ML20046D732
| ML20046D732 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/17/1996 |
| From: | Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Diaz N, Dicus G, Shirley Ann Jackson, Mcgaffigan E, Rogers K, The Chairman NRC COMMISSION (OCM) |
| References | |
| NUDOCS 9609180322 | |
| Download: ML20046D732 (46) | |
Text
-
A I
/
pON s>m 8W q
t UNITED STATES y
j NUCLEAR REGULATORY COMMISSION t
WASHINGTON. D.C. 305t&M01 September 17, 1996 MEHORAN3UM TO:
Chairman Jackson Commissioner Rogers Commissioner Dieus Commissioner Diaz Commist.ioner McGaffigan Janas M. Tavlo/rb FROM:
Executive Dire for.0p.
.cas SUBJECl:
FINAL SAFETY ALYSIS REPORT INSPECTION RESULis AND PLAN"E0 1
IMPROVEMENTS l
In my memorandum to Chairman Jackson dated December 28, 1995, the staff committed to conduct activities that would measure the extent to which problems encountered at Millstone Unit l_ regarding complianca with the final i
safety analysis report (FSAP) existed at other facilit%.
Th u, cemorandum j
provides the results rf the broad-based FSAR inspections and discuses the significance of the identified discrepancies.
It also describ7s shore term and long-term planned improvemants, as well as licensee actions.
Backaround and Methodoloav On January 25,1996, ;ne Office of M.aclear Reactor Regulation (NRR) issued short-term inspectius guidance to ad regional offices to supplement the existing level of FSAR reviews that were accomplished durina routine NRC inspections.
The revised guidance required inspectors to verify selected FSAR commitments by reviewing the applicable portions of the FSAR during inspection preparation and verifying t'iat the commitments had been properly incorporated into plant practices, procedures, and/or design.
ihe guidance was extended indefinitely on March 15, 1996.
The s'.17 monitored the prcyress of the inspections and compiled a table of FFAR discrepancies that were identified during the period from January 25 tarough April 26, 1996.
In my memorandum to the commission dated May M,
1996, I provided the table of inspection results and a copy of the interim inspection guidance. The Ma" 24 memeranoum noted problems and potential violations relating to FSAR ccuracy, design control, and 10 CFR 50.59 implementation by several licensees.
The staff requestM and abtained addit'onal information for the most significant discic: ancies (catcgorized as violations or potential escalated enforcement issues) to determine whether the actions of the licensee and the NRC staff relating to the identification and resolution of these issues had been timel.r.
CONTACT.
D. L. Gamberoni, NRR
""1l ~-""
,e a cua are q'
9604180322 960917 CF ADOCK 05000245 CF
2 The table of inspection results has been modified to reflect this additional iaformation (see attachment).
Finally, the staff performed a probabilistic risk analysis (PRA) screening of all 219 FSAR discrepancies to c+aluate their potential. risk (safety) significance, to ensure that individual issues of potential significance were included in inspection follow-up and enforcement processes.
Results inspection PestJ1ts The short-term inspections, which were documented in 130 inspection rcports from 70 sites, identified 219 discrepancies from January 25 through April 26, 1996. The staff documented all findings, regardless of their significance.
The findings in the table of inspection results do not include the results of licensee self-as; essments. The staff reviewed the discrepancies and noted the follcwing:
Without regard to the safety significance of the specific discrepancy or the scope of the ii.spc-ction activity, FSAR discrepancies were identified at more than 85 percent of the plants.
About one-third of the inspections did not identify any FSAR discrepancies.
Approximately two-thirds of the discrepancies were identifted by the NRC staff; one-third were identified by the licensee.
i The types of discrepancies were divided nearly equally between design, operations /precedures, and edministrative.'
Plant change; may be required to resolve approximately 20 of the 219 FSAR discrepancies.
The significance of approximately 85 percent <f the discrepancies was minor.
Analysis of Sionificant Findinos Eight of the discrepancies have been or will be the subject of escalated enforcement action. Se;en of these eight discrepancies were design problems (either design errors or improperly performed modifications) and had poor 10 l
CFR 50.59 implementation as a significant root cause. Although several of the desigr issues had existed in the plants for many years, in most cases t
' Design - a discrepancy between the plant and the design as described in the FSAR, j
Operations / Procedures - a discrepancy between a plant procedure or operation and the FSAR.
Administrative - minor editorial discrepancy or administrative problem with the FSAR.
3 licensees initiated prompt corrective actions shortly after the identification of the issues. All but one of these issues were addressed in a timely manner after the NRC staff became involved.
In that case (heating, ventilation, and air conditioning isolatior. campers installed in reverse), untimely follow-up by the licensee and the NRC staff allowed it to remain uncorrected for about 4 years. This case is being considered for escalated action.
4 Twenty-sevendiscreppncies(approximately12 percent)resultedinseverity i
level IV violations.
The types of discrepancies included 10 design problems,14 operations / procedures problems, and 3 administrative problems.
All but one of these issues were addressed in a timely manner after the NRC staff became involved, but several of these issues also had existed in the plants for many years.
In most cases, licensees initiated prompt corrective actions shortly after identification of the issues.
4 1
Thirty-fourdiscrepancies(approximately15 percent}arestillbeingreviewed by the regional offices as unresolved items (URIs).
On the basis of the NRC staff's preliminary review of the URIs, only a few appear to be potentially safety or prograamatically significant. One such URI involved a plant that performed an extensive FSAR self-assessment and identified several hundred minor FSAR discrepancies. The regional office responsible for this plant is closely reviewing the licensee's corrective actions.
NRR staff members knowledgeable in PRA performed a PRA screening review (qualitative) of the 219 FSAR discrepancies (summary descriptions only) and i
did not identify any generic risk implications. However, seven individual discrepancies were identified as having some potential risk significance.
These discrepancies primarily involved the potential for common cause failures of redundant trains of safety equipment or spacial interaction (e.g. high temperature steam environment, flooding).
i All seven of these discrepancies were identified as imving some safety significance by the inspection and enforcement programs; three of the discrepancies resulted in escalated enforcement actions, two resulted in level l
2The staff notes that its review of the 219 discrepancies identified that in some cases, discrepancies were identified as deficiencles or weaknesses when violations should have been issued.
NRR has concluded that the most risk significant issues have been appropriately dispositioned.
Therefore, the staff does not intend to revisit the enforcement decisions made or initiate actions where enforcement was not initiated. The consistency of future actions should be increased by use of the revised Enforcement Policy (NUREG-1600) following its approval by the Commission (SECY 96-154, July 5,1996; a revision to the Enforcement Policy to address departures from the FSAR in violation of 10 CFR 50.59 and for failures to update the FSAR in violation of 10 CFR 50.71(e)).
3 i
An unresolved item is a matter about which more information is required to ascertain whether it is an acceptable item, a deviation, or a violation.
For these FSAR discrepancies, the staff will obtain arMitional information and i
1 resolve the issue based on safety and regu1 A ry significance.
I
l 4
IV violations, and two remain under staff review as unresolved items. The staff believes the current inspection and enforcement program treated these l
issues appropriately.
On the basis of the results (small number of potentially s@ if kant issues and the fact that current inspection and enforcement programs successfully identified the potentially significant discrepancies), further detailed l
(quantitative) PRA analysis does not appear to be warranted.
l Relationship to Millstone Lessons Learned Group Activities The FSAR inspection results were provided to the NRC Millstone lessons learned group for consideration in a broader context.
In this regard, it should be i
noted that the level of design information contained in a licensee's FSAR varies greatly, depending on the vintage of the plant.
Even for the most recently constructed plants, however, the FSAR is only a small part of the information that forms a plant's design bases.
10 CFR 50.2 defines design Dases as, "... that information which identifies the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design." A complete design bases document would consist of the design bases (contained in such documents as the Code of Federal Regulations, industry codes and standards, and applicable Regulatory Guides) and supporting design information such as computer codes, analyses and calculations, reports 1
and engineering studies, and engineering evaluations.
Therefore, the FSAR l
review effort does not get at the heart of a licensee's design bases.
Recent inspection findings have indicated that design bases information has not been appropriately maintained and implemented at certair, plants.
These findings raise questions as to whether licensee programs to maintain confit. ration control are sufficient to demonstrate that plant physical and functional characteristics are consistent with the design bases and whether operating plants are being maintained in accordance with their design bases.
Several errors in the FSARs were identified, reflecting a programmatic weakness in maintaining the accuracy and the consistency of information in the FSAR. However, the staff identified much more significant deficiencies involving engineering calculations and analyses and inadequate design modifications at some sites (including Millstone and Haddam Neck).
Corrective actions for generic design deficiencies beyond the scope of the FSAR inspections will be addressed by other staff actions, such as the Millstone lessons learned group reports.
5 l
Planned Improvements Complete or Short-Term linorovements 1.
On March 15, 1996, the short-term FSAR inspection guidance was extended indefinitely, pending a permanent change to the NRC In pction Manual.
j 2.
The staff will review NRC Inspection Manual Chapter 2515. " Light-Water Reactor Inspection Program - Operations Phase," and the operations, maintenance, and engineering core (required) inspection procedures and revise them as necessary to highlight the review and use of the FSAR implementation. Review of FSAR requirements will continue to be a part of future NRC inspections.
3.
The NRC staff will be reminded of the significance of including the FSAR in all inspection activities. This task will be accomplished through greater emphasis on the FSAR at Technical Training Division courses, Fundamentals of Inspection courses, and in upcoming counterpart meetings between headquarters and regional staff.
4.
The staff will resolve violations involving future FSAR discrepancies in accordance with the revised Enforcement Policy, once it is approved by the Commission. The Office of Enforcement will review Notices of Deviations, i.e., FSAR discrepancies which do not constitute violations, prior to issuance.
These steps should improve the consistency of the agency-wide treatment of FSAR discrep ncies.
Onaoina or lona-Term Improvements 1.
Most licensees with FSAR discrepancies have initiated corrective actions that range from performing routine FSAR updates to performing detailed reviews of their FSARs to determine the extent of inaccuracies. NRC regional offices will review the effectiveness of significant licensee corrective actions including the results of licensee FSAR reviews.
2.
The staff will selectively perform safety system functional inspections (SSFIs) at those siias with significant FSAR and 10 CFR 50.59 concerns and at those sites where more information is needed to determine the extent of FSAR and 10 CFR 50.59 implementaticn problems.
Conclusion On the basis of the limited (3-month duration, single-methodology) FSAR inspections and the staff's assessment of the significance of the identified j
discrepancies, the stat..us found few significant FSAR discrepancies.
j However, the staff ident (fied many minor problems and potential violations related to FSAR accuracy, design control, and 10 CFR 50.59 implementation.
i These results indicate that the staff must continue to focus on this area to verify that any significant programmatic problems are identified and corrected.
4
1 6
The staff will no longer compile FSAR discrepancy lists.
Instead, NRC regional office staffs will review and resolve individual FSAR discrepancies in accordance with approved enforcement guidance.
The staff is closely reviewing licensee corrective actions and will independently assess their effectiveness. SSFIs will be used, as appropriate, to aid in the identification and assessment of licensee's problems with their FSARs and 10 CFR 50.59 implementation.
When the short-term and long-term improvement actions are complete, FSAR review will be fully integrated into the normal inspection and enforcement processes.
Attachmeet: As stated cc:
\\
The staff will no longer compile FSAR discrepancy lists.
Instead, NRC regional office staffs will review and resolve individual FSAR discrepancies in accordance with approved enforcement guidance. The staff is closely reviewing licensee corrective actions and will independently assess their effectiveness.
SSFIs will be used, as appropriate, to aid in the identification and assessment of licensee's problems with their FSARs and 10 CFR 50.59 implementation.
When the short-term and long-term improvement actions are complete, FSAR review will be fully integrated into the normal inspection and enforcement processes.
Attachment:
As stated cc:
SECY OGC OCA OPA DISTRIBUTION:
CENTRAL FILES PIPB RF MRJohnson RWCooper, RI EWMerschoff, RII WLAxelson, RIII JEDyer, RIV TTD I
- SEE PREVIOUS CONCURRENCE I
DOCUMENT NAME: GASECY2JSARMEM7.COM ' Tills DOClTilNT WM RITIFWl'D llY Tile TI'CilNICAl. I:DITOR ON OR'05/06*
To receive a copy of this document, indicate in the box: 'C' = Copy without attachment!cnclosure *E" = Copy with attachment /caclosurc 'N' = No copy j
OFFICE PEAS:PIPB*
l SPSB:DSSA*
l DIR:0E*
l PIPB: DISP
- l DISP:NRR*
l NAME DLGamberoni EJButcher JLieberman RWBorchardt FPGillespie DATE 08/06/96 08/09/96/5 08/12/96 08/13/96 08/13/96 l
OFFICE DIR:NRR*
ED0sff f/l l
l l
NAME WTRussell JMTy' r DATE 08/30/96 9/#1/96
/ /96
/ /96
/ /96 y
0FFICIAL FILE COPY i
i I
I l
l l
l
Conclusion On the basis of the limited (3-month duration, single-methodology) FSAR inspections and the staff's assessment of the significance of the identified discrepancies, the staff has found few significant FSAR discrepancies.
