Download: ML19261D847 (36) | |
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MEMORA!'DUii FOR:
Commissioner Ahearne FRO *1:
Harold R. Denton, Direct
, Office of Nuclear Reactor Regulation Jchn G. Davis, Acting irector, Office of Inspection and Enfo cement iHRU:
Lee V. Gossick, Execut Director for Operations
SUBJECT:
STATUS REPORT ON DAVIS-BESSE UNIT NO. 1 This is in response to your memorandem dated t' arch 29, 1979 regarding this subject.
General statements on the status of Davis-Besse Unit 1 are set forth below.
Detailed responses to your specific questions are provided in the attachment.
1.
During startup and testing phases, the number of operating difficulties was not unusually high compared to other plants, and the difficulties appeared to be attributable, in large measure, to equipment shakedown and personnel errors due to lack of an intimate feel for plant operational response.
The duration of the " learning curve" at Davis-Besse has been longer than at other operating sites and the number of operating diffi-culties is higher than normal for a plant with operating experience.
Several of the events at Davis-Besse 1 have been unusual and sig However, we are not aware of any currem. mattech
'ninediate health and safety significance.
2.
Power testing was completed by the licensee on January 15, 1979.
The licensee and IE are still reviewing test results and final determinations have not been made as to the acceptability of some of the results.
The time for completion of tests has been longer c'nreb2e Ch 4
CONTACT:
F. Nolan 2313 136 49-28019 7906260f/9 6
i Cc:.aissioner Ahearne -
than nonnal, and this prompted Region III to obtain a licensee cormitment on January 10, 1979, to complete the progran in a tirely manner, tiost of the delays were due to plant outages during which time the tests could not be conducted.
The reasons for the delays have been nionitored by Region III.
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3, We have become increasingly concerned regarding the licensee's delayed action in evaluating potential problems and in cor-recting identified problems, and regarding the increased rate of noncompliance.
Region III managen.cnt met with licensee manage-c.en M nuuu n W/8 to discuss these problems.
On March 23, 1979, another meting was scheduled for April 4,1979, but was delayed due M "aff adjutnents related to the Three Mile Island incident.
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We on to identjfflicensee mnagerrant control systems which o A' equire change to assure irnrnvo-nt in renulatory performance, C.,
and then n.onitor th~e licensee's progress through periodic followup meetings.
(This apprcach was used w th Corronwealth Edison Company and resulted in a marked improvement in the regulatory perfonnance of the operating facilities).
4.
With one exception, enforcement action has been handled from hM the Region III office, Of the twenty-seven items of noncom-pliance cited in 1978, one was considered to be in the Violation category, and this resulted in a Headquarters I;atice of Violation.
This action, along with the acnagerent moeing mentioned in item 3 above, is the normal prbgression(for escalating enforcerent sanctions, lhd is 24 itb./
5 Major matters under review by f4RR at the present time are tce fuel pool expansion and the fire protection programs.
Each of these is in the final stages of review, and there are no out-standing technical questions,
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,i J hn G. Davis Harold R, Denton Director v Acting Director Office of fluclear Office of Inspection Reactor Regulation and Enforcement Attachments:
(See next p. age) cc:
(See next page) 6 as
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a Coir,issioner Ahearne.
At ta ch.Znts:
1-7. Responses to Specific
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Questions 8.
Backup Information (Coc.r. issioner Ahearne only) a cc w/a tts.1-7 and
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att. 8 - List of Background Infonr.alion Provided Chainc.an Hendrie Cou.aissioner Gilinsky Cor.inissioner Kennedy Commissioner Bradford Samuel J. Chilk, SECY Al Kenneke, PE Len Bickwit, GC 2313 138 e
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i Question What are, the problems that created the difficulty in completing the pcwer ascension testing for this reactor? Have they been adequately resolved? Are there any test results which the staf f has not deter-mined to be satisf actory?
