ML20202F634
Text
. - ~ _ - - , - _ _ . _ - - - . - . - - - . - _ _ ~ . ~ . - . . . - . . . _ . -
I 4;_t, 4-
- j l
From:
To:
/ /
KGL, CJA. DJF1, JBF ,
- Date: 4/16/981:48prn '
l
Subject:
RI-97-A-0145 Susquehanna The vocal alleger called this aftemoon He is still waiting for his letter. I told him that he should have it by next Friday (Dave has draft for concurrence).
Alleger was also VERY upset with NRC's handling of the Keith Davis issue. Alleger believes that it is NRC's fault for l ruining this person's life. j Alleger wrote to Arten Spector.
Abger said that PP&L has not told their employees about 93-03 letter and the implications it has on employees Alleger said that NRC fails in holding licensee accountable for passang informabon to e.4;r - I disagreed with ausger and tried to explain that if we issue a letter that requires a licensee to perform a task, the licensee does not Inform ev4j n :. and the task is not completed ...we would issue an NOV to the licensee. not the employee unless the employee was aware of the requerement and deliberately did not do it. This launched the ,
alleger into why can't we talk to the -.Gioes. Most, if not aH of our charu==ians concoming regulations and NRC l
~
expectations are with the licensee's management.
Alleger said he ran into someone in the grocery store who said, " I don't know where the NRC stands on this issue." _
l Alleger said that this person does not trust the NRC. I said this person can call anonymously to address his g question, Alleger said the NRC is hurtir.g innocent people (like himself) Alleger said that we (the NRC) has not made it clear to Mr. Davis - only to PP&L. I told alleger that there is an ongoing investigation and I cannot say anymore.
Alleger is aware that the NRC sent a letter to DOJ regarding the E diesel generator issue. He was really not happy.
Alleger believes that we are not headed where we should be in terms of dealing with PP&L Alleger said there ,
needs to be less fear and more respect.
Alleger doesn't expect the union to strike.
Alleger expects to hear the arbritrator's decision before he receives our letter. Alleger expects to win and get his job back. ,
. Alleger's hot tip: expect a case before the US Court f,:,randa rights vs. Wyngarden (sp?) Rights (has to deal with union representation during the audit).
That's it.
Kathy CC: DJV,SLJ l
l I
\/
9902040113 990129
.PDR FOIA SORENIBEN99-36 PDR 9965D(s'C/(3 ,. -
\ $1Wb -~ . - ..
_ _ _ _ . . . . _ _ _ _ -_ _ _ . _ . _ . _ . . . . . . _ . _ . _ . _ ~ . _ . .
- 9. u; !
. J, 9 3 _ c' i
l y 7; i L
ALLEGATION DISPOSITION RECORD Rev. 6/6/97 Allegation o.: RI-b-ALN0 7M k N-U / Branch Chief (AOC): l UU '
Site: H/ %
Acknowledged: No N/A Pane Date: T3/ Y Confidentiality Granted: Yes issue discussed (if other than original allegation): ~
Alleger contacted prior to referral to licensee (if applic'a61e)?
mua me u/4 ALLEGATION PANEL S (Previous Allegation Panels on issue: No)
Attendees Chair -
01 Rep. - IACed.
anc hie C) N 31/. b S d J N5rdmja W//Flm RI Counsel - Others . -7WB DISPOSITIOl# ACTIONS: (State actions required for closure (including special -
concurrences), responsible person, ECD and expected closure documentation) NOTE: If filling out j onically, use a larger, bold font to aid individuals in reading this material. '
1) 9 - /.
Responsible Person:
ECD:
f [
Closure Documentation: Completed:
- 2) hdM/I bbld b JYfh A Yl/M $f k D -/ W hfAd + do j7 hl:2?] W/* EWJWC MW f Re e' Person: MM ECD: /
Closure Documentation: Completed:
3) l l
Responsible Person:
ECD:
Closure Documentation: Completed:
I
' Safety Significance Assessment:
- Priority of 01 Investigation
} ARB MINUTES ARE REVIEWED AND APPROVED AT THE ARB 3
/
? .
- l l
- l NOTES: (Include rationale for anv referral to licensee. and identify any ootentially ceneric ellecations)
Issue not to be referred to licensee I l
A. Region 1 should refer as many allegations as possible to the licensee for action and I response unless any of the following factors apply:
I
- Information cannot be released in sufficient detail to the licensee without i compromising the identity of the alleger or confidential source (unless the '
alleger has no objection to his or her name being released).
- The licensee could compromise an investigation or inspection because of knowledge gained from the referral. .-
a The allegation is made against the licensee's management or those parties !
who would normally receive and address the allegation.
e The basis of the allegation is information received from a Federal agency that does not approve of the information being released in a referral.
Even if the above conditions exist, Region 1 shall refer the substance of the allegation to the licensee regardless of any factor if the allegation raises an overriding safety issue, using the guidance in Management Directive 8.8.
Factors to Consider Prior to Referral to a Ucensee in determining whether to refer eligible allegations to a licensee, The Region 1 Allegation Panel shall consider the following:
e Could the release of information bring harm to the alleger or confidential source?
- Has the alleger or confidential source voiced objections to the release of the allegatien to the licensee?
e What is thn licensee's histor/ of allegations against it and past record h dealing with allegations, including the likelihood that the licensee will effectivelv investigate, document, and resolve the allegation?
e Has the a leger or confidential source already taken this concern to the licensee with unsatisf actor / results? If the answer is "yes," the concern is within NRC's jurisdiction, and the alleger objects to the referral, the concerns should normally not be referred to the licensee.
l
1r l .
l l b .
l l
l e Are resources to investigate available within the region?
l Prior to referring an allegation to a licensee, all reasonable efforts should be made to inform
{
allegers or confidential sources of the planned referral. This notification may be given orally ;
and subsequently documented in an acknowledgment letter. If the alleger or confidential l l source objects to the referral, or does not respond within 30 calendar days, and the NRC {
has considered the factors described above, a referral can be made despite the alleger's or l
confidential source's objection or lack of response. In all such cases, an attempt will be l made to contact the alleger by phone just prior to making the referral. ;
Also, referrals are not to be made if it could compromise the identity of the alleger, or if it I could compromise an inspection or investigation. Note: Document the basis for referring allegations to a licensee in those cases where the criteria listed above indicate that it is i questionable whether a referral is. appropriate.
Distribution: Panel Attendees, Regional Counsel, 01, Responsible Persons (original to SAC) l Cotions for Resolution: I y
Licensee Referral (Div. Dir. Concurrence Required (First Consider Factors Prior to Referral) / Document NRC Review of Response - Resp. - AOC)
Referral to Another Agency (OSHA, etc. - Resp. - SAC)
Referral to an Agreement State (MD, ME, NH, NY, RI - Resp. - SAC)
Referral to Another NRC Office (OlG, NRR, Other Regions - Resp. - SAC)
Request for Additional info.(From alleger, licensee, others - Resp. - AOC) l Closecut Letter / Memo (if no further action planned - Resp. - AOC)
Inspection (Resident / Specialist routine or reactive)
IF W&lD INVOLVED
- 1) has the individual been informed of the DOL process and the need to file a complaint within 180 days Yes No (has DOL information package been provided?)
- 2) has the individual filed a complaint with DOL Yes No l
- 3) if the complainant filed directly with DOL, have they been Yes No l contacted to obtain their technical concerns (Resp. - SAC) l
- 4) is a chilling effect letter warranted: Yes No (DOL finding in favor of alleger)
(conciliation w/ licensee prior to DOL decision) i 1
l I
l
. .c . ...
... i
'o (-
I 1
ADDITIONAL NOTES:
9 k
1I acq Al-p nao UNITED STATES g_ NUCLEAR REGULATORY COMMISSION
[
j; E REGION I 475 ALLENDALE ROAD KING oF PRusslA. PENNSYLVANIA 19406-1415 l
Senior Vice President - Nuclear Pennsylvania Power & Light Company 2 North Ninth Street Allentown, Pennsylvania 18101
)
SUBJECT:
NRC INTEGRATED INSPECTION REPORT 50-387/98-01, 50-388/98-01
Dear Mr. Byram:
On March 16,1998, the NRC completed an inspection at your Susquehanna Steam Electric Station (SSES) 1& 2 reactor facilities. The inspections covered routine activities by the resident inspectors, an announced inspection by a Region 1 Operations Engineer, and an announted inspection by a Region i Radiation Specialist. The enclosed report presents the i results of these inspections. The inspectors discussed the findings of these inspections with Mr. G. Jones, Vice President Nuc!aar Operations, Mr. G. Kuczynski, General Manager !
SSES, and others of your staff, at exit meetings at the completion of each individual ;
inspection period. i During the 8-week period of inspection, your conduct of activities was characterized by safe operation and generally conservative decision making. The operator licensing inspection concluded SSES's licensed operator re-qualification program was satisfactory overall in the area of plant support, the inspector found that you continued to maintain an effective radiological controls program. Although as-low-as-reasonably-achievable (ALARA) initiatives to minimize the radiologicalimpact of hydrogen water chemistry (HWC) appeared
~
comprehensive, continued vigiiance to assess and mitigate the radiological impact of HW& -
is warranted.
Based on the results of this inspection, three apparent violations were identified and are being considered for escalated enforcement action in accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy),
NUREG-1600. These violations involve the preconditioning of the Unit 1 standby liquid control system (SLCS) prior to Technical Specification surveillance testing, the adequacy of maintenance procedures associated with the SLCS accumulator charging valve caps, and the adequacy of PP&L corrective action for depressurized SLCS accumulators identified in 1995 and 1996. Accordingly, no Notice of Violation is presently being issued for these inspection findings, in addition, please be advised that the number and characterization of apparent violations described in the enclosed inspection report may change as a result of further NRC review.
l s
+
I !
3 [.
l Mr. Robert G. Byram 2 An open predecisional enforcement conference to discuss these apparent violations has been scheduled for May 5,1998. The decision to hold a predecisional enforcement conference does not mean that the NRC has determined that a violation has occurred or I that enforcement action will be taken. This conference is being held to obtain information to enable the NRC to make an enforcement decision, such as a common understanding of l the facts, root causes, missed opportcaities to identify the apparent violation sooner, !
l corrective actions, significance of the issues and the need for lasting and effective corrective action. As such, we expect you to address the Technical Specification, Final Safety Analysis Report, and Anticipate Transient Without Scram (ATWS) bases for the 1
, SLCS and whether or not you would have met those bases during the pariods of interest. l l Your discussion of the SLCS should include the specific nature and degree of the SLCS l
degraded condition. In addition, this is an opportunity for you to point out any errors in our ,
inspection report and for you to provide any information concerning your perspectives on 1) the severity of the violations,2) the application of the factors that the NRC considers when i it determines the amount of a civil penalty that may be assessed in accordance with
! Section VI.B.2 of the Enforcement Policy, and 3) any other application of the Enforcement Policy to this case, including the exercise of discretion in accordance with Section Vll. :
You will be advised by separate correspondence of the results of our deliberations on this matter. No response regarding these apparent violations is required at this time.
l In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter l and its enclosures will be placed in the NRC Public Document Room.
Sincerely, Original Signed By:
Charles W. Hehl, Director Division of Reactor Projects l Docket Nos.: 50-387;50-388 License Nos: NPF-14, NPF-22
Enclosures:
- 1. Inspection Report 50-387/98-01, 50-388/98-01
- 2. Enforcement Policy: Section V, "Predecisional Enforcement Conferences" l
4 l
Ly 9'. ,j.
l-Mr. Robert G. Byram -3 i l
L cc w/ encl:
i G. T. Jones, Vice President - Nuclear Operations -
j
. G. J. Kuczynski, General Manager J. M. Kenny, Supervisor, Nuclear Licensing G. D. Miller, General Manager - Nuclear Engineering R. R. Wehry, Nuclear Licensing i P. Ray, Nuclear Services Manager, General Electric -l C. D. Lopes, Manager - Nuclear Security A. M. Male, Manager, Nuclear Assessment Services .
H. D. Woodeshick, Special Assistant to the President
'J. C. Tilton, lil, Allegheny Electric Cooperative, Inc.
' Commonwealth of. Pennsylvania 1
i of l
1 I
4 i
l
.. . ~- . - - -. ._. - - _ . . . . . . . - . . -. - - -._ _. -. - -
i i .* - ,j
{
U.S. NUCLEAR REGULATORY COMMISSION REGION I l
1 Docket Nos: 50-387, 50-388 License Nos: NPF-14, NPF-22 l
Report No. 50-387/98-01, 50-388/98-01 Licensee: Pennsylvania Power and Light Company 2 North Nin^h Street Allentown, Pennsylvania 19101 Facility: Susquehanna Steam Electric Station Location: P.O. Box 35 Berwick, PA 18603-0035 Dates: January 20,1998 through March 16,1998 l
Inspectors: K. Jenison, Senior Resident inspector B. McDermott, Resident inspector J. Richmond, Resident inspector J. Caruso, Operations Engineer R. Regland, Jr., Radiation Specialist '
Approved by: Clifford Anderson, Chief Projects Branch 4 Division of Reactor Projects 1
L 1
l f ll
- ~
i ,<:
,p ;
EXECUTIVE
SUMMARY
f Susquehanna Steam Electric Station (SSES), Units 1 & 2 NRC Inspection Report 50-387/98-01, 50-388/98-01 This integrated inspection included aspects of Pennsylvania Power and Light Company's L (PP&L's) oporations, engineering, maintenance, and plant support at SSES. The report
~ covers an 8-week period of resident inspection; in addition, it includes the results of announced inspections by a regional operator licensing inspector, and a regional radiation specialist.