However, the staff identified many minor problems and potential violations related to FSAR accuracy, design control, and 10 CFR 50.59 implementation.
These results indicate that the staff must continue to focus on th's' area to verify that any significant programmatic problems are identified nd corrected.
The staff of the NRC regional offices will review and reso e individual FSAR discrepancies in accordance with approved enforcement gu' ance.
The staff is closely reviewing licensee corrective actions and will ' dependently assess their effectiveness. SSFIs will be used, as appropri e, to aid in the identification and assessment of licensee's problem with their FSARs and 10 CFR 50.59 implementation.
When the short-term and long-term improvement a tions are complete, FSAR review will be fully integrated into the normal inspection and enforcement pree. esses.
The FSAR inspection resul'.s were also pr ided to the NRC Millstone lessons learned group for consid, ation in a br'ader context.
Attachment:
As stated cc:
SECY OGC OCA OPA DISTRIBUTION:
i CENTRAL FILES PIPB RF MRJohnson RWCooper, RI EWMerschoff, RII WLAxelson, RIII JEDyer, RIV TTD JTaylor EDO R/F
- SEE PREVIOUS CONCURRENCE DOCUMENT NAME: G:\\SECY2:FSARMEMS.COM 'A'IS DOCt%fENT WAS REVIEMTD IlY TIIE TirilNICAI, FDITOR ON 08/05%*
To receive a copy of this document, indicate in the box: *C" = Copy witimt attachment / enclosure *E' = Copy with attachment!cnclosure *N* = No copy 0FFICE PEAS:PIPB*
l SPSB:DSSA*
l DIR:0E*
l PIPB: DISP
- l OlSP:NRR*
l NAME DLGamberoni EJButcher JLieberman RWBorchardt FPGillespie DATE 08/06/96 08/09/96/96 08/12/96 08/13/96 08/13/96 0FFlesu. OIR:NRRf2 Sib' EDO l
l NAMF/IMTRusseil JMTaylor DATE /' f/go/96
/ /96
/ /96
/ /96
/ /96
~
OFFICIAL FILE COPY 1
a i
Continuous or Lono-Term Imorovements 1.
Most licensees with FSAR discrepancies have initiated corrective actions that range from performing routine FSAR updates to performing detailed reviews of their FSARs to determine the extent of inaccuracies. NRC regional offices will review the. effectiveness of significant licensee corrective actions including the results of licensee FSAR reviews.
2.
The staff will selectively perform safety system nctional inspections (SSFIs) at those sites with significant FSAR and 0 CFR 50.59 concerns and at those sites where more information is nee /ded to determine the j
exter+ of FSAR and 10 CFR 50.59 implementation' problems, f
Conclusion On the basis of the limited (3-month duration single-methodology) FSAR inspections and the staff's assessment of t significance of the identified i
discrepancies, the staff has found few sig ificant FSAR discrepancies.
However, the staff identified many minor roblems and potential violations j
related to FSAR accuracy, design control, and 10 CFR 50.59 implementation.
These results indicate that the staff st continue to focus on this area to verify that any significant programma c problems are identified and corrected.
/
1
/
The staff of the NRC regional offi es will review and resolve individual FSAR 1
discrepancies in accordance with pproved enforcement guidance.
The staff is closely reviewing licensee corr tive actions and will independently assess j
their effectiveness. SSFIs wil'l be used, as appropriate, to aid in the identification and assessment f licensee's problems with their FSARs and j
10 CFR 50.59 implementation.
When the short-term and lo g-term improvement actions are complete, FSAR review will be fully inte rated into the normal inspection and enforcement processes.
T The FSAR inspection r sults were also provided to the NRC Millstone lessons learned group for co sideration in a broader context.
DISTRIBUTION:
Attachment:
As ated CENTRAL FILES cc:
SECY PIPB RF OGC MRJohnson OCA OPA *SEE PREVIOUS' CONCURRENCE DOCUMENT nap 0:\\$ECY2fSARMEMS.COM *THIS DOCtMENT WAS RFVIEWED BY TIIE TITHNICAL EDITOR ON 08/05/96*
Ta receive a egy of this doement, indicate in the box: *C* = Cgy without att@nt/ enclosure *E* = Cgy we attachment / enclosure *N* = No cgy 0FFICE PEAS:PIPB*
l SPSR':14$A lc. DIRW-lC.PIPB/F SP lE DIfd!;NRR lC /
3 NAME DLGamberoni 0@ut(Rerv l D Jfifb6rman RWSdr4hardt FPGillespie a
/p6 e
e e
n n
9FF CE, DIR:NRR EDO bEC WTRussell JMTay1or DATE
/ /96
/ /96
/. /96
/ /96
/ /96 0FFIC]AL FILE COPY j
4.
NRC regional offices will resolve plant-specific FSAR discrepancies in accordance with approved enforcement guidance.
Continuous or Lono-Term Imorovements 1.
Most licensees with FSAR discrepancies have initiated correct ve actions that range from performing routine FSAR updates to performi detailed reviews of their FSARs to determine the extent of inaccura es. NRC regional affices will review the effectiveness ~ of signifi nt licensee corrective actions including the results of licensee FS reviews.
2.
The staff will selectively perform safety system func ional inspections (SSFIs) at those sites with significant FSAR and 10 FR 50.59 concerns and at those sites where more information is neede to determine the extent of FSAR and 10 CFR 50.59 implementation pr lems.
N Conclusion On the basis of the limited (3-mon h duration, sing e-methodology) FSAR inspections and the staff's assessmqnt of the sign ficance of the identified discrepancies, the staff has found f.w significa FSAR discrepancies; however, the staff identified many mi or proble and potential violations related to FSAR accuracy, design cont 1, and 1 CFR 50.59 implementation. On the basis of these results, the staff m st cort inue to focus on this area to verify that any significar.i. programmatic roMems are identified and corrected.
The staff of the NRC regional offices will view and resolve individual FSAR discrepancies in accordance with approved en rcement guidance. The staff is closely reviewing licensee corrective ac 'ons d will independently assess their effectiveness.
SSFIs will be use., as app opriate, to aid in the identification and assessment of licens e's probl s with their FSARs and 10 CFR 50.59 implementation.
When the short-term and long-term im ovement actions e complete, FSAR review will be fully integrated int the normal inspecti and enforcement processes.
The FSAR inspection results were 1so I,rovided to the NRC Mil 4 tone lessons learned group for consideration n a broader context.
DISTRIBUTION:
Attachment:
As stated CENTRA FILES cc:
SECY PIPB RF OGC MRJohnson x.'
OCA OPA DOCUMENT NAME: G:\\SECY2:FSARMEMS.COf1 'TIIIS DOCtMENT WAS REVIEWED BY TIIE ITCINCAl, EDITOR ON ORf05/96*
Ts aceive a egy of this document, indicate in the box: 'C' = Cgy without attachmeraienclosure "E* = Cwy with attachment / enclosure *N' = N6 cgy 0FFICE PEAS:PIPB 31A.lE PIPB:0lSP l
SPSB:DSSA l
DIR: DISP l
DIR:0E l
NAME OLGambero'nis RWBorchardt EJButcher FPGillespie JLieberman DATE 6//o/96
/ /96
/ /96
/ /96
/ /96 0FFICE DIR:NRR EDO NAME WTRussell JMTaylor DATE
/ /96
/ /96
/ /96
/ /96
/ /96 1
0FFIC)AL FILE COPY
FSAR ISSUES
SUMMARY
The following table was developed following performance of special inspection instructions provided to inspectors regarding the review of the final safety analysis report (FSAR) during preparation and performance of routine inspection activities between January 25 and April 26, 1996. The notes listed below provide descriptive text associated with identification of the issues, the types of issues identified, whether the discrepancy may result in a change in the plant or the way the plant is operated, and the type of followup activities described in the inspection report documenting the FSAR reviews.
NOTES i
- 1. How the issue was identified (Identified):
NRC - identified by the NRC during inspection LIC - identified by the licensee
- 2. Type of issue identified (Category):
l DESIGN - discrepancy between the plant and the design as described in the FSAR.
PROC /0PS - discrepancy between a plant procedure or operation and the FSAR.
EDIT - minor editorial discrepancy or administrative problem with the FSAR.
- 3. The discrepancy may result in a change in the plant or in the way the plant is operated (Chg Op/Eq):
l Yes No Unk - Unknown
- 4. Followup described in inspection report (Followup):
EEI - violation considered for escalated enforcement IFI - inspector followup item I
N/A - no followup required NCV - noncited violation NRR - assistance being requested from NRR URI - unresolved item VIO - violation DEV - deviation no entry - follow up not discussed in inspection report 1
ATTACHMENT 7
FSAR l=ues Summ:ry 27-Ag-w Region Site Ref No Date issue Identified By Category Chg Op/Eq Followsp AdditionalInformation I
Beaver.
%-02 3/19/96 Licenscc identified need to revisc cmcrgency diesel LIC Design Yes N/A Valley generator shutdown procedures or modify the shutdmvn circuitry in order to meet their UFSAR commitment to IEEE Standard 387-1972 for a scenario which could occur within 140 seconds of dicsci shutdown. Mod scheduled next outage.
96-02 3/19/96 The inspectors found control room habitability UFSAR NRC Design No NRR design requirements difficult to interpret. Although the emergency breathing air supply was being maintained in accordance with the UFSAR, the length of time breathing air needs to be supplied was identified as a design basis question the licensee will resolve with NRC.
96 4/19/96 Unit i UFSAR has a shorter rod drop acceptance criteria NRC Edit.
No (2.2 seconds) than does Technical Specifications (2.7 seconds - changed in 1989 amendment) and the Unit 1 UFSAR also states that rod drop testing is donc at no flow cold conditions, contrary to actual practice.
96-03 4/19/96 Unit 2 UFSAR described a separate Manager of NRC Er'it.
No
/
Operations for cach unit and a General Manager Nuclear Operations. Since the last annual UFSAR update. DLC has combined these three positions into one with the titic of General Manager Nuclear Operations. UFSAR resision planned.
96-03 4/19/96 Chemical treatment of the '1 nit 2 dicsci coolers is different NRC Proc /
Unk than UFSAR description. UFSAR describes adding Ops chemicals to the cmcrgency dicsci generator coolers for wet layup from a chemical addition tank to prevent undue corrosion. However. since initial unit startup, it has not been the licensec's practice to place the dicsci heal exchangers in chemical urt layup other than with chemical injection points prmided for controlling algac.
macro invertebrate growth. and silt deposition, and corrosion inhibitors were only recently added.
FitzPatrick 96-02 5/3/96 Use ofIIPCI & RCIC in pressure control mode as RCS NRC Proc /
No N/A depressurization method is not described in FSAR. FSAR Ops to be resised.
1
Itegion Site Ref Na Cate - lesse Identified Ey Category ChgOp/Eg Follownp AdditionalInformation I
Ginna 96-01 The licensee has never exercised the ' Assessment Facility -
.NRC Proc /
Unk-VIO Issueidentified by NRC during
- (backup lab for analyzing PASS samples) for analyzing Ops EP inspection (96-01) 3/1l-both onsite and offsite samples during an emergency.
14/96. Licensee willdevelop Also, the licensee does not have procedures for using this procedures for the Assessment facility as a radiological laboratory. Violation of 50.47(b).
Facility and exercise handling samples.
%01 ' 5/8/96 UFSAR did riot prmide any acceptance criteria for air-NRC.
Proc /
Yes URI cleaning system ventilation tests (control room & aux bldg Ops ventilation systems & plant vent), and the listed values for airf1mv capacity were vague relatist to design
. requirements. UFSAR to be resiewed & updated.
c
% 01 5/8/96 Emergency response plan states Plant Operations Review NRC Proc /
Yes Committec (PORC) is responsible for evaluating plant Ops conditions, resiewing decontamination activitics and
}
necessary repairs prior to giving approval for plant j
I reentry. The inspectors resiewed the PORC charter and noted that it did not mention their recmtry phase responsibilitics. Emergency plan also states members of
[
the recovery organization will be given recovery training annually. Licensee could not verify all PORC members had received this training. PORC charter to be resised.
t
- IIaddam 95-27 t/25/96 An apparent siolation exists for a condition in which LPSI LIC Design Yes EEI NRC notified by licensec Neck system flowrate is nonconservative in relationship to 12/13/95. System engineer had -, l N
assumptions in the accident analysis. Related performance noticed low flow since 1993.