Answer The prcblems that created the dif ficulty in completing the pcwer ascension testing ware primarily equip ent problems which prevented testing. A chronclogy of major outages (5 days or more) is shown in Ap;;ndix A to this attachment.
While some problers can be expected as part of a normal plant shake-dcwn, several unusual problems have occurred at the Cavis-Eesse plant.
The more significant problems have been:
1.
Repetitive failure of auxiliary feedwater pumps during late 1977.
(See Attachment #5 for details).
2.
Turbine condenser tube leaks occurring during late 1977 and early 1978.
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3.
Primary pump seal failures which have repeatedly caused or extended outages since initial startup.
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Removal of burnable poison rod assemblies during the middle of 1978.
5.
Correction of diesel generator electrical problems.
6.
Repair of secondary system steam leaks.
In addition to equipment problems, the testing program was extended because the unit oparated at 75% power during the power shortage caused by the coal strike in February and March 1978 and because of a limitation of 70% power during April 1978 due to the burnable poison rod problem discovered at Crystal River. Also, a. number of the power ascension tests were repeated following the burnable poison removal outage te verify that system parameters were in accordance with design calculations.
Region III concerns over the delays in ccmpleting the testing program prompted a licensee letter to Region III dated January 10,1979, which included a licensee commitment to initiate completion of all major tests by January 13, 1979, or promptly after return to operation.
(The unit was not operating at the time).
The licensee satisfied that conni tment.
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The power ascension testing program was completed by the licensee o January 15, 1979.
The licensee determined from test results that normal opera cions at full power can continue.
.cgion III has no_,,,?
- c. san to question the licensee's determinatjoMvever, some ques-tions related to four tests remain to be resolved.
The tests and the status of action necessary for resolution are as follows:
Test Status of Action 1.
Pcd Drop At the request of Region III, NRR is presently reviewing the adequacy of the online computer to accurately or conservatively monitor core param-ete s.
IE headquarters and NRR will be requested to review the impact of selective loading report on previous NRR analysis of the licansee's rod-drop analytical technique.
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Reference:
IE Report 50-346/79-04, paragraph 2) 2.
Reactivity Coefficients RIII will review the results during at Power and Rod Worth the next inspection to determine the Measurements value of moderator temperature co-efficient.
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Reference:
IE Report 50-346/79-04, paragraph 4.c) 2313
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n 4-Test Status of Action 3.
Shutdown from Cutside Rill will review the results during Control Room the next inspection to determine if the test results are rejectable due to lack of ma'reup tar.k level indi-cation at the remote shutdown panel.
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Reference:
IE Report 50-346/78-30, paragrr 4
Turbine / Reactor Trip R N' will review the licensee's eval -tion for deleting this test based on the results of the unit load r jection test.
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In addition to the above tests, RII! has ni't yet conplet+ d its evalu-at. ion of some of the other test results ac epted by the licensee.
This is planned to be accorcplistied by May 1979 RIII does not have reason s
to auestion these test results.
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i Appendix A CHRONOLOGY OF MAJOR OUTAGES Ou tage Date length i
Started Type (days)
Discussion 8-28-77 Maintenance 5
Frimary coolant pump repair.
9-4-77 Equipment 15 Repair of turbine generator Failure condenser tube leaks.
9-24-77 Equipment 22 Spuricus trip of the Steam -
Failure Feedwater Rupture Control System (SFRCS) and a pres-surizer relief valve failure resulted in primary system depressurization.
(See answer
.to Question 4) 10-23-77 Equipment 5
Spurious trip of the SFRCS Failure occurred.
The outage was extended by troubleshooting of the SFRCS and repair of an auxiliary feedwater pump.
11-29-77 Equipment 5
Reactor tripped due to Failure erroneous load demand signal.
During the transient a trans-fer of auxiliary power was made incorr >ctly resulting
'n' loss of reactor coolant pumps and feedwater pumps.
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Appendix A.
Outage D3te Length Started Type (days)
Discussion Both diesel generators started but one tripped on overspeed.