Onerations e Opereur communications were observed to be clear, concise, formal, and in compliance with SSES operations department procedures. Shift turnovers were detailed and complete, in general, communications between plant control operators and nuclear plant operators were observed to be of good quality. (section 01.1) e- A PP&L management decision, to reduce power in response to a main generator isophase bus duct cooler leak, was well communicated within the operations department and was conservative. The licensee initiated appropriate corrective actions, no violations of NRC requirements occurred, and the failure was documented for maintenance rule tracking purposes. (section 01.2) .
o Operators were observed to respond well to control room alarmed conditions.
Appropriate SSES procedures were adhered to, operability and impact on plant equipment were controlled, and actions were adequately announced and l documented. Operators identified a slow speed drift of one reactor recirculation i pump, on two separate occasions, and responded well to these anomalies. (section 01.3)
I e The licensee's approach to the establishment of alarm setpoints for safety relief '
valves (SRVs), compensatory measures for a Notice of Enforcement Discretion on the "S" SRV, and the control of SRV operability, were acceptable. (section O2.1) i i
e PP&L's corrective actions for three procedure violations, associated with the June 1996 "E" emergency diesel generator circuit breaker misalignment, were acceptable.
Corrective actions focused on improving operator performance, management oversight, and independent assessment. Subsequent licensee audits of operator performance were acceptable and appropriate actions were taken to validate and verify the quality of computer data used to assess operator performance. (section 04.2) e- The inspector concluded that Susquehanna's licensed operator re-qualification training program was satisfactory overall. The written examinations were adequate, but a section for five of six written examinations were weak. Examination administration was good, and operator performance was generally good with some individual operator deficiencies identified for followup. (section 05.1) ii
\
,i .i l
- A selection of Plant Operations Review Committee (PORC) and Susquehanna Review l
Committee (SRC) activities, covering a 3 month period, were reviewed. NRC determined PORC and SRC, in general, conducted in-depth reviews and demonstrated a conservative and safe approach. (section 07.2) l l Maintenance
- Four planned maintenance activities, reviewed during this period, were found to be appropriately conducted and controlled. Interviews with maintenance personnel showed the individuals involved in these activities were knowledgeable, appropriately qualified, and capable of explaining their activities. (section M1.1)
- The surveillance activities observed were adequately performed and appropriately controlled. The activities were accomplished by qualified and trained personnel. No violations of NRC requirements were identified. (section M1.2)
- The "B" Emergency Diesel Generator (EDG) test run was discontinued following receipt of an unexpected turbocharger lube oil low pressure alarm. The cause was adequately identified, and the EDG was repaired and returned to service within the time period allowed by Technical Specification. Overall, maintenance activities were adequate. (section M2.1)
- The licensee implemented several actions, in response to NRC and SSES self assessment identified issues, in the maintenance and work control programs. The performance issues include, in part, work control effectiveness, outstanding work backlog, and maintenance activity control. These actions have not been in place for a sufficient period of time to show improvement in the maintenance area.
(section M7.1)
- NRC review of additional information, regarding the Unit 1 standby liquid control system (SLCS) operability, between September 10,1997, and November 25,1997, identified three apparent violations. The apparent violations contributed to the SLCS being degraded and potentially inoperable. These apparent violations are being considered as escalated enforcement items, in accordance with the NRC Enforcement Policy. (section M8.1)
- Emergency Service Water system hot tapping maintenance activities were governed by procedures with contradictions in the method and depth of drilling and the method of foreign material exclusion. As a result, the activities were not adequately controlled by procedure. The licensee's response to the issue was acceptable and the safety impact of the inadequate maintenance practices was low, in this specific instance, the failure to provide adequate procedures for control of maintenance activities is considered a violation of minor significance, and is being treated as a non-cited violation. (section M3.1) iii
.- ./ i Enaineerino e NRC identified three control room annunciators which alarm after Technical Specification (TS) Limiting Condition for Operation (LCO) action levels are exceeded. '
The issue was discussed with operations management and it was determined the generalissue of annunciator conservatism, including LCO action statement start time, was being addressed in the PP&L corrective action system. Several examples '
of unalarmed TS entries were identified by the NRC, but no violations of the TS allowed outage time were identified. (section E1.1) i
- On February 2,1998, SSES requested and received a Notice of Enforcement Discretion (NOED) for containment penetration leak rate tests that were not performed when required. The licensee's request and immediate corrective actions for the issues were adequate. The licensee's initial NOED commitments were verified to be complete and an unresolved item was opened, pending information on ,
the circumstances which led to this event. (section E1.2) '
1
- The inspectors identified a floor hatch in the reactor building which was maintained open for many years in response to the inspectors questions, PP&L determined the j site tornado analysis assumed the hatch was closed. No safety evaluation was '
performed prior to placing the hatch in other than the analyzed position. A subsequent PP&L calculation determined the result of the tornado analysis was not adversely affected by hatch position. The failure to perform a safety evaluation prior to changing the hatch position was a violation of minor significance and is being treated as a non-cited violation. (section E8.1) 1
- Auxiliary System Operators were not consistently performing radwaste control room panel alarm tests and PCO performance issues were identified regarding performance of main control room annunciator alarm tests in the same time r.,riod when VIO 50-387,388/96-270-01022 was issued. These issues are being treated as a non-cited violation (Section E2.3 and E2.4).
Plant Suonort
- The as-low-as-reasonably-achievable (ALARA) organization was effectively evaluating and implementing radiation dose reduction measures and the health physics staff effectively used the employee ALARA concern program. Although ALARA initiatives to minimize the radiological impact of hydrogen water chemistry (HWC) appeared comprehensive including the implementation of condensate filtration, shielding up-grades, contingencies for chemical decontamination, and improvements in work practices and scheduling, continued vigilance to assess and mitigate the radiologicalimpact of HWC is warranted. A strong commitment to reducing plant contamination was evidenced by the reduction of recoverable-contaminated areas in 1997 from 9.4 to 6.2 percent and performance of a self-assessment in contamination controls. Health physics equipment and facilities were well maintained. Housekeeping and material conditions of plant structures and equipment were good. The condition reporting system was effectively used to identify, evaluate, and resolve radiological control program deficiencies. (section R) iv
i is i _ of I TABLE OF CONTENTS t
/ . .
i . Oper ations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 '
01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 i 01.1 Operator Communications and Shift Turnover . . . . . . . . . . . . .. . 1 01.2 Unit 2 Power Reduction for Isophase Bus Duct Cooler Repair . . . . . 2 01.3 Operator Response to Alarmed and Unexpected Conditions . . . . . . 3 02 Operational Status of Facilities and Equipment ................... 3 i O2.1 Safety Relief Valve Operability .........................3 !
04- Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 4 !
04.1 Operability Determinations ............................ 4 '
04.2 Non-Licensed Operator Performance . . . . . . . . . . . . . . . . . . . . . . 5 :
05 Operator Training and Qualification ...........................7 ;
05.1 Licensed Operator Re-qualification Training Program .......... 7 07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 i 07.1 Quality Assurance Audit of Personnel Training and Qualifications . 10 j 07.2 Safety Review Committee Activities .................... 10 l
08- Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 ;
08.1 (Update) URI 50-388/97-10-01, TS 3.0.3 Entry for Surveillance i Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 08.2 Licensee Event Report Review . . . . . . . . . . . . . . . . . . . . . . . . . 13 l 08.3 Followup of Open items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
- 11. Maintenarice . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 8 L M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 1 M1.1 Preplanned Maintenance Activity Review . . . . . . . . . . . . . . . . . 18 )
l M1.2 Surveillance Test Activity Sample Reviews . . . . . . . . . . . . . . . . 18 i
! .M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . 19 L . M2.1 EDG Turbocharger Lube Oil Low Pressure Alarm - Failed Maintenance Te st . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 9 M3 Maintenance Procedures and Documentation ...................20 l M3.1 Hot Tapping of Safety Related Piping . . . . . . . . . . . . . . . . . . . . 20 i M7 . Quality Assurance in Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 l- M7.1 Review of Maintenance Department Performance . . . . . . . . . . . 22 M8- Miscellaneous Maintenanca issues . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 M8.1 (Closed) URI 50-387,388/97-10 Unit 1 Standby Liquid Control Accumulators Found Depressurized . . . . . . . . . . . . . . . . . . . . . 23 M8.2 Followup of Open items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 Ill . Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 E1. Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 E1.1 Control Room Annunciator Setpoints and TS Entry Conditions . . . 26 E1.2 Primary Containment Penetration Leak Rate Testing - Notice of l Enforcement Discretion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
- E2 Engineering Support of Facilities and Equipment .................28 l
E2.1. "B" Reactor Recirculation Pump Speed Drift . . . . . . . . . . . . . . . 28 E2.2 Engineered Safeguards dystem Transformer Local Panel Alarm Tests )
) ..............................................29 E2.3 Radwaste Control Room Panel Alarm Tests . . . . . . . . . . . . . . . . 30 t
7 v i i
- j i
.i .;
E2.4 Main Control Room Annunciator Alarm Tests ..............31 E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 E8.1 Followup of Open items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 IV. Plant Support . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 R1 Radiological Protection and Chemistry Controls .................33 R1.1 - As-Low As-Reasonably-Achievable (ALARA) . . . . . . . . . . . . . . 33 R1.2 Control of Radioactive Material and Contamination ..........34 R1.3 Hydrogen Water Chemistry - Preparation and Planning . . . . . . . . 35 R2 Status of RP&C Facilities and Equipment ......................36 R8 RP&C Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . . 37 R7- Quality Assurance in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . 38
) R8 Miscellaneous RP&C issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 R8.1 Final Safety Analysis Report Review .................... 39 R8.2 Hydrogen Water Addition Modification . . . . . . . . . . . . . . . . . . . 39 V. M anagement Meeting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 X1 Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 vi
._ . ____.m _ _ _ _ . ~ . _ . . _ _ - - . _ _ _ _ _ _ _ . . _ . . _ _ _ .. _ .
- ._ a ' .. j ,
i l'
Report Details Summarv of Plant Status Susquehanna Steam Electric Station (SSES) Unit 1 operated at 100% power throughout the inspection period, except for four minor power reductions and one larger power i reduction. On February 6,1998, power was reduced to approximately 75% for a control l
rod sequence exchange, and modification work on the reactor feed pumps; power returned to_100% on February 7,1998.
l SSES Unit 2 was operating at 100% power at the beginning of the inspection period. On ;
January 10,1998, power was reduced to approximately 70% for one day to make a control rod sequence exchange. On January 27,1998 an unplanned power reduction to 67% power was made, to support corrective maintenance on an isophase bus cooling bus duct cooling heat exchanger (see section 01.2). On January 29,1998, the unit was allowed to coast from 100% power to approximately 95% power, prior to changing a rod !
pattern. On March 6,1998, power was reduced to approximately 70% to perforr i rod j pattern adjustment.
- 1. Operations 01 Conduct of Operations '
01.1 Ooerator Communications and Shift Turnover
. a. Innoaction Scone (71707)
During control room observations, the inspectors observed shift turnovers and communications between plant control operators (PCOs), nuclear plant operators (NPOs) and unit supervisors (USs).
- b. Observations and Findings Operator communications were clear, concise, formal and in compliance with SSES operations department procedures. Shift turnovers were observed to be detailed and complete.
The inspectors discussed plant conditions with oncoming PCOs and USs following shift turnovers and determined that sufficient information and status was transferred to the oncoming shift to ensure the safe operation of the units, in general, communication between PCOs and NPOs was observed to be of good quality.
- c. Conclusions
' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topics.
l:
.* ./
2 1
Operator communications were observed to be clear, concise, formal, and in compliance with SSES operations department procedures. Shift turnovers were detailed and complete. In general, communications between plant control operators and nuclear plant operators were observed to be of good quality.
01.2 Unit 2 Power Reduction for Isonhase Bus Duct Cooler Reoair
- a. Insnaction Scone (71707)
On January 27,1998, Unit 2 reduced power to approximately 67%, in order to repair a leek in a main generator isophase bus duct cooler. The inspectors reviewed
/
operator actions and the event. A
- b. Observations and Findinas The inspectors discussed the rational used by the Shift Supervisor (SS) to support the power reduction and observed the leak, including its temporary repair. The decision to reduce power was also discussed with the Unit 2 PCOs and US. The inspectors determined the management decision to reduce power was well communicated within the operations department and was a conservative action.
While observing the leak, the irispectors ideritified the service water supply header was vibrating in a manner that could have resulted in fatigue and the area on the connector that had failed was corroded. The Unit 1 isophase bus duct cooling supply line was observed by the inspectors and determined to not be oscillating in the same manner as the Unit 2 cooling system, nor did it have visible leakage.
The licensee issued CR 98-0282 to perform a root cause analysis and determine the corrective actions to prevent recurrence. The inspectors determined the licensee
' initiated appropriate corrective actions, no violations of NRC requirements occurred, and the failure was documented for maintenance rule tracking purposes.
- c. Conclusions A PP&L management decision, to reduce power in response to a main generator isophase bus duct cooler leak, was well communicated within the operations department and was conservative. The licensee initiated appropriato corrective actions, no violations of NRC requirements occurred, and the failure was documented for maintenance rule tracking purposes.
.. .j 3
01.3 Onarator Rannonna to Alarmed and Unarnacted Conditions
- a. Inanaction Scone (71707)
During control room observations, the inspectors observed / reviewed PCO and US response to alarmed and unexpected conditions in order to determine compliance
-with Technical Specification (TS) and SSES operating procedures.
- b. Observations and Findinos Operator responser to the following alarmed conditions were observed to be aggressive and in accordance with TSs and SSES operating procedures.
AR-G16-OO1 "E" Emergency Diesel Generator (EDG) Room Temperature AR-F02-001 Panel OC577E Local Troubh AR-015-C10 "A" EDG Local Alarm AR-231-A04 Recombiner Panel OC145 Lirouble AR-201-001 Area Radiation Monitor Panel 2C605 DNSCALE/INOP AR-102-G03 Recirculation Pump Motor Hi Temp Operators identified a slow speed drift of one reactor recirculation pump, on two separate occasions, and responded well to these anomalies,
- c. Conclusions Operators were observed to respond well to control room alarmed conditions.