}
issues are the lack of LPSI design basis testing and errors OE awaiting NRR team report i
in the supporting analytical calculations for LPSI design on issue prior to issuing l
basis flowrates. Based on review of this issue, one enforcement action. Correctist insenice test procedure for LPSI substantial flow testing actions: Operability confirmed, did not bound the assumptions used for LPSI design safety assessment undertaken.
flourate. Nonconservative accident analysis assumption 50.72 notification made,
{
crrors have occurred in the past, that indicate inadequate additional testing perfornwd.
completeness in resiews_
Licensec looking at modifying LPSI system to increase flow.
confirming testing. and reviewing surevcillance procedure.
6 l
l 3
2
Region Sita Ref My Date-Isene Identified Cy Category Chg Op/Eg Followup AdditionalInferneation j
i Haddam 95-27 1/25/96 UFSAR states that containment narrow ringe sump level NRC Design Unk IFI
- Neck -
monitoring system and containment gaseous and particulate monitoring systems can detect a one gpm leakrate within one hour. The inspector took into account
. instrument accuracy of the containment narrow range !
j sump level indicator, the low and high level alarms on the control board annunciator, and the frequency of operators i
recording the sump level indication (estry 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />), and concluded that operators are prmided no means to alert them of a one gpm leakrate within four hours wheat considering containment narrow range sump lestl t
indicator as irAysAnt from the other parameters to measure reactor coolant system leakage. This conclusion.
and the inconsistences beturen the UFS AR and a TS amendment concerning a sensitivity description of the leakrate monitoring system will be evaluated in future inspections.
96-01 3/11/96 Scnice Water temperature lower than UFSAR design Lic.
f., sign No URI basis. During resiew of proposed TS change for containment air recire fans noted that plant was operating outside design basis due to cold SW temperatures.
l UFSAR assumes minimum ultimate heat sink temperaturc of 35 F. River temperatures as low as 29 F have been noted in the past. Licensec evaluated the condition and operation was qualitatively evaluated for a new SW supply j
temperature of 28 F.
%-01 3/11/% Failure to mcct Operating License DPR-61 Condition #4 Lie Design Yes VIO The basis for the violation is Fire Protection, in the failure to provide a combustible gas that the licenscc was ncycr able I
detection system for the chemistry laboratory. Although a to get the detection system to combustible gas dctcction system was installed during the function properly. Issuc known j
1980 refueling outage, it uns not turned over to operations to NRC & licensec since late j
for usc. The licensee identified the failure to meet License 1970s. In 1985 licensec Condition #4 by letter to the regional administrator dated committed to perform ucekly August 30.1985. along with a plan to make the system che ks for flammable gasses in i
operabic. As of February 20,1996. the combustible gas the lab. Plant maintenance and detection system had not been turned over to operations engineering personnel repaired for use.
and calibrated the system. The
)
system was released to operations for unrestricted use on 2/28/96 i
3
Region Siu Ref No Cate '
Issue Identified Cy Category Chg Op/Eq Followup AdditionalInformation I
Haddam 96-01 3/11/96 NRC inspection report 95-27 described the licensec Lic Design No URI Neck identification of a long standing condition in which LPSI system flow rate was less inan assumed in the accident analysis. This condition n as reported in LER 95-22.
50.72 3/11/96 Cc nmenced TS plant shutdown due to 48 containmcrt LIC Proc /
No N/A Notificat isolation vah es not being verified closed. The vahrs are Ops ion minor instrumentation vahrs or drain vahrs in MS or AFW systems. Stopped rampdown a: 94% Later -
recommenced rampdown when 8 add'l vahrs utre identified. These vahrs are MS PORV sent vahts.
Stopped rampdown at 79% when these vahrs werc verified closed and utre added 1o containment isolation suntil!ance procedure.
96-02 4/26/96 Two instances of setpoint crrors concerning high NRC Proc /
Unk containment pressure isolation of control room ventilation, Ops 1
and annunciator audibic alarm at 4 psig abo r atmospheric pressure in UFSAR.
96-02 4/26/96 Two service water valves (SW-V-852 and SW-V-853) arc NRC Edit.
No listed in UFSAR tabic, but were not installed in plant nor described in plant documentation.
96-02 4/26/96 UFSAR Table 3.9-1 was more prescripthe than technical NRC Edit.
No i
specification table 5.71 on limiting transients for reacic ;
vessel fatigue analysis.
96-02 4/26/96 Operating practices. procedures and logs for PRT pressurc NRC Edit.
No
& temperature ranges difTer from UFSAR.
96-02 4/26/96 UFSAR table orcontainment isolation valves did hot list LIC Proc /
Unk URI filly six main steam line vent. drain. test. and auxiliary Ops feedwater floupath valves.
96-02 4/26/96 Licensee team identified a large number of discrepancies LIC Edit.
Unk beturen UFSAR and operating practices, ranging in significance from minor editorial changes, to more significant changes needed to assurc UFSAR cicarly
- 'lected the current plant design (e.g. the updated nuclear instrumentation system). These findings indicate an apparent weakness in licensee program to update UFSAR and/or assure consistency between operating practices and the licensing basis.
4
Region Sit 2 RefN2 Date Issue '
Identified Cy Category Chg Op/Eg Followup AdditionalInfo:wistion I
Hope Creek %-80 4/24/96 Discrepancies noted in safety rel ted bettery electrol 1e LIC Design
'Unk URI 3
RATI temperature requirements between the UFSAR, the TS and design load calculations.
96-80 ; 4/24/96 Minor organization and qualification discrepancies noted NRC Edit.
No when comparing the actual organization with that described in UFSAR.
96-03 4/26/96 Procedures that fulfill technical specification required LIC Proc /
Unk surveillance testing of drywell-to-torus vacuum breakers Ops did not impicment a "one hour hold
- requirement before test commencement after initial test conditions utre established, contrary to the FSAR description.
%-03 4/26/96 Reactor core isolation cooling and high pressure coolant LIC Proc /
Unk injection system test procedures did not verify automatic Ops operation of all the system valves required by test description in the FSAR.
%-03 4/26/96 Fiftcen pairs of reactor building backdraft isolation NRC Design Yes EEI Damper probicm first dampers installed in IIVAC supply doctwork were documented in inspection installed in a reverse orientation such that protection of report 92-02 as open item important-to-safety equipment following a high cncrgy which was closed later in 1992 line break was not adequately demonstrated.
after analysis and promisc to correct. Issue resurfaced 2/96 with LER stating dampers were still installed backwards.
Dampers reversed 3/96.
Enforcement conference scheduled for 6/II/96.
96-03 4/26/96 Full core ofiloads had been conducted during refueling NRC Design Unk URI outages 3 and 4. It was noted that the operation was accomplished such that design heat rejection rates of the normal spent fuel pool cooling system ucre not exceeded.
Ilowever. it was also found that certain sections of the FSAR indicate that the heat load calculations for the SFP cooling system assumed that core shuffling would occur during refueling outages 3 and 4. NRC is resiewing the i
acceptability of full core off-load on a generic basis.
5
Region Site Ref No Cate Issue' identified Cy Category Chg Op/Eg Followup AdditionalInfernia lon I
Hope Creek 9643 4/26/96 Engineering personnel determined that, folkming o NRC/ LIC Design Yes EEI First known to NRC & licensec station senice water system flow balance, that flow to the during this inspection (2/96),
safety auxiliaries cooling system had been insufficient to Further analysis & f1mv meet the post-accident design criteria specified in the balancing of SW system FSAR since initial plant operation.
completed. Licensceplansto conduct integrated SW system inspection.
96-03 4/26/96 25 containment penetration isolation devices found that LIC Edit.
No are not verified closed by station operating procedure nor listed in TS.
96-03 4/26/96 Procedure that governs control rod speed measurement NRC Design Yes EEI NRC discovered & notified and adjustment did not present cccident analysis licensee ofissue during 3/96 assumptions for a continuous rod withdrawal accident inspection. Rods utre tested 3
during reactor startup. FSAR section which analyzcs rod satisfactorily. Additional withdrawal malfunctions lists a maximum rod speed of 6 actions pending upcoming ips with the speed control vahr failed fully open. This enforcement conference.
speed corresponds to a full stroke time of 24 seconds. Yet, at least six of the cicten rods (some with directional control valves replaced) resiewed exceeded this maximum speed. These speeds should not have been possible since they represented the theoretical maximum speeds achievabic under norst case conditions. The licensec reported the occurrence of excessisc rod speed as a condition outside the design basis per 10 CFR 50.73.
Regarding the excessive control rod speeds (greater than 5 ips). the inspector noted that the May 10,1992 test of rod 22-35 resulted in four results with rod speeds at or above 5 ips. Yet, the licensee restored the rod to an operable status and continued to operate for five more months until the rod was tested during the fourth refueling outage (RFO-4). Full travel stroke time was 20.9 seconds u hich execeded both the expanded GE limit, and the FSAR worst case limit. The licensec did not address this as a condition
^
outside the design basis, and toolr no corrective action at the time. This constitutes an apparent violation of 10 CIP 50 Appendix B Criterion XVI. Correctist Action.
i i
6
Region Site Ref N3 Date issue Identified By Category Chg Op/Eg Followsp AdditionalInformatioit I
Hope Creek %-03 4/26/96 Concern regarding operating procedures that impicment NRC Proc /
Unk URI TS suntillance testing of Automatic Depressurization Ops System vahes during reactor startup. Licensec changed the description of the test in the FSAR (to accommodate vahr vendor recommendations and current operating practice) in a manner which appears to conflict with the intent ofTS requirements.
96-03 4/26/96 Station personnel discovered that dryntil cooling fans LIC Proc /
Unk have been routinely operated (in compliance with Ops operating procedures) in a manner inconsistent with their characterization in FSAR.
Indian Point 96-01 3/27/96 An inconsistency was identified between the plant NRC Proc /
No URI 2
radiation zone descriptions in the UFSAR and actual plant Ops radiation levels. Radiation dose rates in several plant radiation zones utre in excess of the values specified in the UFSAR.
3/27/96 UFSAR lists minimum senice water design temperature NRC Design No 96-01 as 35 F. Louct river temperatures have occurred this winter. To be addressed in a future UFSAR update.
Indian Point 96-01 3/27/96 Waste gas system is described in FS AR as an automatic LIC Proc /
No URI 3
system. The uuste gas system currently is operated Ops manually per SOP-WDS-2. NYPA had presiously r
ideniHicd this discrepancy and a NSE was approved on 2/21/96 which detennined that manual operation was acceptabic. Operation of the waste gas system urnt from automatic to manual operation aller the retirement of the waste evaporator in the mid-1980s and procedure SOP-WDS-2 was changed accordingly. The inspector noted that a 50.59 cvaluation was not completed at that time.
3/27/96 While imestigating oxygen intrusion into the waste gas LIC Proc /
Yes URI 96-01 system, licensec identified that 2 vent header containment Ops isolation vahrs utre in positions contrary to Ihose specified in the FSAR. Resuh ofinadequate 50.59 review of procedure change.
96-01 3/27/96 Discrepancy noted between the FS AR and the plant's 3 RC Proc /
No-cmcrgency operating procedures (EOPs) regarding the Ops termination of NaOli addition.
7
Stegie Site Itet No Date ~ Issue Identitled Cy Category Chg Op/Eg Followup AdditiemalInfonnation i
I Indian Point 96-01 3/27/96 The previous three extmples and other recent occurrences.
LIC -
Proc /
No URI 3
such as the liRing of the CCW relief valve indicate that Ops t
plant procedures nwy not have been consistently and -
adequately evaluated agains; the FSAR as required by 10 CFR $0.59. This issue is leR unresolved pending further NYPA evaluation and NRC resiew to ensure plant pim.Jes are reflectiw of the licensing basis, 9641 3/27/96 Procedure SOP-EL-5 presided instructions for cross-LIC Proc /
No URI connecting 480 y buses while transferring offsite power Ops ~
sources. Performing this procedure uvuld have resulted in both RHR pumps being powered from the same, cross-connected buses. NYPA determined that further cvaluation was required prior to performing this procedure. NRC resier noted that the FSAR described the RHR system as having redundant components. and further stated that equipment was arranged electrically so
[i that multipic items reccised their power from difTerent sources..
l Limerick
%-01 3/22/96 UFSAR states a formal ALARA review is conducted every NRC Proc /
No three years by the Nuclear Resiew Board. By resiew of Ops licensec records, the most recent formal ALARA resiew
(
. conducted by the NRB was dated October 26.1992. This ALARA reticw was conducted by the Limerick Nuclear s
Quality Assurance Group and acsicued by the Nuclear
\\
Resiew Board. The Limerick Nuclear Quality Assurance Group performed another ALARA program resiew in March of 1994. In the audit report introduction section, it states. " Health Physics Operations /ALARA is not a Tech.
i Spec. Assessment therefore NRB concurrence was not l
solicited." The inspector determined that the intent of the -
l UFSAR had been met by performance of a biennial ALARA reticu; however, the perspectim of the NRB was not obtained. The Rad Engineering staff wrote an action request to evaluate the appropriateness of the UFSAR requirement.
i i
f t
i 8
?