Loss of indication of oressur-rcer level or " red during eent.
(S e ann.er to,
Question 4) b-l-6-78 Personnel 11 Feedwater pump put in auto Error at too low power level.
Outage was extended by failure of auxiliary feed-wa ter pump.
2-24-78 Equipment 5
Turbine generator condanser Failure
. tube leak repair.
3-29-78 Personnel 5
Reactor tripped whan oper-Error ator incorrectly adjusted prin.ary temperature controller.
Outage extended for maintenance.
4-29-78 Design and 90 Reactor shut down for removal Equipment of burnable poison rods.
Problems with diesel gener-ator sequencer (See answer to Question 5) and primary pump seals extended outage.
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Appendix A Outage Date Length Started Type (days)
Discussion 8-14-78 Equipnient 15 Reactor shutdown due to two Failure failed radiation n:onitors for the Safety Feature
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Actuation System (SFAS).
Outage extended due to primary pump seal failures.
9-28-78 Equipment 5
Failed Reactor Coolant System Failure (RCS) ficw transmitter caused a transient resulting in over-feeding of a steam generator and a reactor trip due to pric.ary icop low pressure.
10-9-18 Equipment 18 Repair of prir.ary punp seals.
Failure 12-16-78 Equipment 16 Repair of e>t:: action steam Failure line bellows in seccndary system.
1-14-79 Equipment 15 Repair of primary coolant Failure pump seals.
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Attachr.ont 2 Question i.' hat outstanding licensing issues are there? Have any exemptions been granted for this plant? If so, what are the justifications for those e
e> captions?
Answer We are attaching the Project Manager Update Report for Davis-Besse 1 which provides all the currently identified licensing actions for Davis-Besse 1.
Tne report is typical of an operating plant.
The i;iost significant outstanding issues are:
the spent fuel pool expansion; the fire protection safety evaluation report; Appendix I Technical Specification review, the seismic reevaluation; overpres-sure protection; and the inservice testing and inspection programs review.
We will soon be completing our review of the spent fuel pool expansion issue, the HPSI-LPSI flow resistance issue, the change du'e to the reasured pump coastdown curve, and the reactor system flow and pressure drop data.
The other issues are under continuing review.
"No exemptions have been issued for this plant.
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Ouestion Is part of the reason for the low capacity factor attributable to the plant operator personnel? How do the plant operators compare
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with other plant operators in Region III? In particular, I would note the recent notifications of the Commission about problems with the valving errors on the ECCS for both the icw pressure and high pressure safety injection systems.
Answer Some operator errors have resulted in plant outages. However, the impact on the capacity factor of those outages is small. Of the major plant ou, described in Attachment #1, only two were the result of personnel error.
Even though those outages were initiated by personnel error, other complicating factors contributed to the length of the outages.
NRR. experience with the licensing of Davis-Besse 1 operators indicates that the test results compare favorably to the test results of oper-ators at other plants.
For example, only one operator license was d,enied on the basis of NRR examination at Davis-Besse 1.
In comparison, at another facility of similar design licensed during the same time 2313 147 e
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Attachr.cnt 3 frane, licenses were denied seven reactor operators and six senior reactor operators.
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Regarding comparison of plant operators at Davis-Besse to other plants in RIII, the overall number of operator errors during early phases of operation has not been markedly different from other facilities.
However, the rate of operator errors has not decreased as would normally be expected with the gaining of operating experience. The rate has re'nained the same at best and m2y have increased slightly.
RIII has urged the licensee to reduce these errors and this matter was, discussed during October 1977 and August 1978 management meetings.
Because of more ecent sit nificant operator errors, the matter was scheduled to bediscussedada during the planned April 4,1979 meeting.
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g We evaluated the LER's of five reactors (Beaver Valley 1, Crystal River 3, Davis-Besse 1, Salem 1, St. Lucie 1) within a similar time frame and having similar reporting requirements. During the first year of commercial operation, Davis-Besse 1 had the highest n.',ber of LER's (130), but about an average number of LER's attributable to personnel error (20).