Appropriate SSES procedures were adhered to, operability and impact on plaat equipment were controlled, and actions were adequately announced and documented. Operators identified a slow speed drift of one reactor recirculation pump, on two separate occasions, and responded well to these anomalies.
O2 Opercional Status of Facilities and Equipment 02.1 Safatv Relief Valva Ocarability
- a. Inanaction Senna (71707) -
During routine control room tours, the inspectors noted Unit' 1 currently has three Safety Relief Valves (SRVs) with elevated tailpipe temperatures and one with an inoperable acoustic monitor. The inspectors reviewed the licensee's approach to this condition to determine if it ensured the operability of these SRVs.
l
m __m __ . . _ . .
, .: . ; l 1
4 :
- b. Observations and Rndings '
l Elevated SRV tail pipe temperature is addressed in annunciator response procedure j AR-1/210-OO1, SRV High Temperatura. and off normal procedure ON-1/283-001, i
^ Open SRV. The licensee appropriately responded through its Industry Event Review Program (IERP) to NRC Information Notice IN 95-47, Unexpected Opening of a Safety Relief Valve.
! The inspectors reviewed the licensee's setpoint documentation (GEK-R1-1095, ;
i dated 10/20/82) and found that the alarm setpoint, associated with AR-210-001, !
was appropriately justified and that the indication of weeping below 250 degrees
- fahrenheit (*F) is not considered by General Electric (GE) and the licensee to be an i j indication of valve degradation. The setpoint documentation states that tests have -
4 shown that 20 lbs/hr of safety relief valve leakage is acceptable for continued plant )
i operation and can be detected by a setpoint of 250 'F. '
The inspectors reviewed the licensee's compensatory measures for the Unit 1 "S" i SRV and found them to be adequately implemented in accordance with the Notice ;
of Enforcement Discretion dated September 11,1997, i l
i
- c. Conclusions '
The licensee's approach to the establishment of alarm setpoints for safety relief valves (SRVs), compensatory measures for a Notice of Enforcement Discretion on the "S" SRV, and the control of SRV operability, were acceptable.
04 Operator Knowledge and Performance 04.1 Operabahty Determinations
- a. Inanaction Scone (71707)
The inspectors reviewed a sample of operability determinations to determine whether potential degraded conditions were identified, characterized, and resolved in a manner commensurate with their importance to safety.
- b. Observatiois and Findinas
' Seventeen initial operability determinations were reviewed by the inspectors, in general, the operability determinations were found to be acceptable. The inspector raised questinns on two of the operability determinations.
CR 97-2018 identified a potentially significant long term degradation of an emergency diesel generator (EDG) main drive chain. Maintenance activities determined the main drive chain and attached cotter pins were worn. The l operability determination states this condition, if it had been allowed to remain uncorrected, would have resulted in a failure of the main drive chain, and have caused extensive diesel engine damage. The EDG defective parts were replaced.
The inspectors noted that there was no review of the remaining EDGs in the '
l l 0 .' e j l
5 operability determination. The inspectors determined there were no current ,
operability concerns with the remaining EDGs based on a review of completed and '
ongoing maintenance activities. However, the lack of any generic considerations for the other EDGs in the root cause of the operability determination existed, until questioned by the NRC.
CR 98-0491 identified a degraded condition of the "E" EDG day tank. The day tank level slowly decreases, requiring make-up fuel oil addition. The operability determination characterized the leak as " minor" and within the capability of the fuel oil transfer pumps. The inspectors questioned why the engineers did not quantify the leak or discuss the basis for the TS minimum day tank volume (e.g., how long the diesel must be able to run on the day tank alone without makeup). This short coming was discussed with operations management, and the operability determination was revised. The inspectors reviewed the revised operability determination and found that it adequately addressed the TS and design basis for the EDG day tank.
- c. Conclusions Seventeen initial operability determinations (ODs) were reviewed and were determined to be adequate. The inspectors questioned two of the ODs. An emergency diesel generator (EDG) OD did not address the operability of the other EDGs, although they could have been subject to the same degraded condition (worn EDG drive chain). The other OD did not consider the affect of a leaking EDG day tank check valve on an associated Technical Specification requirement for EDG day tank volume. Subsequent revisions of the two ODs, questioned by the NRC, provided adequate bases for operability.
04.2 Non-licensed Goerator Performance
- a. Insoectinn Scone (71707. 92901)
The licensee's corrective actions associated with non-licensed operator performance problems identified in conjunction with the "E" EDG mis-alignment in June 1996 were reviewed. This review specifically examined corrective actions associated with escalated enforcement action VIO 50-387,388/96-270-01022 Items B.2.a, c, and d.
- b. Observations and Findinas On July 4,1996, operators identified that the "E" EDG auxiliary equipment supply breaker, at panel OA510, was not installed. Subsequently, it was discovered that a circuit breaker was mis-positioned on June 14,1996, when an NPO aligned the "E" EDG for serviro. A number of procedural violations were cited in escalated enforcement actions, issued on June 20,1997.
I l
i
6 Item B.2.a of VIO 96-270-01022 identified that an NPO failed to self check during alignment of the "E" EDG breakers at OA510 and consequently did not l align the equipment as specified in the applicable procedure. l
- ' ltem B.2.c of VIO 96-270-01022 identified that an NPO failed to notify the l control room operators after discovering a potential problem with the breaker '
alignment at panel OA510, on July 3,1996. ;
- ltem B.2.d of VIO 96-270-01022 identified that NPOs failed to perform l panel alarm tests for "E" EDG panel OC577E on numerous occasions, between January 1996 and June 1996.
i The inspectors reviewed licensee corrective actions for issues involving procedure l compliance, which included operator training, and first line leadership training. In l addition, PP&L established a number of general corrective actions. The inspectors l
concluded that the licensee's corrective actions for the specific cited violations were ;
acceptable. Several additional issues, with implications regarding personnel '
performance, were also reviewed and are discussed below.
In February 1997, questions regarding the validity of the computer records, for the panel alarm tests at panel OC577E, were reviewed by PP&L and subsequently reviewed by the NRC. The NRC review and inspection activities were documented in NRC Inspection Report (IR) 50-387,388/97-09. The NRC's review concluded that a failed reflash unit had prevented the control room alarm from reflashing on February 13,1997; despite the control room annunciator not alarming, computer )
records were available to show the panel alarm test had been performed. '
in response to the June 1997 escalated enforcement action, PP&L committed to a number of reviews and assessments. PP&L's reviews ultimately included audits of other routine activities required of Nuclear Plant Operators (NPOs), Auxiliary System Operators (ASOs) and Plant Control Operators (PCOs). Three reviews examined routine panel alarm tests required for the engineered safeguard systems (ESS) transformers, redweste control room panels, and main control room panels. NRC review of these alarm test issues and PP&L's investigations are documented in sections E2.2, E2.3, and E2.4 of this report. The NRC inspection included reviews of PP&L's audit reports, computer reports and PP&L's conclusions regarding personnel performance. The inspectors concluded the licensee's audits were acceptable and that appropriate actions had been taken to validate and verify the quality of computer data used to assess personnel performance.
The inspectors concluded that PP&L has implemented appropriate corrective actions
. to address personnel performance related vioic.tions identified in items B.2.a, c, and d of VIO 50-387,388/96-270-01022. Subsequent licensee's audits and self assessments reflect improvements which have occurred in the operations
- f o .
,f 7
department. NRC observations of current operator performance documented in NRC .
IR 50-387,388/97-10 concluded that operator performance was good. Based on !
these findings, items B.2.a, c, and d of violation VIO 50-387,388/96-270-01022 are I closed.
l 6
- c. Conclusions !
t PP&L's corrective actions for three procedure violations, associated with the June i 1996 "E" emergency diesel generator circuit breaker misalignment, were acceptable.
Corrective actions focuscd on improving operator performance, management ;
oversight, and independent assessment. Subsequent licensae audits of operator i performance were acceptable and appropriate actions were taken to validate and verify the quality of computer data used to assess operator performance. ;
l
'05 Operator Training and Qualification !
B 05.1 - Licensed Doerator Re-oumlification Trainina Prooram I
- a. Insoection Scone (71001)
The inspector evaluated the Susquehanna licensed operator re-qualification training ;
(LORT) program using NRC Inspection Procedure 71001, Licensed Operator Re- ;
qualification Program Evaluation, during the week of January 12,1998. The j inspector evaluated the adequacy of the annual operating test and biennial written l examinations, and the administration of the examinations to one operating crsw and I several staff licenses using NUREG 1021, Operator Licensing Examination Standards for Power Reactors. In addition, the inspector reviewed the procedures for maintenance and activation of operator licenses and verified that the requirements were met to reactivate inactive licenses. Administr ative procedures and documents associated with the training program and its imphmentation were also reviewed.
- b. Observations and Findinas Examination Materials The inspector reviewed six written annual re-qualification examinations (i.e., 3 reactor operator and 3 senio, reactor operator) prepared and administered by PP&L this examination cycle. Overall, the written examinations were adequate but sections on " limits and controls" for five of six written examinations had questions that were weak at testing higher cognitive levels of knowledge. This portion of these examinations contained a number (i.e.,30-40%) of direct lookup or memory level questions.
I I
J l
.. ,i l
8 '
l The job performance measures (JPM's) reviewed (4 sets) were adequate but some l
weaknesses were noted. Two of the JPMs reviewed did not have critical steps l annotated (212.004.01, " Remove RPS Set "A" from Service" and 223.009.03, ;
" Shutdown a Containment Recombiner..."). Some of the JPM sets reviewed did not i j include alternate path JPMs. In addition, a number of the JPM tasks, although I l safety significant, were not challenging requiring only one or two action steps to complete the task. The use of direct lookup or memory level questions on the written examinations and weaknesses in the JPMs will be inspector followup items.
(IFl 50-387,388/98-01-01)
The simulator scenario sets reviewed (3 sets) were acceptable.
Samnie Plan 1
The inspector reviewed two sample plans developed for the examinations administered during the week of the inspection and concluded the sample plans provided an appropriate sampling of the material taught throughout the year and adequately sampled the items specified in 10 CFR 55.
Examination Administration The inspector observed PP&L's administration of operating examinations (scenarios and JPMs) to an operating crew and several staff licensed individuals and determined examination administration was good. The PP&L evaluators used good l
techniques in administering the operating examinations. The evaluators were thorough and there were no discrepancies noted during examination administration or in the followup evaluations that documented crew and individual performance.
Ooerator Examination Performance Operator performance was generally good with some individual operator deficiencies identified on the written examinations and during performance of the simulator i' scenarios and JPMs. The licensee evaluators properly identified these deficiencies for followup and feedback to the training program.
Manaaement Oversiaht and Trainina Feedback SvttAln The inspector reviewed management observation forms for 1996 and 1997 and noted that the forms provided constructive feedback on performance. A limited number of observations were documented by senior management above the level of the operation's manager.
The inspector noted several training initiatives were implemented based on operator feedback and operations department requests. These initiatives included the team building simulator activity, plant operators training with the shift, and training on ,
refueling platform modifications. I I-
. - ~ - -- .~ - , - . - _ . , , , - .
I . ._.. _._ _ _ _ i
( ,d [ D s' '
9 i l
The inspector reviewed open simulator deficiencies. There were a total of 414 deficiencies outstanding as of the January 7,1998. There were 212 open items
(
l that had some potential for affecting simulator fidelity, but most of these appeared to be somewhat minor in nature and transparent to the operators. The backlog :
appeared to be well managed with good coordination between the operation's training department and the engineering staff maintaining the simulator to ensure simulator reliability and fidelity was maintained to support training goals.
Remedial Trainina Proaram l
Based on a review of a sample of remediation records for individuals and crews ~who '
had failed cyclic, annual operating and written examinations, the inspector determined this area was satisfactory.
Maintenance and Activation of Ooerator Licenses The inspector reviewed various training attendance, grades, and medical records and also records for six individuals who re-activated their licenses. No weaknessen were identified.
Susquehanna's instruction Ol-AD-010, Summary of Limitations, Requirements and Restrictions imposed on Operations Personnel, provided a very good and detailed summary of limitations, restrictions and requirements imposed on licensed personnel. The procedure specified that licensed operators would have to complete a refresher training program prior to returning to on-shift duties when they are absent [for more than six weeks but less than three months). The inspector concluded that this instruction was a good initiative to ensure maintenance of l operator proficiency in accordance with NRC regulations. i Examination Security and Validitg Security measures for examination development and administration were reviewed and found to be adequate. No instances of examination compromise were
!- identified.
l
! PP&L developed unique static, written, JPM and simulator examinations for each crew tested with limited overlap of examination material between examinations. In i addition, for the simulator examinations a computer data base was maintained to. !
ensure scenarios were not repeated for a crew within a two year period. The !
' inspector concluded that PP&L's examination development practices were satisfactory for maintaining examination integrity.
l 1
l; I l
._,._ -. _ _ _ _ _ _ _ ___ ~ _. . . _ _ _ _ _ . -
r _ -. _ _ _ .. . - . , _ . ._ _ _ . _ _ . . _ .____ . _ . _ _ _ _ . _ _ _ .
a .' . ;
' t 10 i
- c. Conclusions The inspector concluded that Susquehanna's licensed operator re-qualification '
training program was satisfactory overall. The written examinations were adequate
' but certain questions did not test cognitive knowledge. Examination administration i
. was good and operator performance was generally good with some individual l weaknesses identified on the written examinations and during performance of '
simulator scenarios and job performance measures. The program to remediate -
training weaknesses was satisfactory. Various training attendance records, grades, j license reactivation and medical records were reviewed with no weaknesses l identified. Security measures for examination development and administration were
]
p reviewed and found to be adequate in that no instances of examination compromise l were identified. '
07 Quality Assurance in Operations l 07.1- Quality Assurance Audit of Personnel Trainina and Qualifications (71707) l PP&L Nuclear Assessment Services (NAS) Audit 96-139 was reviewed to determine if significant operator qualification or training deficiencies were identified. The l- inspectors determined the audit was adequately performed, findings and l observations were fully communicated to SSES management, and SSES management aggressively completed the appropriate initial corrective actions.