=
l l
l Region Site Ref No Date Issue Identified C'y Category Chg Op/Eq Followsp AdditionalInformation I
Limerick 96-01 3/22/96 UFSAR states, "Two separate shutdown cooling pump sad NRC Proc /
Yes VIO Licensec & NRC aware ofissue heat exchanger loops are prmided," and, " Inter-tics are Ops when identified byinspectors prmided between the suction and discharge lines of the during report period. TS -
RHR pump in the direct injection LPCI loop (C and D change submitted;in the pumps) and the associated RiiR pump in the heat meantime will comply with TS.
exchanger loop (A and B pumps, respectively) to allmy use of the C and D pumps in the shutdown cooling mode, thus providing greater maintenance ficxibility," Inspectors concluded that the LGS interpretation of the SDC loops, with four possible, was incorrect in that only two loops, were possibic. Violation for draining down with only one RHR SDC mode loop operable.
96-01 3/22/96 UFSAR analysis for decay heat removal from the spent NRC Edit.
No fuel pool assumes ofiloading one third of the core, and one and one half year cycles. Ilowever, both plants arc on tuu year cycles and typically more than one third of the core is changed out. UFSAR has not yet been updated to reficct the changes.
Mainc 96-01 4/2/96 Latest core performance analysis repr :. credits SG NRC Proc /
Yes Yankee blowdown isolation on SG low level for mitigating Ops consequences of a loss of feedwater event (an FSAR design basis event). The recently installed SG blowdown isolation system is designed to safety-related requirements, but is not covered by TS. Sysicm is controlled by administrative procedures. Licensec proposed to add the blondown system isolation valves to TS when the issue was raised during the inspection. StalTconsiders the equipment adequate for credit in FSAR safety analysis. Preparing TS amendment.
Millstonc 96-01 4/12/96 FSAR table outlining manual operator actions to align the NRC Edit.
No ECCS from injection to the cold leg recire mode, was not up to date in detailing and sequencing EOP and " Cold Leg Recire Array" steps.
96-01 4/12/96 FSAR dit:repancy in the description of" safety grade cold NRC Proc /
Unk URI shutdown" (SGCS) requirements, implying that Ril5 Ops initiation. rather than the required cold-shutdown conditions, was the desired SGCS cndpoint. The licensec recognized this discrepancy will resise FSAR.
9
. f Itegion Site Itef No Date. Issue Identitled By Category Chg Op/Eg Fellery Additteest Informistion
}
I Millstone 96-01 4/12/96. Hydrogen snonitor operability concern. Containment air NRC Design Unk URI enters the hydrogen monitor through pressure regulators that are set for 10 psig. No iwds checks that the 1
regulator is set for 10 psig and it has not been checked since pon-installation testing of the hydrogen monitors.
l FSAR states that the hydrogen analyzers will not oc subjected to containment pressure, utilizing a pressure regulator in cach sample line to limit sampic pressure to less than 5 psig. The licensee failed to recognize that the I
both trains of the hydrogen monitors have always been j
inoperabic because the surveillance procedure was not written to duplicate, as close a practicabic, the pon-accident conditions in which the equipment would be required to function. The failure to test safety-related systems in this manner has been a recurrent problem at Millstone and has been the subject of presious violations.
% 01 4/12/96 FSAR is inaccurate with respect to design bases for time -
NRC Design Unk requirements for initiation of containment hydrogen monitors and Post Accident Sampling System.
96-01 4/12/96 Number of staw ; %hutdown cycles exceeded the number NRC Design -
No of cycles indicated in the Unit i UFSAR. Tabic 3.9 1
[
states that the number of design cycles for the facility licensed lifetime is 120 cycles. ' To date, the plant has experienced 126 startup/ shutdown cycles. The licensec now estimates that the plant will experience 170 cycles over the facility licensed lifetime. The number of cycles attained has been found by the licensec to l. acceptable. in that total cycles will not result in cumulative cycle usage factors greater than 1.0 over the operating lifttimd of the plant. UFSAR to be revised.
Nine Mile 96-05 3/29/96 Inconsistencies within and between the UFSAR and the NRC Design No IFl Point Individual Plant Examination intohing the value of the
{
' pressure relief capabilitics of the blowout pancis. Also.
i inconsistencies within the UFSAR regarding the design l
basis for the blowout pancis and specific high energy line i
breaks.
i e
[
w t
Region Siu Ref N) Cate issue Identified Cy Category Chg Op/Eq Follownp AdditionalInformation i
Nine Mile 96-01 4/22/96 Licensec installed an emergency temporary mod which NRC Proc /
No VIO Licensec & NRC aware ofissue Point changed design of the Unit 2 circulating water system, as Ops when it was identified by the described in the U MR, prior to the completion of the inspectorsin 2/96. Licensee written safety evaluauon. NMPC common procedure GAP-completed 50.59, within 2 days DES-03, " Control of Temporary Modifications," Revision ofidentification ofissue.
4, allows emergency temporary mods to be installed prior Procedure was revised to to completion of required written 50.59 cvaluations.
remmc prmision for temporary mods w/o prior written safety eval.
Oyster Creek 96-02 4/17/96 An operating procedure provision allowed for operating NRC Proc /
Unk the standby gas treatment system for a purpose other than Ops described in the UFSAR. The specific provision and operational mode uns found to be acceptable, however, the procedure bases did not specifically evaluate and document the prmision. Licensec performed evaluation and determined that operation of the SGTS in thir. manner was no different from normal sysicm running, including during surveillance testing.
Peach 96-01 3/25/96 Inconsistency between the Security Plan and plant NRC Proc /
Yes N/A Bottom practices found relative to de-vitalization of certain areas Ops during outages. PECO revised the Plan to correct the inconsistency.
96-01 3/25/96 Periodic inspections of Borallex coupons in the spent fuct LIC Proc /
Yes DEV pool (SFP) have not been completed for either unit. An Ops engineering evaluation of the SFP determined that the integrity of the Borallex was good. The inspector concurred that the SFP was in a safe condition due to other existing sturcillance methodologics; however the testing required by the UFSAR was not tracked well in that the inspection was not performed in the last ten years.
%-01 3/25/96 Rcsiew of spent fuci pool cooling system design NRC Proc /
Yes N/A documentation descrmined system design heat load was Ops cicarly defined. Inspecte noted, however. that it was not cicar that the refueling procedure provided adequate controls to ensure SFP cooling system design requirements would be maintained during a full core ofiload. Procedure change now precludes fuel movement for 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown. Licensec evaluating issue.
11
~
~..
Region Site Ref Na Cate Issue Identified Cy Category Chg Op/Eg Followup AdditionalInformation i
Peach 96-01. 3/25/96 No discussion of the titeration of the circulating water NRC Edit.
No N/A Bottom discharge flow path in the UFSAR. Testing requirements for the alternate shutdown panels not documented in the UFSAR or an associated reference document.
96-01 3/25/96 PECO was tW testing to ensure that the degraded grid UV NRC Proc /
Yes VIO Licensee & NRC awarc ofissue f
relays on 4 KV safety-related buses functioned within the Ops VIO when identified byinspectors TS required settings (UFSAR specified testing). PECO during report period. Licensee 1
had not been testing to within the TS allowabic values r-~ised test procedure.
l since the relays had been installed in 1989. Further, PECO had not been treating as-found calibration data t
outside the TS allowable values as insttument failures.
The inspector found that the overall safety significance of this issue was low since, although not known cicarly to PECO at the time, the setpoints used were within the calculated analytical values. Howcwr, violations of 10 CFR 50, Appendix B, Criterion XI, Test Controls and Criterion XVI, Corrective Actions, utre cited because the testmg was inadequate to verify operability and the j
calibration testing was inadequate to identify an adverse '
j condition dealing with as-found settings.
Pilgrim 96-01 3/27/96 Groundunter inicakage to torus room not addressed in NRC Design No N/A UFSAR. 50.59 & UFSAR change initiated.
Salem 96-01 3/25/96 On 2/14/96 workers installed temporaryjumpers in the NRC Proc /
Yes VIO Licensec & NRC aware ofissue l
cncrgized 125VDC control circuit for Salem Unit I vital Ops when discovered by inspectors i
bus iB without adequately determining if thejumpers during inspection period in t
modified the plant as described in the UFSAR (50.59 January 96. Licensec Violation).
performed 50.59 and changed procedure.
i 96-01 3/25/96 UFSAR states spent fuel pool cooling is designed with the NRC Proc /
No URI
~
capability to remove heat from a full core discharge.
Ops UFSAR also states that a typical core off-load consists of about one-third of the core. UFSAR describes a full core discharge as unusual circumstances. however, the Salem j
Units typically perform complete core oft-loads during
[
refueling outages. This issue remains unresolved pending further inspection.
12 i
Region Site Ref No Date Issue Identified Ey Categen Chg Op/Eq Followup AdditionalInfonnstion 1
Salem
%-01 3/25/96 Salem does not operate the EDGjacket water cooling NRC Proc /
Yes VIO Licensec A NRC aware ofissue system as described in the UFSAR. Salem staff did not Ops when discovered byinspectors evaluate the change in system operation to determine ifit duringinspcction period in constituted an unreviewed safety question (50.59 January %. Pending response violation). UFSAR states that a ball-float valve controls to violation, licensee utote the makeup uater flow from the dcmineralized water 50.59 to change procedure.
system to thejacket water expansion tank. Contrary to this, operators maintain the demineralized water system isolated and manually make up to the expansion tank as necessary.
26-05 4/22/96 Items from spent fuct pool inspection: 1. Licensec has not NRC Design Unk analyzed spent fuci pool structures and associated systems for boiling. 2. No procedure for using the cross connect between the heat exchangers to support the onc unit with a SFP cxcess heat load. 3. No procedure controls in place that assure that the SFP heat load is maintained below the analyzed value.
Scabrook 96-80 4/3/96 Instances where procedures do not conform with UFSAR LIC Proc /
Unk IPAP requirements relating to the NUREG-0737, item Ill.D.l.1 Ops requirement for a program to reduce leakage from systems outside the containment that could contain highly radioactisc fluid following an esent. Specifically, UFSAR-
/
(Section 1.9) requires: a hand-over-hand type visual walkdown uhile the subject system is in operation (usually during a pump test), work request numbers initiated and recorded on data sheets when leakage is found and the flydrogen Dctcction subsystem of the Combustible Gas Control System. including sample lines for post-accident gas samples. to be included in the scope of the leakage reduction program and to be tested using helium detection techniques. Contrary to these requirements. the Leakage Reduction Program Procedure (EX1801.002): provides the option to perform the inspection after the system has been in operation. does not require recording work request numbers on the data sheets, and sta'es the Combustible Gas Control S stem is excluded from the Leakage 3
Reduction Program and is tested by procedurc EX1801.003. which tests the system using air vice helium.
13
i Region Site RefNJ Date Issue Identified Cy Category Chg Op/Eg Followup AdditionalInformation i
Scabrook 96-80 4/3/96 UFSAR should be considered when upgrading NRC Proc /
Yes IPAP procedures, but discussions with procedure writers Ops indicated a lack of emphasis in this area. Also, a werd search program on the computer for writers to research regulatory, UFSAR, NRC and other commitments does not i
seem to be user friendly. Licensec plans to change to a new innd more powerful program that uvuld improve werd scarch capability.
96-80 4/3/96 UFSAR states cach starting air system is capabic of NRC Proc /
Unk IPAP starting a dicsci generator within 10 seconds at Icast rive Ops titnes without recharging the air receiver. During plant startup, the dicscis were tested to meet five starts from an j
initial air pressure of 560 psig. The starting air compressors were set to start at 560 psig, which met the i
UFSAR intent. Ilowever, the low pressure alarm setpoint for the receivers was 460 psig.100 psig below the design basis value. The low alarm is an early warning to the operators of compressor failure. Possibility exists of having the dicscis in a condition where the five start basis could not be met.
Susquehanna 96-03 4/3/96 Regarding the standby gas treatment and reactor building NRC Proc /
Yes URI recirculation systems, use of probabilistic analysis as a Ops measure of compliance with Ihe sing!c failure criterion of 10 CFR 50.55a(h), and the siting criteria of 10 CFR part
\\
100 as a standard of system operability was unresolved pending further NRC review. An additional unresolved item concerned performance of 50.59 cvaluations for longstanding degraded or nonconforming conditions.
96-03 4/3/96 Apparent violations involving insufficient attention to the NRC/ LIC Proc /
Yes VIO Ucensce & NRC aware of plant licensing basis in operability assessments and failure Ops.
URI issues uhen identified during to identify and correct design deficiencies in a timely Design VIO reporting period. Actions for manner: (1) an engineering evaluation of reactor unter first violation: revised cicanup leak detcetion system capabihty did not reconcile operability assessment &
conflicts between system sensitnity and licensing basis providing training (in requirementst (2) longstanding single failure progress). '.ctions for second vulnerabilitics involving the standby gas treatment and violation: reperformed
{
reactor building recirculation systems were not cotrccted operability assessment.
in a timely manner; (3) scismic-monitoring instruments relocating some scismic were installed contrary to technical specification (and instruments (in progress).