The second year of commercial operation (pro-rated from 6 months data), indicates about the same total number of 9
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Attach-ent 3 LER's for Davis-Besse 1 and an increase of those attributable to perscnnel error.
Beaver Valley also remained fairly constant for the second year of commercial operation, but the total LER's and
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those attributable to operator error decreased at the other three facilities during the second year of cor.imercial operation.
You noted the valving error on the ECCS for both low pressure and high pressure safety injection system which was performed by an auxiliary operator. The operator immediately ra;^rted his actions to his suparvisor and the situation was corrected in about eight minutes.
While the error was promptly corrected, it is a matter of concern.
The situation was reported by Region III as a potential abnormal occurrence, and it is being evaluated for possiu'e escalated enforcement actions.
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n Question In vie.< of the rr-cent Three Mile Island accident, are there any plant systa.as related to the safe operation of the plant which have experienced
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specific difficulties?
In particular, please provide me with the licensing staf f's analysis of the September 24, 1977, event and all major subscquent events that have occurred at the plant.
Answer Thare are sevaral plant systens related to safe operation of the plant which have experienced specific diffic0lties.
Most of these difficulties were the subject of LER's, and Region III assessment of the adecuacy of corrective actions was carried out as outlined in Attachment 5.
For those difficulties not reported as LER's, Region III resolved those matters by identifying them as unresolved items during inspections and subsequently resolving them based on information provided by the licensee.
The licensing staff (NRR) reviews LER's and other matters of signif-icance at cach facility.
However, they do not routinely prepare a documented staff analysis unless requested by IE.
The staff's analysis is in the form of inspection reports and internal memoranda.
The major events and actions taken to resolve them are described below.
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Reactor Coolant System Rapid Depressurization tsL4 g g[
On September 24, 1977, a spurious partial trip developed in
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the Steam and Feede 6ter Control System (SFRCS) while the plant was operating at 9% power.
The SFRCS is designed to automatically provide a heat sink for the Reactor Coolant System (RCS) under postulated operational steam line break.
The spurious trip resulted in the normal feedwater to one of the two steam generators being isolated. The feedwater isolation in turn caused the RCS to increase in temperature and pressure as the steam generator which had feedwater isolated began to decrease in level and thereby lose its effectiveness in cooling the RCS.
There was nothing unusual at this point with regard to system response to the spurious trip.
However, when the RCS pressure increased to the setpoint of the pressurizer power operated relief valve, the valve opened as designed but failed to reseat due, in part, to a missing relay.
This malfunction resulted in a loss of coolant condition through the stuck open valve for a period of approximately 20 minutes until the malfunction was diagnosed and corrected by the operator. The loss of coolant 2313 151 e
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Attaci.. cnt 4 ccndition resulted in a rapid depressurization of the RCS and formation of voids in the core.
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There were several equipment malfunctions involved in this e
event.
Those malfunctions were the spurious SFRCS trip, the relief valve sticking open, and the failure of one auxiliary fee 6sater purp to come up to full sp:ed.
These malfuncticns were corrected and systems and components tasted prior to returning the plant to pcr.cr on October 17, 1977.
The licensee informed the NRC (Regicn III) of the event on September 25, 1977, and an inspector was dispatched to the site the next day. The licensee subsequently reported the event in reports dated October 7,1977 (NP-32-77-15), and November 14, 1977 (NP-32-77-16).
The Region III inspection of this event is documented in Inspection Report 50-346/77-32, Paragraphs 3 and 11. Licensee reports and the Region III inspection report relating to this event are enclosed 'for your information.
It was concluded by both the licensee and Region III that there was no fuel damage, there was no damage to the integrity of the RCS or RCS components, and equipment malfunctions had 2313 152
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Attac!...cnt 4 been corrected.