07.2 Safety Review Committee Activities
- . a. Insnection Scoon (71707) f TS 6.5.1 and 6.5.2 establish the requirements for the Plant Operations Review ;
Committee (PORC) and Susquehanna Review Committee (SRC), respectively. The !
activities of the PORC and the SRC were reviewed / observed for a three month j period ending March 16,1998. A review was conducted to determine whether their o
activities were aggressive in seeking out areas needing improvement, and were overall effective.
'b. Observations and Findinas The inspectors determined, in general, the PORC and SRC conducted in-depth L reviews and demonstrated a conservative and safe approach to power operation.
PP&L management made conservative presentations to the committees and received, when appropriate, recommendations to improve safety and compliance.
- c. Conclusions A selection of Plant Operations Review Committee (PORC) and Susquehanna Review
- Committee (SRC) activities, covering a 3 month period, were reviewed. NRC I
determined PORC and SRC, in general, conducted in-depth reviews and
- . demonstrated a conservative and safe approach.
I.
I
!. ,._ _ - _ _ . . . _ , = ._ _ _
m __ - _ _ - . . _ . . _ _ . _ . _ . _ . . __._ _ _ _ _ _ _ _.__ _ _ _
11 08 Miscellaneous Operations issues 08.1 (Undatal URI 50-388/97-10-01. TS 3.0.3 Entry for Surveillance Activities a inanection Scone (71707. 92901) .
The inspectors reviewed the licensee's corrective actions and observed portions of a PORC meeting that reviewed an October 16,1997, entry into TS 3.0.3, for 4 performance of a surveillance test. '
- b. Observations and Findings On October 16,1997, the Unit 2 "A" rod block monitor (RBM) was taken out of service and declared not operable to support performance of a surveillance test.
During the surveillance, a count circuit output was observed to be below its expected value. The licensee stopped the surveillance, initiated work authorizations to investigate and repair the circuit, and initiated a CR. After the circuit was repaired, the TS 3.1.4.3 LCO expired and the licensee entered TS 3.0.3 to continue the surveillance. '
Inanection Related Activities This event was discussed in NRC Inspection Report (IR) 50-387,388/97-10. A request was made in an NRC letter dated February 4,1998, for SSES management to review their decision regarding the voluntary entry into TS 3.0.3. IR 97-10 identified two issues regarding the licensee's actions in response to the situation on October 16,1997:
- During the first five hours after entering TS 3.0.3, the licensee took no physical actions to initiate actions to place the unit in shutdown, in order to comply with the TS 3.0.3 action requirements.
- At the time the decision to enter TS 3.0.3 was made, the licensee was capable of performing the action required by TS 3.1.4.3(a) action statement.
~
The licensee responded to the NRC request for review by initiating CR 97-3431 and
- conducting a PORC review of the event. The licensee, in part, came to two l
, conclusions: l
- ' During the first five hours after ent' ering TS 3.0.3,' SSES did not establish a 1 clear shutdown plan. l
- L At the time the decision to enter TS 3.0.3 was made, the unit was capable of performing the required TS 3.1.4.3 (a) action statement by tripping the rod block monitor (RBM). However, TS 3.0.3 was entered to allow continued testing of the RBM.'
The above SSES management decisions were reviewed and concurred with by the
- PORC,' and reported by LER 50-388/97-07.
1 I
l m, -
+
, f.
12 i The inspectors reviewed the licensee's responses to NRC questions and supporting documentation, observed in part the PORC activities, and discussed the conclusions with several SSES managers.
Previous Similar Events i
- since January 1993 were reviewed. '
Three of the LERs involved the performance of TS required emergency diesel generator (EDG) testing with one unit in an outage that created a condition not allowed by the operating unit's TS. This condition was identified by the licensee on January 3,1994, and a TS amendment was requested to resolve this issue on
-January 11,1996 (Attachment 3). However, at this time this amendment has not been approved by the NRC. The licensee also has a pending improved Technical Specification (ITS) submittal that would resolve this issue. It is anticipated that the licensee will address this issue in advance of a May 1998 outage.
Eight of the LERs involved tripping /un-tripping equipment to perform operability testing. In addition to the RBM, these issues involve TS required containment radiation monitors and secondary containment ventilation dampers. The containment radiation monitor issues were resolved with a TS amendment.
However, the secondary containment ventilation damper issue is similar to the RBM issue, in that the licensee is waiting for approval of the ITS submittal to resolve the problem.
Twelve of the LERs involved failures and/or events that placed a unit (s) in conditions not specifically addressed by the TS. The inspectors determined that these twelve, involuntary entries in TS 3.0.3 were necessary, and the causes of the events were adequately addressed by the licensee.
Findmgs With respect to the licensee's control of activities after entering TS 3.0.3, the inspectors determined the licensee's post PORC interpretations and corrective actions were adequate.
i With respect to the decision to enter TS 3.0.3, in deference to performing the actions required by TS 3.1.4.3, this part of the unresolved item will remain open until the issue is reviewed by NRR. This URI remains open. !
,.- .m.. -- - . _ - . . ~, - . , . - - , ,-,. .
)
13 ,
- c. Conclusions On October 16,1997, the licensee chose to enter Technical Specification (TS) l 3.0.3, in order to continue with a surveillance test, rather than trip the rod block ;
monitor, as specified by TS 3.1.4.3. Two issues, raised by the NRC: (1) The '
licensee's control of its activities while in TS 3.0.3, and (2) The licensee's capability ;
to choose entry into TS 3.0.3, rather than TS 3.1.4.3. Administrative control issues, identified by the NRC, for the implementation of TS 3.0.3 actions, have been l adequately addressed by PP&L. This item will remain open pending a review of '
PP&L's use of TS 3.0.3, to support surveillance testing, by the Office of NRR. ;
08.2 Licensee Event Renort Review (92700)
(Closed) LER 50-388/97-03-00 )
Required Sample was not Collected and Analyzed within Technical Specification i Time Limit l I
On March 20,'1997, a service water radiation monitor was removed from service. )'
Subsequently a service water sample required by TS was not collected and analyzed within the time specified by TS Limiting Condition for Operation (LCO) Action 3.3.7.10. The TS LCO Action states that with less than the minimum required number of radiation monitors operable, the effluent release pathway may continue for up to 30 days provided that, at least once per eight hours grab samples are collected and analyzed for gross radioactivity at a specific limit of detection. The subject sample was taken and analyzed within fifteen minutes of the required eight hour period. The licensee determined that the root cause of the event was personnel error and entered the involved individual in the PP&L performance improvement process.
The inspectors performed a summary review of the LER, the associated condition report and its corrective actions. In addition, onsite field inspections were performed. It was determined there was no safety impact from the delay in taking the effluent sample, because the results of the sample were normal and as expected. Therefore, this non-repetitive, licensee identified and corrected violation is being treated as a non-cited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. This LER is closed. (NCV 50-388/98-01-02)
(Closed) LER 50-388/97-05-00 )
Reactor Building Vent Continuous Sample Lost for Twenty Minutes On March 25,1997, while the unit was shut down, chemistry technicians were performing a transfer process from the reactor building ventilation stack monitor to the system particulate iodine noble gas (SPING) system. During the transfer, a j spurious reactor building criticality monitor alarmed, requiring the evacuation of the )
area in which the technicians were working. Upon returning to the area the '
i
14 1 technicians realized that there had been an approximately 20 minute period that continuous sampling of the reactor building vent was not mabtained in accordance with TS 3.3.7.11. The licensee determined that the reactor building criticality monitor had drifted low which caused the unanticipated alarm.
The inspectors reviewed the LER, inspected the licensee's corrective actions and root cause evaluation, conducted an onsite field inspection and determined that there were no safety consequences associated with the failure to continuously monitor the stack release. There were no safety consequences because the unit was shut down and there was a clear pattern of data established both before and after the missed time period. With respect to the criticality alarm drift, the drift was in the conservative direction, and there was no significant pattern of spurious alarms. This TS violation resulted from circumstances not within reasonable licensee control, in that the criticality alarm failure could not have been avoided within the parameters of the licensee's surveillance program. Therefore, this non-repetitive violation is being treated as a non-cited violation, consistent with Section VI.A of the NRC Enforcement Policy. This LER is closed. (NCV 50-388/98-01-03)
If,joned) LER 50-387/97-013-00
(
Requirements for Testing Activated Carbon Samples
{
On June 19,1997, while both units were operating at 100% power, the licensee determined that the testing methodology used for activated carbon samples was different than that required by TS. The licensee received a Notice of Enforcement Discretion to operate until it accomplished the required testing. VIO 50-387,388/97-04-01 and a notice of enforcement discretion were issued to the licensee. The licensee responded to the violation in PP&L letter PLA-4666, dated September 4,1997, and affected adequate corrective actions which included a TS change, procedure changes, and technician training. VIO 50-387,388/97-04-01 was closed in inspection report 50-387,388/97-06, through onsite field inspection activities. This LER is closed.
(Clonadi LFR 50-387/97-21-00 and 50-387/97-21-01 Condition Prohibited by Technical Specifications - Technical Specification 3.0.3 Entry On September 21,1997, with Unit 1 at 100% power, the licensee made a voluntary entry into TS 3.0.3, in order to perform a required TS surveillance of the common refueling floor ventilation system. The licensee completed the surveillance and exited TS 3.0.3 within the allowed TS 3.0.3 LCO time limit. The technical decisions made by plant management were adequate.
The inspectors reviewed the LER, inspected the licensee's corrective actions and
. root cause evaluation, conducted an onsite field inspection and determined that the broader resolution of TS 3.0.3 entry discussed in section 08.1 of this report will encompass this issue. Therefore this issue will be treated as an example of the URI in section 08.1 of this report. This LER is closed.
08.3 Followun of Onen iterns (92901)
v.---. ., ...- .- - - ._ _ -- _ _ - - - - - _ _ _ -. _ _
a 15 !
i (Closed) VIO 50-387.388/98-270-01012 I Less Than Four Independent Diesel Generators Operable i
and the "E" EDG is an installed spare, which can be directly substituted for any one 1 i of the other four EDGs to allow maintenance during power operations. On July 4, l 1996, the licensee discovered that the NPO had incorrectly aligned an electrical
, breaker associated with the "E" EDG, such that if the "E" EDG was needed for a loss of offsite power event, the engine would have started, but its auxiliary
, equipment would not have be automatically energized, thus making the "E" EDG j inoperable.
l 3 Between June 14,1996 and July 4,1996, the alignment of the EDGs were such j that the licensee did not comply with TS. The licensee established a number of corrective actions including an event review team, an independent Safety i Engineering Group (ISEG) audit, and a QA audit. The inspectors reviewed the l- licensee's corrective ' actions, including operator training, personnel actions,
- procedure changes, physical equipment modifications, leadership training for first
- i. line supervisors, and several other programs. The licensee's corrective actions were
- adequate to address the TS violation for failure to maintain the required number of
- j. operable diesel generators. This violation is closed.
(Undated) VIO 50-387.388/96-270-01022 i j Failure to implement Procedures as Required by TS 6.8.1 Items P.2.a, c,'and d are discussed in detail in section 04.2 of this report, and are closed. Other items of this escalated enforcement violation remain open. 3 (Closed) VIO 50-387.388/96-270-02013 Containment isolation Valve Open and Deactivated for 24 Hours I During a Unit 1 core spray system outage, the licensee back-seated and de-energized the only containment isolation valve (redundant isolation boundary l provided by a closed system) in a test return line. While the valve was back seated, I its packing was replaced, further affecting the pressure boundary integrity of the valve.
The root cause for the event was an operational decision based on a SSES technical specification interpretation (TSI) that was in conflict with the wording of TS 3.6.1.
The licensee's corrective actions were reviewed and determined to include upgrades to the TSli generic procedure reviews, and engineering training. The inspectors j determined that the licensee's corrective actions were adequate. This violation is i closed.
'e.m*
- . . ~ . . . - .. - . - - - _ . _ _ - - - . - - . _ _ . . . - . _ _ ~ . ~
- e. . .. ;
16 l (Closed) VIO 50-387.388/96-270-03014 I Standby Liquid Control System Heat Trace De-energized
(
On June 12,1996, a NPO de-energized circuit breakers labeled as the normal and j alternate power supplies for the "A" standby liquid control system (SLCS) pump .
heat trace, in preparation for work on the inoperable "A" pump. A day later, the i licensee discovered that the NPO had erroneously de-energized heat trace for the "B" train SLCS pump which was still required to be operable. The licensee's corrective actions were documented on a WA and CR.
The inspectors reviewed the corrective action associated with CR 96-0705 and work authorization (WA) S60168. The licensee's corrective actions included counseling the involved operators, and additional operations department training on the importance of applying status control tags. The licensee's corrective actions ;
were adequate. This violation is closed.
(Closed) VIO 50-387.388/97-04-02 Nuclear Safety Assessment Group NRC IR 50-387,388/97-04 addressed non-compliances with respect to the Nuclear Safety Assessment Group (NSAG) staffing. The NRC requested, as part of the licensee response to tha violation, a discussion of the licensee's planned corrective actions to review and reconcile the activities preformed by NSAG. The inspectois reviewed the licensee's response and corrective actions.