FSAR) location requirements.
FSAR to be changed.
t 14
. ~
Region Site Ref Na Cate Issue Identified Cy Category Chg Op/Eg Followsp AdditionalInformation 1
Susquchanna 96-01 4/9/96 FSAR requires reactor building ventilation system t)
NRC Design Unk N/A maintain air flow from areas oflesscr contamination to areas of greater potential contamination. Contamination found outside CRDM room due to loss of negative pressure in room caused by dirty exhaust lousers and an almost closed exhaust damper. Licensec evaluating.
96-01 4/9/96 During a design basis accident condition, the water scals LIC Design Unk IFI for the feedwater lines as described in the FSAR may not be achievable due to past high FW vah c leakage during LLRTs. As the past as-found Icakage results excceded the criteria for maintaining off-site doses within the regulatory limit, the licensec reported the condition to the NRC.
Recent modifications made to the valves decreased leakage, and acceptable test data from recent outages allowed continued plant operation. The licensee is resiewing various options for long term corrective anions.
TMI 96-01 3/15/96 Inspectors found that minimum ambient air temperature at NRC Design No N/A the river water intake structure exceeds UFSAR minimum of 60 F. Rcsiew of the auxiliary operator log readings for the two cicctrical motor control centre and pump bays revealed that the lowest ambient teniperatures were 52 F and 56 F respectively. UFSAR to be revised based en results of engineering evaluation.
r 11 13rown's 96-03 4/15/96 Several sections of the UFSAR associated 9th clectrical NRC Edit.
No Ferry systems have not been updated to reflect the rett.m arUint 3 to pourt operations.
96-03 4/15/96 UFSAR states that fuel pool high or low levels will actuate NRC Edit.
No i
alarms in the control room. The inspector determined that only low fucI pool level will actuate an alarm in the control room.
96-03 4/15/96 The licensec's QA management directed that a detailed LIC Edit.
No resiew of applicable portions of the UFSAR be included in assessment activitics. This has resulted in one UFSAR inaccuracy being identified as wcll as several areas that should be clarified or enhanced. Several doors between the reactor building and the control building were not being used "for emergency usc" as stated in the UFSAR.
Corrective actions were initiated.
15
Region Siu Ref N2 Date Issue Identified Cy Category Chg Op/Eg Followup AdditionalInfornention 11 Broula's
. %-03 4/15/96 UFSAR crroncously states that refueling takes place on In NRC Edit.
No Ferry approximate annual basis, when in fact refueling takes place about every 18 months.
Catauba 96-05 FSAR does not include any description of the containment NRC Edit.
No Hydrogen Mitigation System (Igniters). The system is included in TS Section 3.6, Containment Systems.
96-02 4/22/96 FSAR figure included incorrect vahr locations for vahrs NRC Edit.
No ISV027A and ISV028A. Vah es werc shown to be associated with incorrect steam generators.
t 96-02, 4/22/96 FSAR describes the refueling trolley and hoist and NRC Edit.
Unk references Rod Control Cluster (RCC) mast / handling desices. The RCC was removed on Unit 1 in 1995. FSAR descriptions ofinterlocks are not valid for Unit 1.
%-02 4/22/96 FSAR described the design of the Spent Fuct Pool (SFP)
NRC Design Unk VIO Identified by licensee pre-and stated that no connections would result in inadvertent inspection scif assessment week draining of the SFP below a Icycl of ten Icct above the of 3/4/96. FSAR being resised.
i racked fuct assemblics. The safe shutdown system interface to provide reactor coolant pump seal water could
. permit draining of SFP to top of racked fuel assemblics.
96-02 4/22/96 FSAR described the refueling bridge trolley and stated that NRC Design Unk raising of fuct assemblics was limited by a limit switch and mechanical stop to prevent raising fuct above a level required for shiciding (10 feet of water above the fuct t
assembly). There was a limit switch, but no mechanical T
stop was presided.
~
96-02 4/22/96 FSAR describes efTect on the SFP at maximum heat Ivad NRC Design Unk of a Sarc Shutdown event due to boil oIT and reactor coolant pump seal unter supply. This description is not valid because with a full core offload required by the maximum decay heat load. RCP scal water was not required. Lcret loss in this condition would be due to boil ofTonly. In the normal heat load condition, time to j
boiling would be dilTerent than the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> stated.
i 16
Region Site Ref Na Cate ' - Issue Identified By Category Chg Op/Eg Followup AdditionalInfernistion Il Catauba
%-02 4/22/96 Unit vent monitors contin charcoal elements. FSAR NRC' Design No VIO First known tolicensee & NRC states elements contain sihcr zeolite whenidentified byinspectors 3/1/96. Licensee regiening proceden:s, calculations, A collecting efficiency resicus of specificissue and performing complete FSAR resiew to identify and change all discrepancies.
%-02 4/22/96 FSAR described the SFP loading conditions for normal NRC Proc /-
Unk and maximum decay heat loads and included the criteria Ops of a 7-day decay time before a full or one-third core ofiload. This decay time was included as an assumption in the decay heat load analysis for the loading conditions.
No administrative controls assure this 7-day criteria is met.' Licensec records indicate the criteria had not been exceeded.
Crystal River 95-21 2/26/96 A weakness was identified for failure to maintain the NRC Proc /
No FSAR to be resised FSAR accurate for the engineered safeguards closurc Ops system for the containment purge vahrs.
95-21 2/26/96 FSAR states that cIcctrical systems satisfy the criteria of LIC Design Yes VIO NRC informed via Licensec sufficient physical separation. cicctrical isolation, and probicm rpt.1/25/96 Lic.
redundancy to prevent common failure. While aware 11/2/95. during imestigating the containment purge valve wiring. it was resolution of precursor 95-noted that the control circuitry was routed through a non-2501. Lic.cvaluating safety related cabir.ct.
alternatives for resching isolation of safety and non-safety-rclated circuitry.
96-01 4/8/96 The licensec made a modification to the make-up system NRC Edit.
Unk VIO First identified by NRC in this regarding an interlock installed to open the borateci water report. FSAR will be resised.
storage tank isolation valves on a low MUT water Ictcl. A submittal to the NRC was made, but no resision was made to the FSAR.
96-01 4/8/96 UFSAR accident analysis for a HPl line SBLOCA LIC Design Yes VIO identified in problem report concurrent with a LOOP and the loss of either vital battery 2/15/96. Ha-dware mods train was not properly analyzed. it failed to consider completed to meet FSAR several pieces of equipment.
requirements.
L 17 I
. ~..
Region Site Ref Na Date issue Ideetified Cy Category Chg Op/Eg Follown AdditionalInfersnation 11 Cr3ttal River 96-01 4/8/96 Tb-6tn basis of the spent fuct pool systcm was not NRC Edit.
Unk VIO Firstidentified by NRC in this incorporated into the UFSAR as folloux FSAR incorrectly report. FSARwillbe resised.
states that i180 fuct asemblies are allowed vs. the 1357 of license amendment 134; FSAR incorrectly states 16 refuelings can be handled vs 191/3 oflicense amendment 134; FSAR incorrectly rercrences a max spent fuel pool temp of 140oF vs 157oF amendment i34. FSAR incorrectly states that leakage from the spent fuct poni through the leak trench is monitored daily.
IIarris 95-13 9/28/93 FSAR stated all leakage from ECCS pcst-LOCA LIC Desig:
Unk recirculation system would be filt: red by the Reactor Auxiliary Building Emergency Exhaust System prior to release offsite. Certain portions of the system were not enclosed in the emergency exhaus: system boundary and therefore could not be riitered during an accident.
96-02 4/9/96 Spent Fal Pool Cooling System assumptions as describcd Ll^
Design Unk IFI in FF AR utre not consistent with actual plant configurat;on. Condition was identified during licensce resiew following recen' industry _ issues on the subject.
Prelim calculations indicate design basis of SFP cooling l
- ystem was not exceeded as result of en 3rs. FSAR revision to l'c submitted.
l 96-02 4/9/96 Lic:nsec has dcycloped a FSAR improvement plan to be Info managed by site licensing group RECD 3!30/97).
b h
r 1
4 1
18
Region Site Ref Na Cate lesse Identified Cy Category Chg Op/Eq Followup AdditionalInformatioit 11 McGuire 96-01 4/3/96 (7) FSAR discrepancies noted: 1. Unit 1 operated with one LIC Design Unk component cooling water pump running. FSAR implies Proc /
two pumps should be running. 2. The fuel crane underload Ops Edit.
switch opens the rmin fuel hoist drive circuit when the suspended load drops to 2100 pounds or less. This setpoint was actually 1740-1780 pounds due to a change to the calibration procedure based on manufacturer design -
specifications. 3. Manipulator crancs contain positive stops which prevent the top of the fuel pelicts in a fuct assembly Sm being raised to within ten feet of normal water lestl. Actually, the upper limit exitches on crancs limit height but do not ensure ten feet of water costr. 4.
The highest level above the fuct racks that the fuct assembly can be dropped is 3 fect, two inches. The re-rack modification changed the height of the fuct racks such that the highest level would bc 3 fect, six inches. 5. The hoists supporting the wrir gaics utre connected by two separate cabics, cach cabic supportiag the entirc load. Actually, the wrir gates are connectL1 to hoist by one cable. The accident analysis accounts for dropped weir gate. G. Spent fuct cask lifting height was limited to 12 inches with cask shock absorbing cover not installed. There ucre no admin. limits or physical restrictions on the crane to crsure this limit. This 12" liinit is used in drop analysis.
- 7. Fuct lifting and handling desiccs utre capable of supporting maximum loads under Safe Shutdown Earthquake (SSE) conditions. No documentation was available to validate this scismic capability.
96-01 4/3/96 (3) FSAR discrepancies noted: 1. The reactor manipulator LIC Editj Unk crane was det>gnet to prevent disengagement of a fuci Design assembly frorr, o gripper in an SSE. Ne documentation was available to support this scismic capability. 2. Decay heat of spent IW as analyzed for turhe month refueling cyrk strent analysis addressed refueling c3ric of greater thaa twelve months. 3. Long term SFP makeup sources included the reactor makeup water storage tank (RMWST) and the refueling w ter storage tank. both at 2000 ppm boron. The RMWST uns not a borated water source. (SER Supplement 6. Section 3.3) 19
Region Site Ref Na Cate Issue Identified Cy Categen Chg Op/Eq Followup AdditionalInformation 11 North Anna %-01 3/21/96 ' la response to IN 95-54, licensee identified that Llc Proc /
Yes URI administrative controls to limit component cooling water Ops temperaturc were necessary to emre that the 140 F SER limit associated with installation of high density storage racks not be exceeded for a normal (fuil core) off-load.
96-01 3/21/96 Operation of the non-scismic refueling purification system NRC Design Yes to purify water in the scismic RWST is discussed in the UFSAR; however, it was not clear if this was allowed at pourr. Also, operation of this system for RWST temperature control was a mode of operation not described in the UFSnR.
96-01 3/21/96 The EDG testing technique specified on UFSAR page NRC Edit.
No N/A 8.3.21a was not being performed. An administrative oversight resulted in this page not being deleted uhen resision 23 was issued.
96-01 3/21/96 UFSAR stated that the process vent monitors were NRC Edit.
No N/A clectrically powered from cmergency 480 rac power panci l'Il-l. No mention is made of the alternate power supply panel 111-1.
96-01 3/21/96 Six blowout pancis in instrurr,entation tunnel access hatch NRC Proc /
Yes VIO lssue identified by NRC during l
locked shut. FSAR requires they open on.5 psi Ops inspection period 2/13/96.
differential pressure across the panel.
Action taken to restore plant to original configuration.
x 96-01 3/21/96 incorrest pages found in UFSAR controlled copy. Licensec NRC Edit.
No N/A resicued & corrected errors.
Oconce 95-30 2/22/96 Spent fuel pool level of 23.5 fcct (rcquired by FSAR) was NRC Proc /
Yes DEV maintained by plant procedures at 21.5 fect.
Ops i
20
~
Itegion Site Ref Na Cate lesne.
Identitled Cy Category Chg Op/Eg Follownp AdditionalInfonnation II Oconee 96-03 4/4/96 (4) Inaccuracies noted in SER: 1. An SER amendment NRC DesignPr Unk stated that if SFP water temperature was initially 125 F '
oc/ Ops boiling would occur greater than 9 hoors aner loss of SFP -
cooling. Calculation OSC-4998 for Unit 1/2 Heat Up Rate, determined that the actual timef te boil could be less than 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> for higher heat loads.. An SER amendment 2
stated that the :cquired make up rate will be less than 70 gpm for Unit 1/2 SFP. *Ihis addressed water loss due to boil off only and did not account for the 29 gpm RCP seal supply. Combined losses muld exceed the 70 gpm value.