Based on these findings, it was concluded that resumption of operation of Davis-Besse 1 did not constitute a hazard to the public health and safety.
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Region III kept NRR inforn.ed of the event but did not request an NRR analysis.
I;RR personnel from appropriate technical groups visited the site prior tc resu;c.ption of operation and reviewed event information with Region III inspectors.
Input from those NRR personnel was used by Region III in evaluating the event.
b.
Loss of Indicated Pressurizer Lavel On November 29, 1977, a false power demand signal daveloped while the plant was cperating at 40% power.
Power auto-matically increased to 50% and at that point the plant was automatically shut down. Although it did not significantly
-affect the event results, a subsequent operator error in trans-ferring electrical loads resulted in a loss of offsite power condition.
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O Attach.r.ent 4 5-During this event the water level i the pressurizer dropped below the lowest indicated level.
ii..ever, there was no information to suggest that there was ormation of voids in the core such as occurred during the September 24, 1977, event.
As a result of inspector concerns about the potential for emptying the pressurizar and formation of steam voids in the RCS during cooldown operational occurrc.es such as this event, the licensee was requested to perform an analysis to determine the lowest pressurizer level attained during this event.
The licensee's analysis indicated the pressurizer did not.erpty and calculated the lowest pressurizer level to be approximately 34 inches above the bottom of the pressurizer.
An outgrowth of the Region 111 review of the minimum pres-surizer level issue was the development of information which indicated that, under certain postulated conditions, the pressurizer could empty and allow formation of steam voids
. in the RCS.
These postulated conditions did not exist during the November 29, 1977, event.
Upon receipt of this information, Region III requested that prior to plant startup the licensee perform a bounding analysis to determine if an unreviewed safety question existed.
Region III also requested NRR 2313 154 e
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Attachmant 4 6-assistance in determining the acceptability of the licensee's analysis.
The licensee submitted his analysis to Region III and NRR on December 22, 1978. The analysis concluded that emptying of the pressurizer under the conditions postulated did not represent an unreviewed safety question since system and fuel failures would not occur.
Region III and ',RR agreed prior to plant startup that an unreviewed safety question did not exist.
g There were two other significant problems identified during the November 29, 1977, event.
One was'the overspeed tripping of an e7.crgency diesel generator (EDG) and the other,;as related to a design deficie.ncy concerning General Design Criterion 17. The licensee was unable to determine the cause of the abnormal EDG operation but took several actions to reduce the probability of recurrence. The GDC 17 design problem was resolved at a later date and is discussed in
. Paragraph 4.c below.
Region III concluded that licensee corrective actions including those to prevent recurrance of EDG overspeed trips, to modify
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procedures to prevent operators from inducing a loss of offsite 2313 155 ew w
l-ttachmant 4 power condition, and to assure the RCS thermal-pressure transient did not compromise the RCS integrity were acceptable.
NRR assistance was requested and obtained in the resolution of the problems related to pressurizer level and GDC 17.
Review of the November 29, 1977, event and resolution of the pressurizer level and other concerns 'ics docur.ented in the following inspection reports:
50-346/77-34, Paragraphs 5 and 11 50-346/78-06, Paragraphs 2_and 9 50-346/78-17, Paragraph 5 50-346/78-27, Paragraph 8 50-346/78-30, Paragraphs 2 and 15 These inspection reports are enclosed for your information.
Also enclosed is (1) the licensee's analysis of the pres-surizer level question which was submitted to Region III and NRR on December 22, 1978; (2) a Transfer of Lead Responsi-bility from IE to NRR dated January 16,1979; and (3) the LER NP-32-77-20 dated December 12, 1977, describing the loss of power situation.
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Design Deficiency in Automatic Electrical Bus Fast Transfer (GDC 17 Issue)
As a result of the operator-induced loss of offsite power condition described in the precceding paragraph, Region III requested NRR assistance in assessing the acceptability of the plant electrical design for fast bus transfer.