PP&L responded to the notice of violation in a letter dated September 4,1997 (PLA-4666), stating that.two additional persons, with a bachelor's degree in engineering, had been assigned as dedicated full-time NSAG engineers to assure TS compliance, !
and that PP&L was in full compliance. The inspectors verified that for the period of January 2,1998 through March 1,1998, that the NSAG was adequately staffed. l l
In addition, several issues related to the violation were identified, including I management control and oversight of NSAG. These included organization changes that could compromise the independence of the NSAG function. The NRC requested that the licensee review and reconcile these issues. The licensee l performed a review and determined the NSAG function to be independent and unaffected by the identified issues. The inspectors performed additional reviews and did not identify an instance where the identified issues affected the safe operation of the units. Corrective actions in response to a violation for inadequate staffing of the independent Safety Engineering Group (ISEG) was acceptable. This violation is closed.
! (Closed) VIO 50-387.388/97-04-03 Quality Assurance Program (QA) Changes in February 1995, the licensee made changes in the accepted QA program, without prior NRC approval, that affected the span of control of the manager of QA. As part of its corrective actions, the licensee completed an evaluation of the QA program
- changes, in accordance with NASP-QA-104, Evaluation of Proposed Changes to the I -- .
. : . i 17 Quality Assurance Program Description. The inspectors reviewed the licensee's corrective actions and determined the corrective actions, which included a safety evaluation, an QA organizational chsage and QA program implementing procedure ,
changes, were acceptable. This violation is closed.
(Closed) URI 50-387.388/97-06-04 Control Room Emergency Outside Air Supply System (CREOASS) Operability l On July 24,1997, the licensee identified an open intake plenum access door on the ;
"A" train of CREOASS, which was contrary to the design of the system. The '
impact on system operability, licensee's operability determination, and corrective '
actions were reviewed in NRC IR 50-387,388/97-06.
Three issues were identified to the licensee in an NRC letter, dated October 27, 1997, and responded to in a PP&L letter, PLA 4839, dated February 5,1998. The three issues addressed the initial cause of the open CREOASS door, the. j consequence.s of the open door on system operability, and the control afforded the positioning of the door by operations personnel. In its response, the licensee was 4 not able to determine the definitive cause of the open CREOASS door, but determined the system would have operated at its design capacity following initiation, and agreed that the control of CREOASS panel doors needed improvement.
The inspectors evaluated the licensee's corrective actions and written response to I the NRC letter. The licensee's discussions of operability and root cause were adequate.- The licensee's control of the CREOASS doors prior to the event was determined to have been inadequate to ensure TS 3/47.2 system operability. ,
Because of the lack of consequence in this particular case, the response of the panel - l doors following system actuation and adequate licensee corrective actions, the !
failure to establish and implement controls to ensure the position of CREOASS doors, in support of TS 3/4.7.2 operability requirements, was considered a violation ;
of minor significance and is bt.ing treated as a non-cited violation, consistent with Section IV of the NRC Enforcement Policy. This unresolved item is closed. (NCV 50-387,388/98 01-04)
Subsequent to the discovery, the licensee established adequate controls on the positioning of plenum access doors for CREOASS and other HVAC systems The license has implemented admirdstrative controls to enter the appropriate LCO when the systems are breached. In addition, the licensee will block the automatic start of the CREOASS system, when personnel have the system breached and are performing activities within the system plenum.
i t
ls .
l 18
- 11. Maintenance
.M1 Conduct of Maintenance M1.1 Preolenned Maintenance Activity Review l a. insnaction Scone (62707) l l The inspectors observed / reviewed selected portions of pre-planned maintenance activities, to determine whether the activities were conducted in accordance with NRC requirements and SSES procedures.
- b. Observations ard Findinas l
Maintenance activities performed by the following WAs were observed / reviewed during this inspection. In addition, selected personnel qualifications, equipment, permits (tagouts), procedures, drawings, and/or vendor technical manuals associated with the maintenance activities were also reviewed. *
\
l V80190 "A" EDG 5 Year Maintenance S80379 "B" EDG Fuel Oil Drain Line
.V80011 Local Power Ra'nge Monitor L S86011 "B" EDG Lubrication Oil Pressure Indication interviews with maintenance personnel showed the individuals involved in the-maintenance activities to be knowledgeable and capable of explaining their function.
- c. Conclusions l
Four planned maintenance activities, reviewed durire this period, were found to be appropriately conducted and controlled. Interviews with maintenance personnel showed the individuals involved in these activities were knowledgeable, appropriately qualified, and capable of explaining their activities.
' M1.2. Surveillance Test Activity Samole Reviews l j
- a. Insoection Scone (617261 The inspectors observed / reviewed selected portions of pre-planned surveillance
. activities, to determine whether the surveillance tests conformed to TS requirements and SSES administrative requirements.
H _ ~ _
__ . _ _ . _ _ . _ . . _ _ _ . _._ ___ . . _ _ _ ~ _ _ . - . _ _ _ _ _ ~ _
l
, . , l 19
- b. Observations a~n d Findinas Portions of the following pre-planned surveillance activities were observed / reviewed:
SC-134-104 System Particulate lodine Noble Gas Quarterly Functional Radiation Monitor -lodine Channel S0-024-001 "A" EDG Monthly Operation SO-156-001 Weekly Control Rod Exercising !
SO-024-013 Class 1E Operability Test ' !
SC-234-104 System Particulate lodine Noble Gas Quarterly Functional I Radiation Monitor - lodine Channel SI-283-320 Quarterly Calibration of.Condensor Vacuum SI-155-302 - 18 Month Calibration of Control Rod Scram Accumulator Leak Detectors . i l' SO-149-A02 "A" Loop RHR Full Flow Test i S0-013-010 Moathly Fire Protection System Valve Alignment Check i
The activities were determined to conform to the requirements of TS and satisfied !
PP&L administrative requirements (approvals, personnel qualifications, scheduling l and permits). Equipment was properly removed from service and, when appropriate the TS LCOs were documented and met.- The surveillance activities were determined to have been accomplished by qualified and trained personnel.
- c. Conclusions The surveillance activities observed were adequately performed and appropriately controlled. The activities were accomplished by qualified and trained personnel. No violations of NRC requirements were identified.
M2 Maintenance and Material Condition of Facilities and Equipment M2.1 EDG Turbocharoer Lube Oil Low Pressure Alarm - Failed Maintenance Test a,. Insoection Scone (62707)
A maintenance test run on the "B" EDG was discontinued and a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO was entered following receipt of an unexpected turbocharger lube oil low pressure alarm.
The inspectors observed / reviewed portions of the troubleshooting and restoration,
- b. ~ Observations and Findinas On February 9,1998,'a maintenance run was commenced on the "B" EDG, as part
. of a post-modification test, following replacement of the fuel oil crossover drain line.
Shortly after the start of the run, the turbocharger tube oillow pressure alarm was p
lN p
,.* i.
- . \
20 ,
received at the local EDG control panel. A local pressure gauge indicated the turbocharger lube oil pressure to be lower than normsl. Operations personnel ,
shutdown the EDG using the emergency stop. The "B" EDG was declared 1 inoperable and SSES Unit 1 and Unit 2 entered a TS 3.8.1.1(b) 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO.
l l
_ The inspectors reviewed / observed a turbocharger boroscope inspection, instrument calibration checks and re-calibration, functional checks of the turbocharger tube oil pressure regulating valve, and turbocharger lube oil filter inspection and replacement.
The indicated low oil pressure and alarm, seen by the operator at the local EDG control panel, was attributed to air intrusion into the oil instrument line. Two other L
utilities, contacted by the licensee, which have similar Cooper-Bessemer model L diesel engines, stated they had experienced similar low oil pressure indications and l low oil pressure alarms due to air intrusion into the oil instrument line. Cooper-Bessemer also stated air intrusion would not result in the actual turbocharger oil i pressure being below minimum requirements.
The inspectors observed that instrument calibration checks were performed as minor maintenance activities. Overall, maintenance activities were adequate. No violation of NRC requirements were identified.
- c. Conclusions The "B" Emergency Diesel Generator (EDG) test run was discontinued following receipt of an unexpected turbocharger lube oil low pressure alarm. The cause was i adequately identified, and the EDG was repaired and returned to service within the ;
time period allowed by Technical Specification. Overall, maintenance activities were '
adequate.
M3 Maintenance Procedures and Documentation M3.1 Hot Taooina of Safety Related Pioina
- a. Insosction Scone (82707)
The inspectors observed / reviewed selected portions of pre-planned maintenance activities related to modification of the Emergency Service Water (ESW) system.
The inspections were intended to determine whether the activities were conducted in accordance with NRC requirements and SSES procedures.
- b. Observations and Findinas The selected maintenance items included boring activities on safety related concrete wall structures, and modification activities on ESW system piping. The licensee implemented the ESW pipe modifications to attach chemical addition tubing to a main ESW header, by using a " hot tapping" connection. The hot tap procedure
,: c f
21 welds a piping connection onto a header pipe, then uses a special drilling rig, with a pressure retaining drill bit assembly, to drill into the pressurized header pipe. The attached piping connection was intended to be used to supply biocide treatment to the ESW header.
Maireenance activities were authorized and conducted by the following:
MRP-QA-3806, Hot Tapping of Piping Systems WA C73282, Hot Tap of Division i ESW Piping IP-100, Core Drilling Machine Vendor Manual DCP 96-9048, ESW Design Change Package Maintenance personnelinvolved in these activities were knowledgeable of their assigned duties.
There were several contradictions identified by the inspectors between the governing procedures, the vendor manual and field pract_ ices, for performance of hot tapping activities.. The contradictions included calculation errors, contradicting methods and depths of drilling, contradicting methods of foreign material exclusion (FME) control in the drilled cavity, and different applications of vendor supplied materials. Because of these discrepancies, the inspectors determined the observed maintenance activities on safety related piping were not adequately controlled by j
- procedure, contrary to the requirements of TS 6.8.1. Because of the reasons j discussed below, this failure constitutes a violation of minor significance and is i being treated as a non-cited violation, consistent with Section IV of the NRC Enforcement Policy. (NCV GO-387,388/9841-05)
The inspectors' observations were discussed with SSES maintenance department -
management. Adequate changes were made to the procedures controlling hot tapping activities by procedure revisions to MFP-QA-3806, on February 24,1998.
The licensee's corrective actions were viewed as prompt and appropriate. No unreviewed safety question was identified. Because ESW is a low pressure raw water system, there was no safety impact. This finding was not considered a precursor to a more significant event.
- c. Conclusions Emergency Service Water system hot tapping maintenance activities were governed by procedures with contradictions in the method and depth of drilling and the
. method of foreign material exclusion. As a result, the activities were not adequately controlled by procedure. The licensee's response to the issue was acceptable and the safety impact of the inadequate maintenance practices was low, in this specific
- instance, the failure to provide adequate procedures for control of maintenance activities is considered a violation of minor significance, and is being treated as a non-cited violation.
2
- .' s .-
22 M7 - Quality Assurance in Maintenance M7.1 Raview of Maintenance Denartment Performance
- a. Insoection Scone (62707)
The inspectors reviewed SSES actions to improve maintenance controls.
- b. Observations and Findinas Performance issJes exist in Outstanding Work (corrective action and maintenance authorization) Sacklog, maintenance activity control, work implementation and completion, and work control effectiveness. Based on the performance of plant equipment, the inspectors determined the licensee is appropriately prioritizing corrective maintenance and has adequate control of the corrective maintenance backlog. Based on the licensee's lack of success in implementing work as planned, the inspectors concluded the licensee continues to be challenged with the scheduling and planning aspects of maintenance. However, the inspectors found no discernable impact on operational safety due to the current maintenance and corrective action backlog.
The inspectors determined that SSES management has recognized that performance issues exist in the maintenance and work control areas and has initiated actions to address these performance issues. The inspectors reviewed these actions and discussed them in detail with SSES management. PP&L actions are aggressive and safety oriented. However, there are presently no clear results from these new actions.
- c. Conclusions The licensee implemented several actions, in response to NRC and SSES self assessment identified issues, in the maintenance and work control programs. The performance issues include, in part, work control effectiveness, outstanding work backlog, and maintenance activity control. These actions have not been in place for a sufficient period of time to show improvement in the maintenance area.
I 23 M8 Miscellaneous Mainter ance issues M8.1 (Closed) URI 50-387.388/97-10 Unit 1 Standbv Liould Control Accumulators Found Deoressurized
- a. Inanection Scone (82707)
In response to NRC violation 50-387,388/97-07-06, Standby Liquid Control System (SLCS) Accumulator Operability, the licensee discovered additional periods in which the SLCS was inoperable. These events were reported by the licensee and reviewed in NRC IR 50-387,388/97-10. The inspectors reviewed additional information associated with these issues.
- b. Observations and Findinom
Background
On November 25,1997, the Unit 1 "A" and "B" SLCS discharge accumulators were discovered by the licensee to be partially depressurized. The NRC identified several
'iesues associated with the root cause of this condition and reviewed the licensee's immediate corrective actions. On November 26,1997, PP&L identified a maintenance practice as the potential cause for the accumulators losing pressure.
After questions from the NRC, on Decoinber 2,1997, the licensee made a 4-hour ENS notification stating that preliminary results showed the low accumulator pressure Jeopardized the system's ability to perform its intended safety function.
On January 2,1998, the licensee submitted LER 50-387/97-25-00 and stated that the full safety function of the SLCS was lost. Since then PP&L has reconsidered whether the 'SLCS safety function was lost. By letter dated February 4,1998, the NRC acknowledged that PP&L was re-evaluating the issue and requested this re-evaluation be completed within 20 days. The unresolved item (URI 50-387/97 04) was opened because more information was needed from PP&L to ascertain whether the SLCS pumps were actually inoperable and whether violations had occurred.