This was not a concern since the refill capacity exceeded 150 gym. 3. An SER amendment stated that the times of 15 and 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for Unit 3 SFP boiling in the normal and i
abnormal heat load condHons respectively, utre sufl;cient to prmided emergency SFP makcup. The emergency procedure for Refilling SFPs specified 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> for corr.pletion of the pumping system for SFP refill and 72 i.ours as the upper limit to begin pumping to the pocl. 4.
l An SER amem wnt references maximum normal and 5
abnormal predicted heat loads, values w hich will hot be accurate when the higher enrichment fuct assemblics are transferred to the SFP in future refueling outages.
L Sequoyah 96-02 4/22/96 Spent Fuct Pool craporation rate in FSAR is incorrect.
LIC Edit.
No
[
FSAR states 55 gpm under certain conditions but licensec's calculations show it to be 103 gpm. FSAR i
resision planned.
l L
%-02 4/22/96 UFSAR states low pressure CC' system is tested in LIC Design No i
accordance with code requi<
.ts. The CO2 system for the cable spreading room I-st been tested since 1982.
This system is not listed.
equired system by the TS j
but is a backup to the automatic sprinkler system; j
therefore, the licensec does not consider this system is j
required to be tested. This item uns identified by TVA t
I during a 1995 audit. TVA's resolution was to rernove this system from the cable spreading room and delete the i
UFSAR reference to this system. NUREG-0800 Standard Review Plan requires a water suppression system for cabic spreading rooms but does not require a backup CO2 system.
i L
[
t 21 2
Region Site Itef NJ. Cate Issue Identified Cy Category Chg Op/Eg Followsp AdditionalInformistles 11 Sequogh
% 02 4/22/96 UFSAR Section 6.8 desenbed the licensee's pump and LIC Edit.
No vaht insenice testing program for the first 10-3 tar insenice inspection period. Houtser, the licensee commenced the second '0-year insenice inspection period in December 1995. The is:nsee intends to remme Section 6.8 from the UFSAR as discussed in NUREG-1482, GUIDELINES FOR INSERVICE TESTING AT NUCLEAR POWER PLANTS.
r 96-02 4/22/96 Spent Fuct Pool cooling section indicates it normally NRC Edit.
No handles a 40 percent core offload although elsewhere FSAR clearly states that a full core omond is typical.
St. Lucic 96-03 2/22/96 Unit I procedures for adding a mixture of dcmincralized NRC Proc /
Unk eel Identified by NRC during
'[
water and boric acid to the RCS (manually to the suction Ops inspection period 1/22-23/96.
of the charging pumps) did not implement the method Part of aggregate Lesti 111 siol.
t stated in the FSAR (automatic and to the volume control FSAR change prepared. Also l
tank) and had not donc so since January,1976.
rcccived viol for inadequate 50.59 in making procedurc change in response to FSAR t
inconsistency j
96-01 3/18/96 2B CS pump casing valve uns leaking past it's seat and NRC Proc /
Unk IFl the 2A LPSI pump vent uns also leaking. Licensec has Ops not yet determined whether leakage is within assumptions of FSAR accident analysis.
Summer 96-02 4/8/96 FSAR states that a special, narrow band, d-c voltage relay NRC Design Unk monitors Class IL hattery voltage and initiates an alarm in the control room if ottery voltage falls slightly below normal float voltage. With the current control room i
annunciator setting, this monitoring capability is not cfrective to indicate a slightly below normal float voltage.
j 96-02 4/8/96 Isolation of feedunter pump discharge vaht powcr circuits N.T Design Unk IFl consists of a magnetic breaker, a con: actor / thermal overload desice and fuses. The control poner to the contactor/ thermal overload desice was non-safety and i
hence. no isolation credit was taken for this desicc. A' I
magnetic breaker without starter thermal orcrioads was not included in the FSAR description of Class IE overcurrent desices. Licensec evaluating.
t I
r b
22 7
Region Sit 2 Ref No Cate issue Identified Cy Category Chg Op/Eg Followup AdditionalInformation 11 Summer
%-02 4/8/96 Inconsistency betmen FSAR and TS concerning time NRC Edit.
Unk URI requirements for monthly operation orcontainment atmosphere cleanop trains to reduce moisture buildup.
Licensec complies with TS.
89rry 96-02 4/19/96 UFSAR section on Control Room and Relay Room NRC Edit.
No Ventilation stated that 3 refrigeration chillers sen'c the ventilation system and that all chillers are located in hER-
- 3. The actual configuration of the refrigeration chillers 2
had been modified by a design change to add two additional chiiicts located in a separate hER, called MER-
- 5. This design change added redundancy and separation to refrigeration chiller system. UFSAR revision planned.
96-02 4/19/96 UFSAR describes outside containment isolation valve,2-NRC Edit.
No Ril-MOV-200, as a motor operated gate valve. The motor has been electrically disconnected and this valve is now a manually operated gate vahr. This inconsistency also exists in for the Unit I application.
Turkey Point 96-02 4/22/96 Licensec scif h/cssment & audit of 9 UFSAR chapters in LIC Info.
Unk March % noted 56 (-ficiencies. Team wrote condition report & prepared UFSAR changes for Oct 96 update to reschc deficiencies.
964)2 4/22/96 (5) UFSAR discrepancies noted: 1. UFSAR does not NRC Edit.
Unk URI rcilect the racetrack, ball field. and air show field in the transient population section. 2. UFSAR describes abandoned equipment in the rad waste building as active.
- 3. Permanently installed equipment in the refueling cavity that has been abandoned in place still referenced in the UFSAR m attive. 4. UFSAR does not reflect the primary use of temporary reactor cavitt filtration systems nor temporary liquid radwaste processing systems, in place since the 1980's. 5. Black Start and C Bus modifications completed in the fall of 1995 are not updated in UFSAR.
96-02 4/22/96 Full core ofiloads for normal refueling outages were not NRC Proc /
Unk URI 95-19 analyzed for heat load in UFSAR. Also portions of Ops UFSAR continue to tellect other than full core ofiload as normal refueling method.
l 23
Region Site.
Ref No Date. Issue Identified Ey Category Chg Op/Eg Followup AdditionalInformation 11 Vogtic 95-28 1/10/96 Unit 1 FSAR wording has normal core omoad of one-NPC Edit.
No N/A third, max normal omond is 40%, and emergency omond is 100*A. Normal practice has been to do a full core omead and ultimately end up with a third of the core in the pool. Wording in Unit 2 FSAR says normal is one-third but did analysis for up to 100% omoad. SER Sup 8 (1989) says to do full core omoad every refueling. FSAR change to be made to make U-l FSAR read like U-2.
Watts Bar
%-02 3/7/96 FSAR outdated regarding the description of the main NRC Edit.
No N/A control room habitability system area. Reference was made to certain rooms by titics that are now used for other purposes. The licensec initiated an FSAR change.
111 Big Rock
%-02 4/10/96 UFIISR states that all work in radiation areas and all NRC Edit.
No Point entrics to high radiation, contamination, and airborne areas requires the use of a RWP.11ourver, Technical Specification 6.12.1 gives exemptions to entering high radiation areas without the use of a RWP under special circumstances. The li:ensee is currently in compliance with the TS and planned to initiate a change to UFIISR to incorporate the exemptions.
96-02 4/10/96 The number of security omccrs used to counter security NRC Proc /
Unk IFl events during training drills exceeded the number of Ops personnel availabic under the security plan criteria. The protection strategy emplo3cd does not agree with the security plans.
Eh ron 95-13 3/20/96 in the UFSAR referenced security plan some capabilitics NRC Edit.
No IFI of certain security components utre not accurately described. In cach case howes cr. the existing capabilitics of the security components equalled or exceeded the capabilitics described in the security plan.
Cook 95-15 I/29/96 50.59 for design change package (DCP) for relocating NRC Proc /
No EDG starting relays uns inadequate. in response to the Ops question
- Docs the proposed design change represent a change to the plant as described in the SAR. Emergency Plan or Security Plan?" the licensee responded with
" Ec dicscl generators are not explicitly described in the 3AR. This design change does not afTect the dicscl generator controls as described in sections 6.1.1 and 8.4 of the UFSAR."
24
-~_m._.
__.-m.
_ -.m. _.. _...
...__m
. i Reglen Site Ref Na Cate - Issue.
' IdentlGed Tiy Category Chg Op/Eg Feueway AdditionalInformselen 111 -
Cook 95 1/29/96 During review of the large bore piping seconstitution -
NRC Proc /.
- Unk-
- program support modification g-kgn, the inspectors
' Ops -
determined that the licensee was not performing specific operability evaluations for each support found to be outside thelicensing basis.
96 4/2/96 The placement of a tarp inside the Unit 2 containment mis -
NRC Proc /
No VIO Item identified by NRC during performed without a 50.59 review of UFSAR Ops inspection period (report dated -
commitments.
4/2/96). Licensee remmed tarp.-
% 02 ' 4/2/96. UFSAR states that containmem recirculation sump will NRC Proc /
No IFI i
have alarms and redundant lesti indicators reading out in Ops the control room. The licensee removed the recirculation e
sump level indicators and mmed them to the adjacent ~
sump which is connected. De licensee failed to properly j
update all pertinent sections of the UFSAR at the time of
[
the modification.
' [
%-02 4/2/96 New Fuct Vault criticality monitor was not addressed NRC Proc /
No URI FSAR. Arca radiation monitors / criticality monitoring Ops j
1 desices were a part of the licenscc's design and licensing l
basis that should be in the USAR. -
l E03 4/17/96 UFSAR did not document acceptable temperature ranges NRC Edit.
Unk j
for CVCS process fluid.12% boric acidc The system in many cases was being maintained above the 175 l' high temperature alarm setpoint.
E03 4/17/96 UFSAR indicated that the CVCS diaphragm valves were NRC Edit.
No installed in a 200 F portion of the system. However, the l
system temperatures were not being maintained below 200 F. Subsequent resiew found that vendor data j
qualified the valves to 300 F.
t l
% 03 4/17/96 UFSAR did not document the permissible voltage ranges NRC Edit.
No j
for acceptable system operation of 4160 volt electiical rystems. The design basis prmided by the licensee noted that the acceptable voltage range was 3600 to 4400 volts.
I
%03 4/I7/96 UFSAR CCW fabrication description noted that tie NRC Edit.
.Unk
[
system was fabricated using stainicss steel piping and components which were welded where applicabic. The team noted that a temporary modification had instatted a chemistry sampling unit to the supply and rewarn lines of the CCW system using tygon tubing. The piping was considered to be safety related and could only be isolated manually.
.25
Region Site Ref N3 Date issue Identified Cy Category Chg Op/Eg Followup AdditionalInformation
.III Cook 96-03 4/17/96 CCW flow balance suntillance was not consistent with NRC Proc /-
Unk the UlliAR requirements specified for sample cc ner flow Ops (142 GPM vs 240 GPM). CCW flow through containment air recirculation units was not addressed during flow balance suntillance.
96-03 4/17/96 UFSAR indicated that gross gamma analysis of the reactor NRC Edit.
No coolant system would be performed 20 minutes after sampiing as util as a periodic analysis via gamma spectroscopy. Licensec was currently performing the I
gamma spectroscopy activitics but were not performing gross gamma analysis. Licensee has a letter from the NRC relieving them from the gross gamma commitment l
based on the gamma spectroscopy that was donc and plans a revision to the UFSAR.
Davis-Besse 95-09 2/8/96 UFS AR section on transient analysis indicated that if a NRC Edit.
No IFI LOOP utre to occur with the unit at full poutr, the reactor and turbine would not trip. Near the end of the UFSAR LOOP analysis, the UFSAR indicated that the reactor may trip from 100 percent power. Although this
" add-on" portion of the analysis appeared to be accurate, most of the analysis addressed a sequence of crents that j
uns no longer valid.
95-09 2/8/96 During review of an unexpected SFP radiation alarm. a NRC Design Unk IFl N
UFSAR figure indicated that the maximum dose expected for general areas adjacent to the spent fuel pool while operating with I percent failed fuct uns < 15 mr/hr. Fucl manipulation resulted in an approximately 23 mr/hr field in portions of the general area. about i I/2 times the UFSAR anticipated value.
95-10 4/19/96 Valve MS-853 is considered a containment isolation NRC Edit.
No IFI boundary. but did not appear in the UFSAR listing. The UFSAR states tiu am*.ts of this type utre controlled in other administrative programs such as the " blue cap" program. The licensec is determining if these vahrs.
should be placed into the UFSAR.
95-10 4/19/96 Some values for hot leg volumes as listed in UFSAR LIC
. Design No IFI tables and Figure 5.1-1 were inconsistent. Potential errors in computing operating limits in the Core Operating Limits Report was a concern.
26
.=
Region Site Ref Na Date Issue identified Cy Category Chg Op/Eq Followup AdditionniInformation III Davis-Besse 95-10 4/19/96 UFSAR has misicading description ofIntegrated Control NRC Edit.