NRR concluded the design did not satisfy the requirements of Criterion 17 of Appendix A tc 10 CFR Part 50.
Region III obtained a commitaant from the licensee, as confirmed in an immediate action letter, to place the plant in an electrical alignment that satisfied GDC 17 and was consistent with all other license conditions. A folicwup inspection was also conducted.
Identification and resolution of this design deficiency is documented in the following:
Inspection Report 50-346/78-13, Paragraph 13.b Memorandum from J. F. Streeter to R. W. Woodruff, dated June 6,1978 2313 157 e
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Attachc. ant 4 Memorandum from R. W. Woodruff to J. F. Stoltz, dated August 17, 1978 -
Memorandum from J. F. Stol to E. L. Jordan, dated October 5, 1978
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M:aorandum from E. L. Jordan to G. Fiorelli, dated October 13, 1978 Ic. mediate Action Letter fecm Regicn III to the licansee dated October 31, 1978 Letter (LER 78-104) from the licensee to Region III dated
!?cvember 3, 1978 Inspection Report 50-346/78-26, Paragraphs 6 and 7 These documants are enclosed for your infon.ation.
Althouch the licensee has not made per.anent modifications to allow more flexibility in alignment of the electrical systems, continued operation with the present alignment is acceptable.
d.
Inoperability of Emergency Diesel Generator Sea,uencer On June 5,1978, during a routine refueling surveillance test, the licensee discovered design and initial installation 2313 158 e
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At tachn.ent 4 deficiencies that would have prevented the automatic sequencing of safety components on to the emergency diesel generators (EDGs) during a loss of offsite power in con-junction with a loss of coolant accident (LOCA). However, the sequencer would have worked as designed in the event of a loss of offsite power or LOCA not in conjunction with each other.
An inspector was dispatched to the site on June 6,1978, to assess the adequacy of the licensee's cor-rective actions. The Region III inspection of the event which is documanted in Inspection Report 50-306/78-19 identified the following four basic problem areas:
(1)
Inadequate control of terminal slide links in safety related panels.
(2)
Inadequate testing methods to demonstrate operability of control systems.
(3) Conflict between design drawings and the scheme and vendor as-built drawings.
(4)
Compromise of safety systems design feature by authorized procedur"s.
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Attachn.ent 4 Based on these problems, a correctiva action program was established to insure that the circ ;it problems associated with the EDG seouencer were corrected and did not exist in other safety related systems. This program..;as confirmed in an Ir...adiate Action Letter from Region III to the licensee on June 12, 1978.
Further licensee review of the original problem uncovered additional problems including some related to procedures, incorrect relay settings, and wiring errors. These are discussed further in LER's NP-32-78-05, HP-32-78-06, NP-32-78-07, and NP-33-78-87.
E 3ed on the results of the corrective action program and licensee action on identified problems, Region III concluded that return of the plant to operation did not constitute a hazard to the public health and safety.
Region llI did not request NRR review of the problem.
However, Region III did consider the event to be of such significance that a potential abnormal occurrence report was issued to IE Headquarters and the enforcement action was issued from IE:HQ to highlight our concern.
In addition, this problem was highlighted in 2313 160 g
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Attachr<ent 4 12 -
a r:.anagement meeting conducted August 16, 1978, and it is docu-c.ented in Inspection Raport 50-346/78-23.
4 Licensee reports, Region III inspection reports, the Irr:diate Action Letter, and enforcen ent corresrcndence relating to this e.ent are enclosed for your infor.ation.
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Question Have any safety systems been subject to a large number of LER's?
Please provide details of the staff's review of these items.
Answer There are several safety systems at Davis-Sesse 1 which have been subject to a large number of LER's.
Among these systems are the Safety Features Actuation System, Auxiliary Feedsater System, Cocay iicat Remaval System, Steam and Feedwater Line Rupture Contr:1 System, and the Reactor Protection System.
All of the LER's submitted cn these systems were reviewed in accor-dance with the IE procedures pertaining to LER review contained in IE Manual Chapter 2515, Procedures 907128 and 9270G3.