Additional information During this inspection period, PP&L re-evaluated the as-found condition of the Unit 1 SLCS on November 25,1997. PP&L determined the SLCS pump relief valve set points are essential to their final conclusion on SLCS operability. Consequently, PP&L informed the NRC that their re-evabtion of SLCS could not be completed until after the SLCS relief valves are Lnch tested during the Unit 1 refueling outage scheduled to begin April 14,1998.
The inspectors reviewed a summary of the PP&L historical test data anc! discussed system' design and setpoint information with cognizant engineering personnel. The inspectors found that w th the accumulators de-pressurized, the expected SLCS discharge pressurs during accident conditions is 1398 pounds per square inch gauge
' (psig) and the relief valve setpoint is 1400 psig (-0, + 3%). PP&L historical relief valve bench test test data indicated that 83% of as-found setpoints would result in
I 24 the relief valves prematurely opening at < 1398 psig. However, PP&L has ,
questioned the historical bench test data because the relief valves have not lifted i prematurely during quarterly surveillance tests.
Based on the inspectors' review, several issues were identified that appear to be contrary to SSESTechnical Specifications and NRC regulations:
)
On September 19,1995, and September 22,1996, _a single Unit 1 standby liquid control accumulator pressure was discovered to be below the acceptable pressure range specified by maintenance procedure MT-053-OO3 and no condition report was initiated for this condition adverse to quality as i required by NDAP-OA-702. These issues are considered apparent violations I which led to the SLCS being degraded to the extent that a detailed evaluation is necessary to determine whether the system was operable. The {
issues e.re being considered for escalated enforcement in accordance with the l NRC Enforcement Policy. (eel 50-387,388/98-01-08)
On September 10,1997, the procedures controlling the standby liquid control system maintenance activity being performed were not adequate to ensure the accumulator charging valve cap was installed in accordance with the vendor's instructions. As a result, the caps for both standby liquid control pump accumulators were over tightened and caused the loss of accumulatar pressure discovered on November 25,1997. These issues are considered apparent violations which led to the SLCS being degraded to the extent that a detailed evaluation is necessary to determine whether the system was operable. The issues are being considered for escalated enforcement in accordance with the NRC Enforcement Policy. (eel 50-387,388/98-01-07)
On September 10,1997, the Unit 1 Quarterly Standby Liquid Control Flow Verification surveillance test procedure, S0-153-OO4, specified that maintenance personnel pre-charge the standby liquid control nitrogen accumulators to the range specified in maintenance procedure MT-053-OO3.
This activity resulted in the standby liquid control pumps being tested in a ;
condition that was different from the as-found condition, thereby potentially affecting the validity of the surveillance test results. These issues are considered apparent violations which led to the SLCS being degraded to the extent that a detailed evaluation is necessary to determine whether the system was operable. The issues are being considered for escalated i enforcement in accordance with the NRC Enforcement Policy. (eel 50- .
387,388/98-01-08) l Unresolved item 50-387/97-10& is closed.
.- . - _ ...... . . - - . - . - . . . . - . - - - . ~ ~ . - _ - - ~ . _ - - . - -.
l 25
- c. I Conclusions NRC review of'additionalinformation, regarding the Unit 1 standby liquid control system (SLCS) operability, betwun September 10,1997, and November 25,1997, identified three apparent violations. The apparent violations contributed to the SLCS being degraded and potentially inoperable. These apparent violations are being
[ considered as escalated enforcement items, in accordance with the NRC l Enforcement Policy.
M8.2 Followun of Onen items (92902)
(Closed) URI 50-387/94-14-01 and 50-388/94-15-01 Evaluation of Thermal and Pressure Locking This item was opened to track the status of PP&L's corrective actions for gate valves susceptible to pressure locking and thermal binding. On August 17,1995, the NRC issued Generic Letter (GL) 95-07 requesting licensees to formally evaluate
. pressure locking and thermal binding for motor operated valves. PP&L provided 60-day and 180-day responses to the GL on October 16,1995 and February 13,1996, respectively. PP&L's program continued to evolve during its implementation and on November 7,1996, a revision to the 180-day response was submitted to the Office of Nuclear Reactor Regulation (NRR). By letter dated April 29,1997, PP&L updated NRR on the status and schedule for additional corrective actions. The adequacy of PP&L's response to the issues of GL 95-07 are currently under review by NRR and, pending resolution of any concerns, NRR will close out the GL by direct correspondence to PP&L. No violations were identified. This unresolved item is closed.
(Closed) VIO 50-388/97-09-02 Reactor Recirculation Valve Bonnet Vent Line Failure On September 17,1997, Unit 2 was' shut down to investigate increased drywell leakage. The licensee identified the source of the drywell leakage to be a 180 l
degree through-wall crack on a reactor recirculation discharge valve bonnet vent line. The crack resulted from the failure to ensure that an adequate vent line support was maintained at the completion of maintenance activities.
The licensee responded to the violation in PP&L letter PLA 4836 dated February 2, l 1998. The inspectors reviewed / inspected the corrective actions reported in the l
- PP&L letter. The corrective actions, which included repair of the bonnet, i replacement of an associated hanger, and procedure changes, were determined to be acceptable. The inspectors reviewed the results of three non-destructive examinations (NDEs) conducted under WA C72397 and WA C73718 and two hanger inspections conducted under WA V72413.' The NDEs were intended to I
l
- - r r. - - y --
I
- 1 26 ensure that there were no additional failures on similar plant equipment. The inspectors determined that the NDEs were adequately performed and did not identify y problems similar to the failure that resulted in the Unit 2 shutdown. This violation is closed.
Bl. Engineering E1 -_ Conduct of Engineering -
E1.1 Control Room Armunciator Setooints and TS Entry Conditions l a. Innoaction Scoon (37551) ;
.i On January 18,1998, the "8" EDG was declared inoperable, when an NPO, on i rounds, discovered the air start receiver pressure was below the 240 psig requirement of TS 4.8.1.1.2.(a).7. The inspectors reviewed the root cause of this occurrence.
l-l
- b. Observations and Findinos i 1
The inspectors determined the finding was the result of good NPO round !
performance and the licensee took adequate initial corrective actions, includmg !
initiation of CR 98-0175. However, this is a repetitive occurrence (September 5, l 1995, and November 11,1996) and the air start receiver low' pressure annunciator I
- setpoint was not adequate to detect that the pressure was below the 240 psig TS j minimum pressure requirement. As a result, the control room operators were i
unaware the EDG had become inoperable, until the NPO identified the condition by observation of a local pressure gauge. l I
The inspectors reviewed other control room annunciator setpoints and alarmed conditions and determined two other annunciators (for containment hydrogen and oxygen levels) are also less conservative than the TS allowable values. - The consequence of this paring of TS and annunciator setpoints, can also result in unalarmed TS LCO entries. The inspector noted that LCO entry times were logged by the operators at the time that the alarm took place, which was not conservative.
l .
The inspectors discussed the issue with operations management and determined the general issue of annunciator conservatism, including LCO action statement start time, was being addressed in the PP&L corrective action system since 1988. '
Although several examples of unalarmed TS entry were identified, no violations of l- TS LCO requirements were identified.
I:
s
- 27 l-
! c. Conclusions i NRC identified three control room annunciators which alarm after Technical Specification (TS) Limiting Condition for Operation (LCO) action levels are exceeded.
The issue was discussed with operations management and it was determined the generalissue of annunciator conservatism, including LCO action statement start
- time, was being addressed in the PP&L corrective action system. Several examples of unalarmed TS entries were identified by the NRC, but no violations of the TS allowed outage time were identified.
E1.2 Primary Containment Penetration Leak Rate Testina - Notice of Enforcement Discretion
- a. Inanaction Scoon (375511 l
On February 3,1998, SSES requested and received an Notice of Enforcement i
~ Discretion (NOED) to allow operation without taking action in accordance with TSs )
4.0.3 and 4.6.1.2. The inspectors observed / reviewed portions of licensee activities '
associated with the NOED.
- b. Obaarvations and Findinas On February 3,1998, at midnight, the licensee entered TS 4.0.3 upon discovering it
. had not leak rate tested two primary containment penetrations in Unit 1, and five .
primary containment penetrations in Unit 2, in accordance with requirements of TS 4.6.1.2. Because of the complexity of performing the leak rate tests while at power, the licensee chose to pursue an NOED.
SSES PORC reviewed and recommended the NOED request in meeting 97-2-2 after reviewing the conective actions recommended in response to CR 98-0342. The PORC review activities were observed to be insightful and detailed. The inspectors
- observed the actions of operations management and shift supervision in support of the NOED request and determined that the actions were conservative and safety
- oriented.
The NOED was granted by the Director of Project Directorate 1-2, NRR, at approximately 4:30 p.m. on February 3,1998, based on the completion of a telephone conference with the NRC.
The inspectors reviewed / observed portions of three of the five penetration leak rate tests for Unit 2. The two Unit 1 penetration leak rate tests will be performed during the upcoming Unit 1 outage, in April-May 1998. Based on satisfactory test results
. from the Unit 1 testing, the remaining two Unit 2 penetration leak rate tests will be postponed until the next scheduled Unit 2 refuel outage (Spring 1999). Penetrations X-90A, X-90D, and X-223A were tested by surveillance procedures SE-259-109,
- SE-259-110, and SE-259-113, under WAs A80357, A80358, and A80359. Each of the three surveillance tests was completed satisfactorily and the licensee I c completed its initial NOED commitments, as indicated in PP&L letter PLA-4844. The L decision to exercise enforcement discretion by issuing an NOED does not change-1 I
I
_ . . _ _ . _m _ _ ._ __ .._. _ m.__ _ _ _ .. _ . _ _ ___.__ _ ___. _ . _ _ __ _ _ _ ,
. .' e r 28 l- the fact that a violation will occur, nor does it imply that enforcement discretion its being exercised for any violation that may have led to the violation at issue. '
Pending additional information from the licensee regarding the root cause of this event, this issue will be tracked as an Unresolved item. (URI 50-387,388/98-01-09)
- c. Conclusions On February 2,1998, SSES requested and received a Notice of Enforcement Discretion (NOED) for containment penetration leak rate tests that were not performed when required. The licensee's request and immediate corrective actions for the issues were adequate. The licensee's initial NOED commitments were verified to be complete and an unresolved item was opened, pending information on the circumstances which led to this event.
i E2- Engineering Support of Facilities and Equipment l E2.1 "B" Reactor Recirculation Pumo Somed Drift I
- a. Insoection Scoon (37551)
A design condition described by the licensee as recirculation pump speed control system drift, was reviewed by the inspector. As a result of the control system drift, the SSES Unit 1 "B" reactor recirculation pump speed was observed to be slowly changing without any operator action. In addition, the inspectors observed and reviewed the initial troubleshooting efforts for this problem.
- b. Observations and Findings On February 14,1998, the Unit 1 "B" recirculation pump speed increased by approximately 11 rpm over a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period, for no apparent reason. There was no change in the "A" recirculation' pump speed during this period. All recirculation pump and recirculation motor-generator (M-G) set parameters appeared normal. On two different occasions, during this six hour period, recirculation flow was reduced with the master recirculation flow controller; both pumps responded normally.
On February 17,1998, on Unit 1 during a planned increase in recirculation flow, to maintain 100% reactor power, a momentary reduction in "B" recirculation drive flow (e.g., a reduction in pump speed) was noted, followed by a slower than expected increase in the "B" drive flow. A 3 rpm speed decrease over a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> period was observed for the "B" recirculation pump. As before, there was no change in the "A"
. recirculation pump speed during this period. The licensee initiated CR 98-0498 and CR 98-0518 to evaluate this problem.
i y f - g , - ,~.....m. - rc_,,y y -- - -
i - .
[.
1
- l. , /
29 Since February 14, the PCO's have been closely monitoring recirculation speed and I flow, however, no speed or flow drift has been. observed since February 17. The l recirculation pumps operate at about 1500 rpm, therefore, the observed speed drift problem represents a change of approximately 0.1% per hour. SSES engineering and instrument and controls personnel analyzed the system performance data for these two events, based on a computer history review, and concluded that this was probably caused by a drift in the control signal. SSES engineering has further concluded, based on the information available, that a rapid speed change or failure of the control system is not likely. A temporary data recorder was installed in the recirculation flow control system to obtain additional diagnostic information not available from the plant computer system. SSES engineering response was adequate.
Operators responded well, to control reactor power and monitor plant parameters, on two occasions when the recirculation pump speed drift resulted in unanticipated reactivity additions. The inspectors reviewed the CR and supporting data, and discussed recirculation speed control, reactivity addition, and the resulting effects on reactor po'v.'er with operations supervision, system engineering supervision, and PCOs. The operability determination was found to be adequate. The licensee's I initial actions appeared to be reasonable and conservative. j c.- Conclusions i i
The Unit 1 "B" reactor recirculation pump speed was observed to be slowly changing without any operator action,'on two separate occasions, resulting in unanticipated reactivity additions. The inspectors reviewed the operability determination and the licensee's initial corrective actions, and found them to be adequate.
E2.2 Enaineered Safeauards Svatem Transformer Local Panel Alarm Tests (37551, 71707)
Nuclear Plant Operators (NPOs) perform a local panel alarm test for several engineered safeguards system (ESS) transformers as part of their routine round activities. The inspectors reviewed two computer reports, for ESS transformer local l . panel alarm actuation data, and a PP&L corporate audit report, dated October 15, 1997, for ESS transformer local panel alarm test performance.
l l
L t
.. .' e;e 30 The inspectors concluded the computer reports showed data recording deficiencies for some Unit 1 ESS transformer local panel alarms. A comparison of the Unit 1 and 2 computer reports with in-plant maintenance records, and system configurations, provided evidence that PP&L's conclusions were reasonable with respect to NPO local panel alarm test peiformance. After accounting for the computer data recording deficiencies, computer records covering the periods of January to July 1996, August to December 1996, and January to March 1997, all indicate NPOs were consistently performing the required ESS transformer local panel alarm tests.