No IFI System capability. UFSAR states that the ICS was 7
designed to allow a 100% load reject sia the turbine bypass vah es and code safety vahts without a scram.
However, the Powtr Operated Relief Vah c (PORV) setpoint uns char:ged to a value above the overpressure trip setpoint and would have caused the plant to scram afict a 100% load reject according to another UFSAR section.
Dresden 95-15 3/26/96 Scrart discharge volume gallery steel structural design 1.IC Design Yes IFI margins not met (AISC deviation). Unit 2 modification corrplcte. Unit 3 at next refuel.
95-15 3/26276 Unit I ventilation practices, hot shop and contaminated LIC Proc /
Yes URI stora'c on turbine deck, and asbestos removal desiations Ops with respect to various decommissioning licensing documents. Comed perfoiming 50.59.
96-05 4/l1/96 Undocumented and unanalyzed structural steel load LIC Design Yes eel NRC learned ofissue 8/24/95 changes in LPCI corner rooms were known to exist since during review of Quad Ci ics 1991 and the structural stect design margins utre known document. Licensee corponte to be exceeded since at Icast January 1994. Existing plans office awarc in 1991, site aware would not have resolved these nonconforming conditions ofissue 8/95. 50.59 until approximately six years aller initial identification.
cvaluation performed. plant
( \\pparent Violations for inadequate design control &
modifications to be failure to take prompt corrective actions) impicmented to restore plant to code.
Duanc 95-11 1/25/96 (2) Discrepancies. I. Certain MOVs utre removed from NRC Proc /
Unk Arnold the GL 89-10 program because they have no safety Ops function to re-position. They are usually in th :ir safety position. cxcept for brief periods during surveillance i
testing or other " secondary modes" of operation. The UFSAR states that if an initiation signal occurs while HPCI system is being tested, system valves align automatically to the injection mode. A 50.59 resiew was l
performed in 1994. 2. The licensec does not enter LCOs for tect;nical specification required surveillance testing.
cren though in some cases. that testing may render the system technically inoperable (but usually "availabic" with some operator actions). There is no 50.59 review for this item.
27
Region Sit 2 Ref Na Cate Issue Identified Cy Category Chg Op/Eg Followup AdditionalInformation III Duanc 9642 4/12/96 11,0 gallons of uiter entered the HPCI turbinc exhaust LIC Design Unk--
LER Arnold piping in Dec.1995. The licensee concluded the water was drawn up from the torus due to a leaking check valve in the exhaust line, and that the installed vacuum breakers were functioning properly. UFSAR specifics that the installation of the vacuum breakers was to ensure that durirtHPCI system operation and subsequent shutdou11 no differential pressure would exist that could cause torus water to enter the exhaust lines and cause water hammer.
This inconsistency will be resicurd during closure of LER 95-013.
96-02 " 4/12/96 UFSAR describes the fire protection system as having NRC Proc /
Unk IFI pressure maintained by a jeckey pump and accumulator Ops combination. The accumulator has been isolated and tagged out since 1992. No 50.59 safety evaluation had been performed.
96-02 4/12/96 In February 1995, the licensee identified, through testing, LIC Design Unk that the ESW makeup flow rate to the spent fuci pool was less than design and less than specified in UFSAR. A 50.59 cvaluation documented the rationale for the conclusion that there uns no unresicurd safety question.
This inconsistency will be revicurd by NRR as part of the Spent Fuct Pool Licensing Basis Review 96-02 4/12/96 DC ponned RCIC steam supply valve MO 2401.
LIC Design Unk IFI clectrically back scated. may exceed the UFSAR design closure time of 20 seconds under design basis conditions.
Calculations showed that under degraded voltage and full flow conditions. stroke time would be 22.7 seconds. The licensec subsequently resolved the issue and documented the basis for operability, w hich included a statement that the values for closure time in UFSAR are nominal in l
nature and not based upon detailed analysis. This inconsistency will be resiewed further.
1 6
4 4
28
Itegion Site Ref N3 Date Issue Identitled Cy Category Chg Op/Eg Fellowup AdditiemalInfennation
.III
. Duanc 96-02 4/12/96 The normal supply ofco...,M tir for the safety-related NRC Design Unk IFI Arnold standby gas treatment (SBGT) system is not safety-related. UFSAR specifies failure of the normal compressed air system will not affect operation of the system because of the safety-related seismic category I standby air compressors (IK-3 and IK-4), available if the main plant compressed air system fails. The definition of operability in TS includes the statement that necessary attendant auxiliary equipment required for a system to perform its function are also capable of performing their related support function. Duane Arnold did not enter a TS LCO or consider the SBGT inoperable when IK-4 was out of sent:c on 1/2/96. This inconsistency will be rcsicurd further.
I Fermi
%-02 4/3/96 UFSAR stated that the " heart of the permanent Fermi 2 ~
NRC Edit.
No URI solid radwaste system is the radwaste volume reduction and solidification system," and that a wndor system is to be used w hen the aspha:t system (above) is not working or at plant management discretion. The system has not been used for several years. The vendor system has been used exclusively. The UFSAR also stated that two frcon dry-cleaning units are used for cleaning contaminated laundry. Hourver, the licensee used an offsite vendor facility for this purpose.
96-02 4/8/96 UFSAR listed the maximum flow rate through the shcIl NRC Proc /
Unk.
IFI-side of the EESW/EECW heat exchanger as 1450 gpm.
Ops Routine suntillance runs used a flow rate of 1670 gpm.
96-02 4/8/96 UFSAR stated safety equipment cabling was color coded NRC Edit.
No URI orange for Division I and blue for Division 2, while BOP cabling was black. or magenta. During construction extra divisional color coded cabling was used to complete BOP clectrical distribution.
Kcwaunce 96-03 4/15/96 USAR is not clear on which valves are required for turbine LIC Edit.
No
- FI overspeed protection. One section states both the tchcat steam stop and intercept valves close on mrrspeed u hile another section and the suntillance procedure imply operability testing of only the turbine stop and gmtrnor vahrs. However, periodic testing was performed on all turbine stop snd governor valves and reheat stop and intercept vahrs.
29
Region Sit]
Ref NJ Cate Issue Identified Cy Category Chg Op/Eg Followup AdditionalInformation til LaSalle 96-02 4/9/96 The humidification equipment installed in the corttrol LIC Design Yes room and auxilia y electric equipment room ventilation systems were not scismically supported. 50.59 cvaluation determined that an unresiewed safety question does not er ist. Afodifications to be made to return the system to full compliance with the UFSAR.
96-02 4/9/96 Refueling practices were not fully consistent with FSAR NRC Edit.
No N/A wording. The FSAR was open to interpretation on whether or not a full core omoad is an emergency heat load. Comed's calculations show that during cooler months v. hen the lake temperature is much less than design temperature, a full core omoad is not an emergency heat load on the spent fuel pool. As a result of the resiew, Comed resised the FSAR to be consistent with the current refueling practices and clarify the wording.
96-02 4/9/96 In 1994 Comed discovered that the condenser mechanical LIC Design Unk N/A vacuum pump did not have an automatic trip feature as described in the FSAR. A 50.59 was performed and determined to be a unresicurd safety question.
Notification was made to the NRC and an was SER issued by NRC. The FSAR has not been up-dated to reflect this.
Afonticello 96-02 3/21/96 USAR contained two maximum spent fuci pool LIC Edit.
No IFI temperatures.125cF and 140cF. The licensee believed that 1250F referred to normal omoads whereas the 1400F referred to the cmcrgency omoad. USAR resision planned.
96 02 3/21/96 Inspectors observed instrument and control personnel lift NRC Edit.
No N/A covers clTofinstrumentation during surveillances. USAR stated that operations personnel must remove the cover plate, access plug. or scaling device from instruments.
This discrepancy did not have safety consequences. The licensee planned to resisc this requirement.
95-10 4/9/96 Two containment isolation vahrs failed to meet separation LIC Design Yes IFl criteria in that the recirculation system process sampling containment inboard and outboard isolation valves position indicators ucrc poutred from the same electrical source. A single failure of the power source would cause a loss of position iridication for both valves w hich was contrary to the licensec's commitments in response to Generic Letter 82-33. " Supplement I to NUREG-0737 l
Requirements for Emergency Response Capability,"
30
Region Site Ref Na Cate
!ssue -
Identified Cy Cat" gory Chg Op/Eq Followup AdditionalInferniation III Monticello 96-02 4/9/96 Discrepancies in documentation regarding tornado design LIC Des.gn Yes IFI requirements for the r.fuct floor area superstructure. One USAR section stated that the Class I structures except the steel superstructure were designed to withstand a tornado load. Hourser, another section referenced a GE report,
" Tornado Protection for the Spent Fuct Storage Pool,"
which stated that the superstructure was designed to withstand tornado winds. Licensec planned to submit a letter describing issue and corrective actions.
Palisades 96-02 4/8/96 Palisades operating procedures did not comply with FSAR NRC Proc /
Unk URI and 10 CFR 50 Appendix K regarding assumed Ops instrument uncertainty in measuring reactor power. This
~,
may have resulted in the licensec operating the plant at power levcis up to 100.99 percent, in difference to the assumed initial conditions for transient analysis in the
" Safety Analysis" section of the FSAR.
Perry 96-02 4/1I/96 UFS AR table showed an cmcrgency lighting unit NRC Edit.
No IFI illuminates Emergency Closed Cooling Vaht P42-F0551.
The light was ineffective because it was too far away from the vahr and there was a wall betuten it and the vahr.
96-02 4/11/96 UFSAR states " Perry normally has a minimum of five NRC Edit.
No IFI operating shiR crews. Four shin crcus may be established during certain phases such as startup testing or extended outages to maximize training." inspectors obscited licenscc was using three shins during the refueling outage and that they had begun using the three shin schedule about 3 wecks before the outage. UFSAR being changed to delete reference to the specific number of shin crews required during plant shutdown conditions.
96-02 4/11/96 UFSAR does not address free fall of the polar cranc's hook LIC Design Unk IFI or uhether the cranc should be single failure proof.
Licensec evaluating the need to make the polar cranc sing!c failure proof. An interim administrative control halted the use of the polar crane over the open reattor vessel cavity for activitics not specifically addressed in the UFSAR.
31
Region Site Ref N3 Cate issue Identified Cy Category Chg Opra Followup AdditionalInformation 111 Perry 96-02 4/11/96 Structures supporting the suppression pool shield doors LIC Design Unk URI had not been analyzed for pool swell structural loeds while open. Subsequent analysis restaled some structural components were not adequate to meet design code requirements during a pool surli. License amendment requested. Approval of the amendment request would allow the doors to be open briefly at low reactor pourt levels until the next refueling outage. -
Point Beach %-02 4/17/96 Violation of 50.71 requirement to update FSAP annually NRC Edit.
No VIO NRC identified issues during or 6 months aller cach refueling outage. Minimum senice inspection period (11/95 -
water flow to the containment accident fan coolers is 920 2/96). FSAR being updated.
gpm sice 1000 gpm as stated in FSAR. Actual operating Process review team reviewing average coolant temperature of both Units is $70 F vice entire FSAR for discrepancies 573.9 F as specified in FSAR. An analysis and a FSAR as part of reticw/ update process.
change increased the design senice water temperature to 75 F; howcrer, this temperature has not been updated in all sections of the FSAR for systems using senice water cooling. These conditions existed a minimum of six months prior to the presious FSAR update.
96-02 4/17/96 Recently, station blackout mods hast connected two NRC Edit.
Unk additional dicsci generators; so FSAR descriptions and diagrams do not accurately reflect the actual conditions.
Licensee has presided operations temporary diagrams.
Licensee updates its FSAR crery June.
96-02 J/17/96 Despite an analysis and FSAR change that noted design NRC Edit.
No SW temperature change to 75 F. this temperature has not been updated in FSAR sections for systems using SW cooling.
96-02 4/17/96 FSAR states there should be essentially zero leakage from NRC Design Unk the mechanical scal of RilR pumps. CS pumps and SI pumps. All these pumps exhibit leakage to some degree.
96-02 4/17/96 FSAR indicates spent fuct pool cooling system heat NRC Edit.
No exchanger inlet temperature is monitored. What is actually r
measures is SFP temperature at a location on opposite side of pool from suction line to the cooling system.
32
Region Site Ref No Date Issue Identified By Category Chg Op/Eq Followup Additionniinformation Ill Point Beach %-02 4/17/96 FSAR regarding the leakage provisions of the spent fuel NRC Design Unk.
pool cooling system states "The normal operating pressure of the senice water system is higher than the normal operating pressure of the spent fuct cooling system. In the event of a heat exchanger tube leak, this differential pressure will result in leakage from the service water system to the spent fuel pool cooling system." This is contrary to obsened pressure drops across the heat exchanger. During in-senice testing of the SFP cooling heat exchanger, SW inlet pressure was 52 PSIG, the outict 7 PSIG. The SFP cooling inlet pressure was 30 PSIG, the outlet 6 PSIG. The pressure drop the pool unter experienced while traveling through the tubes was 24 PSI.