These proce-dures require that each LER be screened and evaluated in the IE Regional Office to determine such things as (1) if the event consti-tutes an Abnormal Occurrence, (2) if there is a need for immediate site inspection effort, (3) if the event represents a matter generic to other components within the facility or to other facilities within the Region, (4) if there is c need to inform Licensing Boards, and (5) if the licensee's corrective action is appropriate.
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. Not all of the LER's are required to be subsequently reviewed on site by an inspector; however, all of the LER's which are required to be reported prcmptly (telephone and 14 day written re,nort) are reviewed
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on site and at least 10% of the LER's submitted in a longer time frame (30 day written report) are required to be revie-ed on site.
Additionally, the procedures allow for on-site review of other LER's decced by the inspector or his surarvisor to require site inspection.
LER's reviewed on site are documented in inspection reports.
The Region also maintains docunentation on its in-office review of all LER's.
Using this review program, many of the large number of LER's submitted on safety systems at Davis-Besse 1 were reviewed on site.
As an example, the following table shows that 6 of the 18 LER's submitted for unplanned events on the Auxiliary Feedwater System were reviewed at the site.
Inspection Report Event Date LER Number Documentino IE Review 7/27/77 NP-32-77-ll 50-346/78-01, 50-346/78-13 8/3/77 NP-33-77-46 8/4/77 NP-33-77-53 8/6/77 NP-33-77-45 50-346/78-13 8/7/77 NP-33-77-51 8/7/77 NP-33-77-52 S
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Attachc. ant 5
- 8/21/77 NP-33-77-62 50-346/78-13 10/16/77 NP-33-77-80 10/20/77 NP-32-77-17 50-346/78-01 10/25/77 NP-33-77-83 11/8/77 NP-32-77-18 50-346/77-34; 50-346/78-01 12/11/77 NP-33-77-110
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12/16/77 NP-33-77-113 50-346/78-04 12/28/77 NP-33-77-Il6 1/6/78 78-005(NP-33-78-06) 3/16/78 78-027(NP-33-78-33) 1/2/79 79-002(NP-33-79-03) 2/20/79 79-030(NP-33-79-33)
The office and site revicas of LER's related to the Auxiliary Feed-water System resulted in a determination in December 1977 that the large number of LER's indicated a need for an exhaustive engineer-ing evaluation.
Region III obtained a licensee commitment for such an evaluation which was completed in January ~1978. This ccamitment and licensee action is documented in Inspection Report No. 50-346/
78-01, Paragraphs 8 and 11.
Since that time, corrective actions have been taken and only 3 LER's have occurred in the year since the evaluation was completed. Typical in-office and onsite review documentation of the Auxiliary Feedwater system problem is shown in the enclosed screening and evaluation form.
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. Althcuch exhaustive engineering evaluations by the licensee, such as the one for the Auxiliary Feedwater System, have not been deemed by Region 111 to be necessary for all of the safety systems which have been subject to a large number of LER's, Region III does require such evaluations if it appears necessary frem periodic LER trend analyses.
The licensee's corrective actions on the individual LER's have been judged by Region III to be adequate to assure proper system performance.
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Attachmant 6 Question How do the facility change requests for this station compare with other operating reactors which have been licensed in about the same time frame?
Answer
're cor.ipared the number of facility change requests approved for impicmantation at Cavis-Sesse 1 during the interval April 22, 1977, (cperating license issuance date), thrcugh calendar year 1978, to the nunber approved for implementation during a similar interval at other cporating reactors which were licensed in about the same time frame.(D. C. Cook 2, Browns Ferry 3, Crystal River 3 and Farley 1).
We also made a ccmparison to a similar operating reactor (B&W design) which was licensed on February 6,1973.
The reliability of these numbers is somewhat questionable because of different methods of defining design changes, the variable status of plant completion at the time of OL issuance and the different selection criteria for change requests to be. counted.