Although some deficiencies were identified with the recorded computer data, the inspectors found that PP&L accounted for the computer data deficiencies and reached reasonable conclusions regarding the data. Based on the computer records discussed above, the inspectors concluded that the ESS transformer local alarm tests were consistently performed by the NPOs.
E2.3 Radwasta Control Room Panel A! arm Tests (37551,71707)
Auxiliary System Operators (ASOs) perform radwaste control room panel alarm tests as part of their routine shift activities. The inspectors reviewed computer reports and PP&L audit data dated February 10,1998, for radwaste panel alarm tests for selected periods in 1996 and 1997.
The inspectors concluded the computer reports showed data recording deficiencies for some radwaste control room panel alarms. A comparison of the Unit 1 and Unit 2 computer reports with in-plant maintenance related conditions, and system configurations, provided evidence that PP&L's conclusions were reasonable, with respect to ASO redwaste control room panel alarm test performance. PP&L's audit determined that their management expectations were not clear and had not been consistently implemented throughout all operation shifts.
Records for the period January to July 1996, indicated the ASOs were not consistently performing the radweste control room panel alarm tests, in that these tests were, at times, documented as completed but not actually performed. This issue is considered part of personnel performance problems which existed prior to July 1996. Licensee corrective actions in response to VIO 50-387,388/96-270-01022, "E" diesel generator misalignment event which occurred in the same time period, are considered applicable to this issue.
I NRC reviews of the records covering the periods August to December 1996, and January to March 1997, indicate ASOs were consistently performing the required radwaste control room panel alarm tests. The inspectors concluded that the licensee's corrective actions, with respect to ASO performance, were acceptable based on direct inspection of ASO performance, interviews of ASOs, and additional record reviews.
o
a .. m ,
31 Although some deficiencies were identified with the recorded computer data, the inspectors found that PP&L accounted for the computer data deficiencies and reached reasonable conclusions regarding the data. PP&L's conclusions and corrective actions to improve ASO performance, management oversight, and independent assessment were found to be acceptable.
Failure of ASOs to consistently perform radwaste control room panel alarm tests is being considered a non-cited violation consistent with Section Vll.B.1 of the NRC enforcement policy. These issues were identified by the licensee's processes and occurred during the same time frame and were of the same nature as the "E" EDG misalignment event (Section 04.2). The corrective actions and root cause, of failure of management to communicate expectations, from the "E" EDG misalignment event were also applicable. The "E" EDG misalignment event, in part, formed a basis for a $210,000 civil penality issued on June 20,1997. (NCV 50-387,388/98-01-11)
.E2.4 Main Control Room Annunciator Alarm Tests (37551,71707)
Plant Control Operators (PCOs) perform main control room annunciator alarm tests as part of their routine shift activities. The inspectors reviewed PP&L audit data dated December 11,1997 for performing control room annunciator alarm tests.
The inspectors determined that there was no computer data which would indicate whether specific main control room annunciator alarm tests were or were not performed. The inspector also determinr,d that the method used to accomplish the annunciator alarm tests varied from shift to shift. The licensee also concluded, that there was no computer data to support a performance review for specific main control room annunciator alarm tests. Therefore, the licensee conducted a review of main control room annunciator alarm tests using PCO interviews, Unit Supervisor (US) interviews and control room log reviews. The licensee determined that in the January to June 1996 time period there were PCO performance issues with the performance of main control room annunciator alarm tests which included, at times, the test was documented as completed but not actually performed.
The inspectors concluded that the corrective actions in response to VIO 50-387, 388/96-270 01022, "E" diesel generator misalignment event which occurred in the same time period, were applicable to this issue and that PP&L's initial actions to evaluate the PCO performance issues were conservative. In general, the inspectors concluded the corrective actions, for varied PCO alarm test practices, were reasonable and overall PCO performance was good, based on direct inspection, interviews of PCOs, and additional record reviews.
The inspectors found that PP&L accounted for the absence of supporting computer data'and reached reasonable conclusions. PP&L's conclusions and corrective actions, to improve PCO performance, management oversight, and independent assessment, were found to be acceptable.
>..o #
32 PCO performance issues regarding main control room annunciator alarm tests is being considered a non-cited violation consistent with Section Vll.B.1 of the NRC enforcement policy. These issues were identified by the licensee's processes and occurred during the same time frame and were of the same nature as the "E" EDG misalignment event (Section 04.2) The corrective actions and root cause, of failure of management to communicate expectations, from the "E" EDG misalignment event were also applicable. The "E" EDG misalignment event, in part, formed a basis for a $210,000 civil penality issued on June 20,1997. (NCV 50-387,'
388/98-01-12)
E8 Miscellaneous Engineering issues E8.1 ' Followup of Open items (37551,92903)
(Closed) URI 50-387.388/97-07-09 Reactor Building Truck Bay Hatch During a routine tour of SSES reactor buildinos, the inspectors observed that a large floor hatch, on elevation 749 of the Unit 2 reactor building, was open and appeared to have been that way for many years. A subsequent review by PP&L found that the hatch was assumed to be closed in the tornado analysis for the reactor building.
CR 97-1950 was opened to document this discrepancy and two calculations were performed to evaluate the as-found condition (EC-012-2207 and EC-012-2209).
PP&L concluded that the open hatch is an acceptable configuration.
As discussed in NRC IR 50-387,388/97-07, the licensee had not performed a 10 CFR 50.59 safety evaluation prior to placing the hatch in a position contrary to the original tornado analysis. The inspectors reviewed the results of PP&L's recent calculations and determined the tornado analysis was not adversely affected by having the truck bay hatch in the open position. No unreviewed safety question was identified. No safety impact was identified, this finding does not represent a programmatic problem, and this finding was not considered a precursor to a more significant event, This failure constitutes a violation of minar significance and is being treated as a non-cited violation, consistent with Section IV of the NRC Enforcement Policy. This unresolved item is closed. (NCV 50-387,388/98-01-10)
The inspectors identified a floor hatch in the reactor building which was maintained open for many years. In response to the inspectors questions, PP&L determined the site tornado analysis assumed 1he hatch was closed. No safety evaluation was performed prior to placing the hatch in other than the analyzed position. A subsequent PP&L calculation determined the result of the tornado analysis was not adversely affected by hatch position. The failure to perform a safety evaluation prior to changing the hatch position was a violation of minor significance and is being treated as a non-cited violation.
l'
^ *
.. *v 33 IV. Plant Support R1 Radiological Protection and Chemistry Controls R1.1 As-Low-As-Reasonablv-Achievable (ALARA)
- a. Innoaction Scone (83750)
A review was performed of the controls to maintain radiation exposures as-low-as-reasonably-achievable (ALARA). Information was gathered by reviews of ALARA evaluations written for radiation work permit 1997-0061, " Application of paint and epoxy" in the emergency core cooling system rooms on the lower elevation of the unit 1 and unit 2 reactor buildings, through discussions with cognizant personnel, and tours through the plant.
- b. OFmationa and Fir.dir.as ALARA reviews were well detailed and included person-rem estimates, work planning information, external and internal exposure controls, health physics operational concerns, dosimetry and radiological monitoring, anticipated dose rates, additional comments, a work flow synopsis, and lessons learned from previous jobs.
Examples of ALARA measures implemented for painting of the residual heat removal (RHR) rooms included use of temporary shielding, system flushes, use of long handled tools, radiation source postings, and use of pictures for briefings. "ALARA (work) in-progress reviews" were performed as the project evolved, and one notable lesson learned was that a preplanned flush of the unit 2 RHR shut down cooling (SDC) line was canceled without consultation of the cognizant ALARA specialist.
The ALARA in-progress review highlighted the need for improved communications between operations and health physics to ensure the success of future RHR SDC system flushes.
During tours of the plant, the inspector examined temporary shielding installed in the RHR rooms. Lead blankets were suspended from the upper grating and were hung beneath the RHR shut down cooling lines in the overhead of the RHR room.
Licensee records showed that the shielding reduced general area dose rates by approximately 25-35 percent. Although the shielding was installed to reduce dose rates for the painting project, approval had been obtained to allow the shielding to j remain in-place until the end of the next outage on each unit, thereby increasing the effectiveness of the shielding. Shielding packages were neat and orderly, showed evidence of detailed planning, and were noted as excellent by the inspector.
The inspector also observed a willingness of the health physics staff to use the employee ALARA concern program. A health physics technician assigned to provide job coverage activities noted that, in an ongoing effort to clean and remove stored materials from the plant, a camera had been removed from the control rod drive
. (CRD) rebuild room. One of the uses for the camera was to perform an air damper inspection. With the camera removed, personnel had to enter a posted high radiation area to inspect the air damper. In response to this observation, the health physics technician submitted an Employee ALARA Concern to the ALARA
. . - - . - - - - - - . - - - ~ . . . - - . . - . - _ - ~ . - . . - . . - - _
- .. L .' . i. ;
34 ;
i organization for review. This suggestion was accepted for evaluation. The j
- inspector concluded that this was an effective use of the employee ALARA concern program.
i
- c. Conclusions I The inspector concluded the as-low-as-reasonably-achievable (ALARA) organization ;
was effectively evaluating and implementing radiation dose reduction measures and !
the health physics staff effectively used the employee ALARA concern program. 1 l
R1.2 Control of Radioactive Material and Contamination l
- a. Inspection Scone 183750)
A selected review was performed of contamination controls. Information was gathered through tours of the facility and discussions with cognizant personnel. ;
- b. Observations and Findinos The inspector observed a noticeable reduction in areas classified as contaminated areas. Many areas hed been de-contaminated and contaminated areas had been moved back to room doors. The best examples were the emergency core cooling -
system (ECCS) rooms on the lower elevation of the unit 1 and unit 2 reactor buildings which had been de-contaminated and painted. Licensee staff indicated that goals had been established to reduce the total square footage of areas classified as recoverable contaminated areas. Records showed that at the beginning of 1997, approximately 9.4 percent of plant areas were classified as recoverable contaminated areas, and this was reduced to 6.2 percent by the end of 1997. A health physics supervisor stated that future contaminated area reduction goals included de-contaminating the refuel floor, and reducing the total square footage of recoverable contaminated areas to zero percent by the end of the year 2000.
Records for internal dose assignments resulting from bioassay showed that there were no assignments greater than 10 mrem which indicated that controls to limit internal dose had been effective.
l A commitment to improve contamination controls was evidenced by the performance of a two week contamination control self-assessment performed in ;
December 1997,
- c. Conclusions l The inspector concluded that a strong commitment to reducing plant contamination was evidenced by the reduction of recoverable-contaminated areas in 1997 from 9.4 to 6.2 percent and performance of a self-assessment in contamination controls.
- ..~ s i
, 35 i
R1.5 Hvdronen Water Chemistrv - Pranaration and Plannina i i
- a. Insoection Scone (83750) :
A review was performed of preparation and planning for hydrogen water chemistry (HWC). Information was gathered by a review of an evaluation of the radiological :
impact of HWC and efforts taken to mitigate the radiological impact of HWC.
-l
- b. Observations and Findings ,
Background
i Reactor system components fabricated of high-alloy metals are subject to stress i corrosion cracking (SCC), and this type of cracking has become increasingly evident !
in reactor vessel internal components. Industry experience has shown that cracking ;
begins after about 10 years of operation, and the start of visible SCC has been ;
identified in the welds of the core shroud and the steam dryer at SSES. The area of
)
greatest concern is the potential development of SCC in the lower regions of the reactor vessel such as in the core support structures, vessel penetrations, and vessel internal structural attachment welds. In these areas, access for inspection is limited and the methods and technology to repair damage has not i,een fully developed and demonstrated. Research has shown that injection of hydrogen into reactor feedwater could reduce SCC by inhibiting the radiolysis of water and promoting recombination of radiolytic components. Potential savings with HWC were calculated by estimating the avoided cost to repair potential damage due to SCC minus the cost to install and operate HWC. This calculation forecasted that operating under conditions of HWC would result in significant monetary savings over the life of the plant, and based on that information a decision was made to install HWC in both units during the next two refueling outages. "
HWC Radiological Imnact N-16 activity is produced in the reactor coolant by the O(n,p)N reaction in the core region. N-16 has a short half-life (7.1 seconds), but emits high energy gammas upon decay. Under HWC, as hydrogen concentration increases, volatile forms of nitrogen such as NH3 are produced and swept from the reactor with main steam.
Licensee staff predicted that dose rate in steam related areas would increase by a factor of five during power cperata. This is expected to increase radiation dose to personnel working in and around the turbins building, and for personnel working adjacent to steam related areas outside the radiologically controlled area (RCA). A notable impact is that many personnel who previously received non-detectable radiation dose (i.e., radiation dose less than the 10 mrem detection limit of the thermoluminescent dosimeters (TLDs) used for personnel monitoring), were predicted to receive reported radiation dose under conditions of HWC. The most j significant dose impact was predicted to come from increased radiation levels associated with primary systems. Under HWC, changes in primary water chemistry l
cause in-core corrosion product layers to loosen, resulting in the re-distribution of '
activated corrosion products to out-of-core regions. This has been predicted to add more than 290 person-rem to a typical refueling outage, in total, even with
, 36 l mitigating measures, HWC was predicted to more than double the total yearly collective radiation dose at Susquehanna. !
Mitigatina Measures 1
Licensee staff had taken a number of initiatives to mitigate the radiological impact of hydrogen water chemistry. Examples included installation of condensate filtration to reduce feedwater iron concentration; preparation and plans for chemical de- ;
contamination if excessive dose rates were encountered; preparation and plans for depleted zinc injection; improvements and increased use of shielding including identification and evaluation of multiple permanent shielding locations, installation of l
' shield supports in the drywell, replacement of some drywell grating with steel plates, and construction of shield walls at the feedwater heater bay; installation of :
cameras in steam related areas including reactor water feed pumps, condenser bay, l moisture separators, turbine deck, and control valves; revised work practices and '
efficiency improvements to minimize time in areas with elevated dose rates; revised work scheduling such es movement of some 10 year in-service inspection j requirements from the unit 2 ninth refueling outage to the unit 2 eighth refueling '
outage and scheduling of some steam related work during outages; and initiation of multiple staff communications to inform station personnel of the benefits and impact of hydrogen water chemistry.