Thercrore, if a tube leak occurred in the area near the shell side outlet the leakage uvuld be the rcyctse of that stated abo c.
96-02 4/17/96 EDG starting air from the G-01 and G-02 storage tanks is NRC Edit.
No listed in the FSAR as being admitted at a working pressure of 200 psi. This is the high pressure cut-ofTpoint for the air compressor pressure switch. The starting air pressure for these tanks is normally kept at 185-190 psi. with a Technical Specification basis mini..um pressure of 165 psig.
96-02 4/17/96 Conflicting FSAR statements concerning location of spent NRC Edit.
No fuel pool cooling system syphon breakers.
Prairic Island 95-14 2/6/96 Change made to the head botting on one of the dry casks NRC Design No VIO NRC identified issue during that resulted in less thread engagement and a longer inspection period (approx.
bending radius for the head bolts than assumed in the 12/11/95). Licensec knew of USAR. Cask vendor had performed an engineering bolting change mid-95, but evaluation of the change but *mt a documented safety didn't see the need for 72 'M cvaluation in accordance uit.10 CFR 72.48. (Severity evaluation. Perform ' '
Level IV violation issued) 12/20/95; developt scrcening form A trained stalT.
joined industry initiative on enhaneed SE guidanec; completed scif-assessment on SE program using industry experts.
33
Region Site.
RefNJ Date issue -
IdentifledIy Category ChgOp/Eg Follownp AdditlegalInfonmation j
111 PrairieIstnd 95-14 2/6/96 Test 13 measure Iknv through the emergency intake line to -
NRC Proc /-
No VIO NRCinformed ofissue sia the cooling water pumps should haw been considered a Ops 50.72 call 11/20/95. Licensee special test and should have had a safety evaluation per 10 revisedearthquake2,mn A CFR 50.59. The test was described in the USAR as a indw revised Safety preoperational test but had newr been done at power Evaluation to take credit for before.
non-seismic canal-NRR currently resiewing.
95-14. 2/6/96 50.59 safety evaluation donc for response ofintake bay to NRC Design No URI scismic event may have resulted in an unresicued safety NRR question not identilled by licensee. The licensee assumed that the sides of the bay would not instantly stuff off. That assumption, u hile probably reasonabic, appears to bc
+
contrary to the licensing basis.
t
% 02 3/19/96 USAR has a misicading description of material used to NRC Edit.
No N/A construct waste gas tanks.
% 02 3/19/96 Pipe rupture analysis for fire protection piping used LIC Design No URI
. assumptions that inay be inconsistent with plant configuration.
i Quad Citics %-02 4/17/96 UFSAR and other plant procedures require resision to LIC Proc /
Unk IFI
{
reflect the frequency of full core offloads and presions Ops licensing commitments from the SER issued for high density fuel racks.
% 02 4/17/96 FSAR design basis for ventilation system for control room.
NRC Design Unk IFI turbine building and reactor building assumes outside air temperature range of-6 F to 93 F. Temperatures during inspection period ucre as low as -28 F. Licensec evaluating.
%-02 4/17/96 Some structural stect beams and connections supporting LIC Design Unk URI RIIR heat exchangers utre determined to be overstressed relative to U iAR allowabic stress limits. Comed completed an _perability determination with supporting functionality crahiation. The operability determination t
shourd that the analyzed beams meet functional criteria.
but did not meet UFSAR allowable stress limits. Initial -
discovery was made during contractor reticus and walkdowns of associated piping supports, where a number i
of pipe supports utre not accounted for in existmg r
calculations.
I t
34
?
Region Site Ref N; Date Issue Identified By Category Chg Op/Eq Follownp AdditionalInformation 111 Zion 96-05 4/5/96 A longstanding non-conforming condition where the NRC Proc /
Yes DEV licensee did not maintain operability of a three-hour rated Ops fire barrier as specified in the UFSAR referenced fisc hazard analysis report.
e IV Arkansas
%-14 3/13/96 Discrepancy between the Units I and 2 descriptions of a NRC Edit.
No shared component contained in the UFSAR was identified.
%-02 4/8/96 FSAR indicates maximum temperatures expected in fuel NRC Design Yes URI pools are based on a max lake temperature of 85 F,3rt lake temperatures routinely exceed this valuc in the summer.
96-01 4/8/96 As a result of the problems identified at other plants, LIC Info licensec has initiated their own review of FSAR accuracy.
%-01 4/8/96 FSAR states that a complete description of the licensec's NRC Edit.
No 96-1I resiew and audit program is discussed in Section 6.0 of TSs. Ilowever, the licensee had moved the program descriptions to the Quality Assurance Manual. FSAR change initiated.
96-01 4/8/96 A modification removed the flow balance function of the LIC Edit.
No ilPSI header isolation tahts and installed manual i
throttling vahrs in cach header to balance the flow between the headers. FSAR change tcqucst initiated.
Callaway 96-02 4/2/96 inconsistencies noted beturen FSAR. TS. A surveillance NRC Edit.
No N/A procedurcs regarding allowable EDG start times. FSAR changes initiated.
Comanche 96-01 3/19/96 UFSAR stated that the external alternate AC input source NRC Proc /
No N/A Peak voltage was 120 Vac plus or minus 10 percent nominal Ops (107 to 132 Vac). Procedurc MSE-CO-5810. "10kVA Elgar Inverter Calibration and Adjustment." indicated that the restrse transfer lockout setpoint range was 131.4 to 132.6 Vac. This minor inconsistency was identified for their evaluation and correction, as appropriate.
35
Region Site Ref N3 Date Issue Identified Cy Category Chg Op/Eq Followsp AdditionalInfonnation IV Cooper
%-04 3/l1/96 On i1/9/95, the licensec declared the main steam tu nel NRC Design Yes eel issue identified by NRC blouvut panel sections inoperable due to a fiberglass 11/15/95. Licensee corrected coating, which strengthened the pancis and would have blowout panel to conform to presented them from operating at 0.52 psig as designed.
safety analysis /FSAR,imprmed A 1985 roodification to the panels, which was not reflected design control processes and in the UFSAR, rendered them incapable of relieving at maintenance unrk request design pressurcs. Sestrity Level Ill 50.59 violation.
implementation, and audited for other past unauthorized or unanalyzed modifications to the plant.
96-04 3/11/96 An apparent violation was identified for the installation of LIC Design Yes eel item identified during the 0.25 inch diameter J-tubes on the DG muffler bypass valve inspection period (3/11/96 solenoid exhaust ports without formal approval or report) Licensee remmed J-analysis. The J-tubes intermittently prevented the muffler tubes to make installation bypass vahr from opening when actuated, resulting in conform to FSAR, improved potentially inoperable DGs under scismic and tornado design control processes and conditions. (SL IV siol-App B, Crit 111) maintenance work request impicmentation, and audited for other unanalyzed plant mods.
96-03 4/15/96 Listing of penetration-2nd a listing ofIST boundary NRC Edit.
No vahrs in UFSAR had many errors of minor safety significance. UFSAR change initiated.
Diablo 96-02 4/10/96 proficiency training required by fire protection program LIC/ NRC Proc /
Unk VIO Issue 6rst known during Canyon and the UFSAR for individuals assigned to the fire brigade Ops inspection period (report was not completed.
4/10/96) Licensec has l
cstablished a new program for tracking fire brigade training and linked it with operator training, since all fire brigade members are in operations.
with exception of fire marshals.
96-02 4/10/96 Apparent discrepancy in the assumptions utilized to NRC Design Unk determine the plant's radionuclide source term. UFSAR assumed the plant uvuld operate on a 12-month cycic at a capacity factor of 80% Currently, Diablo Canyon Units I and 2 arc operating on an 18-month cycle and have historically exceeded an 80*4 capacity factor. Licensec resiew determined current analysis bounds source term.
UFSAR to be resised.
36
Region Site
- Ref Ma Cate Issue Identitled Cy Category Chg Op/Eg Fellowup Additteest Intersnaties 3
IV :
Diablo Not yet 4/15/96 Licensec task force identified sewrd hun' ed UFSAR LIC Info
- URI Canyon iss'd deficiencies. Licensee plans to reissue UFSAR in its
- o entirety; target date 11/96.
Fort Calhoun 96-01 4/8/96 Experiencing failures ofincore detectors. USAR LIC Design No IFl committed to having 21 incere detectors strings operabic.
w After the fiAh detector string (of 28 strings) wis' declared inoperable, a 50.59 evaluation to allow continued operation with as few as eight detectors, two per quadrant, was completed and apprmtd. The licensec intends to incorpo' ate evaluation results into the USAR.
Grand Gulf %-06 t 3/21/96 Storage locker in the remote shutdown panel room had not NRC
' Design Yes VIO NRC identified issue 3/21/96.
been scismically evaluated or properly secured in Licensee secured cabinet &
accordance with UFSAR guidance.
verified otheritems properly stored..
River Bend
%-02 3/21/96 Control room habitability related to toxic chemicals is LIC Proc /
Yes NCV based on the amounts orchemicals stored emite and listed Ops in USAR Tabic 2.2-5. The licensec store _
- frcon onsite than the values listed in the tabic. A.so, the licensec stored R-1 I onsite but the chemical is not listed in 4
the table. The licensce had not performed a 10 CFR 50.59 cvaluation. The licensec is evaluating procedure changes to assure that materials brought onsite would be compared to the USAR requirements.
%-01 4/18/96 Spent fuct pool and reactor cavity pneumatic gate scals NRC Proc /
Unk VIO Issue identified by NRC during were not controlled as safety-related equipment.
Ops inspection period. Licensee corrective actions not yet determined.
l San Onofre 95-30 3/6/96 Possible unanalyzed release path concerning RWST NRC Design No.
NRR-suction isolation valves failing. UFSAR de,u;bcd the -
design basis as has ing these valves shut (manually) on a recircubtion actuation signal. and because with these valves not shut. a possible unanalyzed release path existed (from emergency sump to RWST). the issue was referred to NRR for further evaluation.
4/1I/96 UFSAR was inconsistent with respect to local controls for NRC Edit.
No IFl 96-02 starting and stopping motor-driven auxiliary feedwater 3
t pmp 4/11/96 RCS wcid materials listed in UFSAR Tabic were not NRC Edit.
No URI
%-02 updated to reflect recent design changes.
37'
Region Site Ref N2 Cate issue Identified Cy Category Chg Op/Eq Followup AdditionalInformation IV Waterford 95-22 3/26/96 The actual inten:1 between performing the integrated leak NRC Proc /
Yes VIO NRCidentified issue during test for systems containing primary coolant outside of Ops inspection period (report issued containment (21 months) was not consistent with the 3/26/96). Licensec resised the FSAR, which required the test to be performed at inten als applicable procedures to specify r.ot to exceed each refueling outage (i.e.,18 months or the correct frequency for less).
performance of the test.
95-22 3/26/96 Use of an unresicued engineering evaluation uns NRC Design No N/A inconsistent with engineered safety features systems allowable leakage limits described in the FSAR.
96-03 4/17/96 h1ultiple examples of conflicting information between NRC Design Unk URI UFSAR and other design basis info for EFW system design basis requirements.
WNP-2 96-02 3/19/96 OfTgas system gas coolers have been operable for only 2.8 NRC Design VIO NC identified issue 3/19/96.
years c :t of 10. Violation of 50.59 for failure to perform liensee aware ofissue in written safety evaluation which prmides the bases for the 1988. FSAR to be chaaged.
determinadon that vault coolers described in the FSAR, but no longer operable, did not im olve an unresiewed safety question.
96-02 3/19/96 FSAR states that all three DG ventilation systems operate NRC Design Yes VIO NRC identified issue 3/19/96.
automatically to maintain ambient temperature at Licensec awarc ofissue in equipment operability limits during all modes of 1989. FSAR to be changed.
operation. Electric heaters are designed to maintain DG rooms at a minimum temperature of 70 F during extreme x
ucather conditions. This prmides a 7 F margin above the minimum temperature of 63 F. which is xquired to assure DG operability. The existing system camiot maintain the DG rooms above 70 F dming cxtremely cold ucather.
Violation of 50.59 written for failure to perform safety evaluation.
WolfCrcck 96-02 3/4/96 UFSAR lists minimum operating temperature for the NRC Proc /
Yes DEV clectrical penetration room and the charging pmp rooms Ops as 60 F. During inspection obsened 52 F in Electrical Penetration Room A on 1/23/96 and 52 F in Charging Pump B room on 2/6/96.
96-02 3/4/96 Requirements removed from TSs by Amendment 89 were LIC Edit.
No N/A not concurrently added to the UFSAR. Caught by shift sapenisor when document senices was making change to TSs without a concurrent UFSAR change.
O 38
.....