However, recognizing these variables, our review indicated that the number of facility changes at Davis-Besse 1 approved for implementation during the period 2313 166 e
AttacInc.ent 6 reviewed is comparable to the number of changes approved at the other operating reactors.
Part of our review was completed prior to receipt of your letter and is documented in Inspection Report 50-346/79-05, Paragraph 6.
Wnile this review shows the number of change requests to be comparable to other sites, we have expressed our ccncern to the company that such requests have not been handled properly.
Our concerns relate to the adequacy of the staff size, the systems and the timaliness of action.
(Also, see Attachment 7).
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Attachmant 7 Question Are defects which have been identified in safety-related systems analyzed and corrected in a timely fashion by the licensee?
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Answer Defects identified in hardware components have, in general, been corrected in a timely manner by the licensee.
Honaver, we have had continuing pecbiccs with timeliness of the licent e's analysis, resolution, and correction of potential problems.
The effectiveness of the licensee's managemant control systems'to assure timely action was discussed during a. eeting with corporate management in August 1978 (Inspection Repcrt No. 50-346/78-23). A recent investigation high-lighted continuing problems in the licensee's timely resolution of facility change requests.
This matter was again scheduled to be a topic at the meeting scheduled for April 4,1979.
This matter will be discussed at the next management meeting.
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Si??C? ?:.Z IS?07"ATICS FCR.ES?0SSE TO CC12CSECNIR.'.;r.A?5E'S LET.'IR TO H. DESTON AND J. DAVIS, DATED MARCH 29, 1979 30CL.fENT APPLICA3LE ATTACEMENTS Inspection Report 50-346/78-30 1, 4 Letter dated 1/10/79, J. Grant to J. Keppler 1
Inspectica Report 50-346/79-04 1
Proj.ect Mar.ager Updat'e Report for Power Reactors 2
Inspectica Rcport 50-346/78-23 3, 4, 6, 7 Inspectica Report 50-346/77-31 3, 6 Inspectica Reper: 50-346/77-34 4
Inspecticn Reper: 30-346/78-06 4
- nspecticn Ecper: 50-346/73-13 4
Ins pectica Repert 50-346/78-17 4
Inspecticn Eeport 50-346/78-19 4
Inspectica Rcper: 50-346/75-26 4
Inspecticn Repor: 50-346/75-27 4
Letter dated 9/22/73, J. williamson to N. "aseley 4
Tm:eciate Action Letter dated June 12, 1973 4
letter dated 9/01/78, N. Moseley to J. 'aillic: son 4
.M:crandu dated S/14/75, J. Keppler to N. "eseley 4
"eco ran d u dated 6/9/75, J.
Streeter to 7 Woedruff 4
12:eciate Action Letter dated October v, 1978 4
-Reportable occurrence 50-346/77-16 (N?-32-77 Supplement) 4 Reportable Occurrence 50-346/78-055 (N~c-32-7S-05) 4 Repertable occurrence 50-346/73-057 (N?-32-75-06) 4 Repor:able Occurr ance 50-3c6/ 7 8-061 -(N?-12-7 5-07) 4 Feportable Ocrurrence 50-346/73-104 (5?-32-73-11) 4 Reportabic Occurrence NP-32-77-20 4
Menorandum dated S/17/78, R. Woodruf f to J. Stol:
4 Licensee Analysis of Pressurizer level, dated 12/22/78 4
Mascrandua dated 10/5/78, J. Stol: to E. Jordan 4
Macorandus dated 10/13/7S, E. Jordan to G. Fiorelli 4
Raportable occurrence NP-33-78 ;'
4 Reportable occurrence NP-32-77-16 4
Transfer of Lead Responsibility, dat-d 1/16/79 4
Inspection Report 50-346/77-32 4
Inspection Report 50-346/78-01 5
Screening and Evaluation Form for NP-32-77-18 5
Inspection Report 50-346/79-05 6
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