- c. Conclusions The inspector concluded that hydrogen water chemistry (HWC) was predicted to have a significant radiological impact, essentially doubling the total yearly collective dose at SSES. The licensee has made reasonable efforts to assess the radiological impact of hydrogen water chemistry. Although as-low-as-reasonably-achievable (ALARA) initiatives to minimize the radiological impact of HWC appeared comprehensive, including the implementation of condensate filtration, shielding up-grades, contingencies for chemical de-contamination, and improvements in work practices and scheduling, continued vigilance to assess and mitigate the radiological impact of HWC is warranted.
R2 Status of RP&C Facilities and Equipment
- a. Inspection Scone 183750)
The inspector performed selected tours of the Unit 1 and 2 reactor and turbine buildings to evaluate the condition of facilities, housekeeping, and hedth physics equipment. !
l l
l
-__ _ . _ . _._ _ . . .__ _ ._ - . _ _ _ _ _ ._. _ _ _.._._m- . _.
1
.e ; i I
37 !
- b. Observations and Findinas I
Overall, material conditions of plant structures and operating equipment were very good. No significant water leakage was observed from pumps or valves, catch
.' containments were installed as needed, and there was no observable ground water intrusion in the lower elevations of the reactor buildings. Many locations in the
[ reactor buildings had worn or chipped paint on floor surfaces; however, fresh paint i ' had been applied to floors, walls, structures, and piping in the lower elevation of the 1
unit 1 and unit 2 reactor buildings that housed emergency core cooling systems (ECCS).
/
Overall, plant housekeeping was good as evidenced by clear isles and walkways, and the volume of stored materials had been noticeably reduced in the radiologically controlled areas as evidenced by the removal of cages used to temporarily store tools and equipment.
Contamination monitoring equipment observed to be in-use including friskers, personnel contamination monitors (PCMs), tool contamination monitors (TCMs), and continuous air monitors (CAMS) appeared to be well maintained and in good working condition.
- c. Conclusions Based on this review, the inspector concluded that health physics equipment and facilities were well maintained, and housekeeping and material conditions of plant structures and equipment were good.
R6 RP&C Organization and Administration
- a. Insnaction Scope (83750)
The inspec'or reviewed the organization and administration of the health physics organization. Several changes had been made including the appointment of a new Supervisor-Health Physics. A review was performed to determine if the individual's qualifications met the requirements specified in technical specifications. Information I was gathered by a review of an organization chart, interviews with cognizant l personnel, and review of a resume.
l
- b. Observations and Findinas Technical Specification 6.3, " Unit Staff Qualifications," required the Supervisor- i Health Physics to meet or exceed the qualification requirements of Regulatory Guide 1.8, " Personnel Selection and Training," dated September 1975. The inspector i compared the qualifications of the Supervisor-Health Physics that were documented on a resume, to the qualification requirements presented in Regulatory Guide 1.8.
This review indicated that the individual met the qualification requirements for the l position of Supervisor-Health Physics (Radiation Protection Manager). l
- c. Conclusion !
l I
' , .; %s l
l 38 Based on this review, the inspector concluded that the newly appointed Supervisor-Health Physics met the qualification requirements outlined in plant technical specifications 6.3.. )
i l
t R7 Quality Assurance in RP&C Activities
- a. Insoection Scone (83750)
]
A review was performed of the effectiveness of the condition reporting system for resolving radiological control deficiencies. Information was gathered by a selected l
review of condition reports, interviews with cognizant personnel, and tours through the facility.
- b. Obmarvations and Findinos A selected review of records showed that a variety of low threshold radiological control issues were placed into the condition reporting system as evidenced by the following examples: individual performed minor personal de-contamination without l health physics assistance; uncontaminated tools with purple paint were found ir. the L Combo shop; and informational postings of radioactive material were found to be j inconsistent. The inspector noted that when items were entered into the condition !
l reporting system, immediate corrective actions were taken, a significance review !
was performed, causes/ causal factors were identified, past experience was l reviewed, and actions to prevent the condition were investigated or implemented.
L Interviews revealed that health physics technicians had increased confidence in and readily used the condition reporting system to resolve radiological control program deficiencies.
l . Condition report number CR 97-2624 was selected to evaluate the effectiveness of the condition reporting system in resolving deficiencies. CR 97-2624 was originelly l written after an instrument and control (l&C) technician notified the health physics
! staff that he received radiation exposure at a greater rate than anticipated while i
working riear the unit 1 fuel pool cooling unit precoat tank. The I&C technician had l been assigned to perform work in an area with dose rates of 10 mrem /h; after 10 l minutes, the individual received 7 mrem of dose. Followup surveys showed that work area dose rates had increased from 10 mrem /h to approximately 40 mrem /h i and that the elevated dose rates originated from the bottom of the unit i fuel pool 5 filter domineralizer precoat tank. The licensee concluded, and the inspector i concurred, that the individual did not exceed an allowable exposure because the l individual was equipped with a personal alarming dosimeter, and the I&C technician l maintained a questioning attitude. During the initial review, licensee staff noted l' that the condition reporting system contained related condition reports involving elevated dose rates related to the change-out of the FPC filters and processing of l_ the wastes. A multi-disciplined team includir.g representatives from operations,
! health physics, effluents, scheduling, and engineering was assembled to investigate l
the event and implement corrective actions to prevent recurrence. Deficiencies in equipment, system design, and procedures associated with the fuel pool cooling domineralizer and waste processing systems were identified. Examples of actions
( taken to prevent creation of un-posted or un-barricaded high radiation areas included l
j < e / u.. w I
39 1
revising operations procedure OP-135-001/OP-235-001 to require health physics (HP) notification'and specific " hold-points" when a backwash of the FPC filter domineralizer was necessary, and revisions to health physics procedure HP-Hi-073,
" Notification of Plant Evolutions and Expected HP Actions," Rev.11, to ensure proper posting, barricading, and access control at the fuel pool hold pump room, the fuel pool precoat tank, and fuel pool cooling back wash receiving tank upon change out of the fuel pool cooling unit filter domineralizer. Other actions to prevent recurrence were to change the precoat tank alarm response to isolate the filter from the precost tank, and to evaluate methods for waste stream flushing. Corrective actions were broad based and appeared adequate to prevent recurrence.
- c. Conclusions The inspector concluded the condition reporting system was effectively used to identify, evaluate, and resolve radiological control program deficiencies.
R8 Miscellaneous RP&C lasues R8.1 Final Safety Analvsis Renort Review (83750)
A recent discovery of a licensee operating their facility in a manner contrary to the
- Final Safety Analysis Report (FSAR) description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the FSAR description. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the FSAR that related to the areas inspected.
The inspector reviewed selected sections of Chapters 12.1 - 12.5, " Radiation l Protection," of the FSAR, pertaining to radiological controls, to evaluate the l accuracy of the UFSAR regarding existing plant conditions and practices.
No FSAR discrepancies were identified during this review.
R8.2 Hydrogen Water Addition Modification (71750)
During a review of TS and the SSES on going Hydrogen Water Addition (HWA) system modification work, the inspectors determined the licensee was directing a !
substantial effort to reduce the impact of sulfate concentrations on the integrity of the reactor coolant system (RCS). RCS oxygen levels are reduced by the HWA system in order to prevent intergranular stress corrosion cracking initiation related to high suifate concentrations.
i i
. .- .. . , j
, 40 V. Management Meetings i
X1 Exit Meeting Summary
( The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection report period on March 16,1998. The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. The licensee identified a proprietary Electric Power
! Research Institute (EPRI) document, to which the inspectors had access. The licensee did not disagree with any of the findings presented at either exit meetings.
! The licensee stated at the exit meeting that their evaluation of the as-found standby liquid control accumulator pressure, on November 25,1997, would be completed prior to June 1 30,1998, i
I e
f
e g ,e -
1 41 ITEMS OPENED, CLOSED, AND DISCUSSED Onened 50-387,388/98-01-01 IFl Licensed Operator Re-qualification Training Program (section 05.1) 50-387,388/98-01-06 eel Unit 1 Standby Liquid Control Accumulators Found Depressurized (section M8.1) 50-387,388/98-01-07 eel Unit 1 Standby Liquid Control Accumulators Found Depressurized (section M8.1) 50-387,388/98-01-08 eel Unit 1 Standby Liquid Control Accumulators Found Depressurized (section M8.1) 50-387,388/98-01-09 URI Primary Containment Penetration Leak Rate Testing -
Notice of Enforcement Discretion (section E1.2)
Uodated 50-388/97-10-01 URI Technical Specification 3.0.3 Entry to Support Surveillance Activities (section 08.1) 50-387,388/96-270-01022 VIO Failure to implement Procedures as Required by TS 6.8.1 (section 08.3)
Closed 50-388/98-01-02 NCV Required Sample was not Collected and Analyzed within Technical Specification Time Limit (section 08.2) 50-388/98-01-03 NCV Reactor Building Vent Continuous Sampf s Lost for Twenty Minutes (section 08.2) 50-387,388/98-01-04 NCV Control Room Emergency Outside Air Supply System (CREOASS) Operability (section 08.3) 50-387,388/98-01-05 NCV Hot Tapping of Safety Related Piping (section M3.1) 50-387,388/98-01-10 NCV Reactor Building Truck Bay Hatch (section E8.1) 50-388/97-03-00 LER Required Sample was not Collected and Analyzed within Technical Specification Time Limit (section 08.2)
-. g ,,. . .
42 I
) '. 50-388/97-05-00 LER Reactor Building Vent Continuous Sample Lost for l Twenty Minutes (section 08.2) l l
50-387/97-013-00 LER Requirements for Testing Activated Carbon Samples (section 08.2) 50-387/97-21-00 LER Condition Prohibited by Technical Specifications -
Technical Specification 3.0.3 Entry (section 08.2) l I
50-387/97-21-01 LER Condition Prohibited by Technical Specifications - l Technical Specification 3.0.3 Entry (section 08.2) l l
50-387,388/96-270-01012 VIO Less Than Four independent Diesel Generators
- Operable (section 08.3) 50-307,388/96-270-02013 VIO Containment isolation Valve Open and Deactivated for 24 Hours (section 08.3) 50-387,388/96-270-03014 VIO Standby Liquid Control System Heat Trace De-energized (section 08.3) 50-387,338/97-04-02 VIO Nuclear Safety Assessment Group (section 08.3) 50-387,388/97-04-03 VIO Quality Assurance Program (QA) Changes (section 08.3) 50-387,388/97-06-04 URI Control Room Emergency Outside Air Supply System (CREOASS) Operability (section 08.3) i 50-387/97-10-04 URI Standby Liquid Control Accumulators Found Depressurized (section M8.1) l- 50-387/94-14-01 URI Evaluation of Thermal and Pressure Locking (section M8.2) 50-388/94-15-01 URI Evaluation of Thermal and Pressure Locking (section M8.2) l l
50-388/97-09-02 VIO Reactor Recirculation Valve Bonnet Vent Line Failure (section M8.2) i 50-387,388/97-07-09 URI Reactor Building Truck Bay Hatch (section E8.1) !
l l
50-387,388/98-01-11 NCV ASO Performance for Radwaste Control Room Alarm Tests (Section E2.3) i l
50-387, 388/98-01-12 NCV PCO Performance of Control Room Annunciator Alarm Tests (Section E2.4) l
l~ .).n <'
I' 43 LIST OF ACRONYMS USED i
ALARA As Low As Reasonably Achievable '
ATWS Anticipate Transient Without Scram ASO Auxiliary Systems Operator CAM Continuous Air Monitor CFR Code of Federal Regulations l- CR Condition Report l CREOASS Control Room Emergency Outside Air Supply System j CRD Control Rod Drive
! ECCS Emergency Core Cooling System EDG Emergency Diesel Generator eel Escalated Enforcement item l EPRI Electric Power Research Institute ESS Engineered Safeguard System ESW Emergency Service Water FPC . Fuel Pool Cooling FME Foreign Material Exclusion FSAR Final Safety Analysis Report GDC General Design Criteria GE General Electric !
GL [NRC] Generic Letter i l gpm gallons per minute HP Health Physics HPCI- High Pressure Coolant injection i HWA Hydrogen Water Addition {
HWC Hydrogen Water Chemistry l I&C Instrument and Contrcl j IFl Inspection Follow-Up item IR [NRC] Inspection Report ISEG Independent Safety Engineoring Group ITS Improved Technical Specification JPM Job Performance Measure kv Kilovolts Kw Kilowatts I LPRM Local Power Range Monitor l LCO Limiting Condition for Operation LORT - Licensed Operator Re-qualification Training LER Licensee Event Report M-G Motor-Generator MOV Motor Operated Valve mrem millirem .
l' mrem /h mram per hour
[ N-16 Nitrogen-16 i NAS Nuclear Assessment Services NCV Non-Cited Violation NDE Non-destructive Examination NOED- - Notice of Enforcement Discretion NOV Notice of Violation
- .I . i y -
44 NPO Nuclear Plant Operator NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSAG Nuclear Safety Assessment Group NSE Nuclear System Engineering i OD Operability Determination PAD Personal Alarming Dosimeter PCO. Plant Control Operator PCM Personnel Contamination Monitor PORC Plant Operations Review Committee !
psig Pounds per Square Inch Gauge QA - Quality Assurance RBM Rod Block Monitor RCA Radiologically Controlled Area RCS Reactor Coolant System RG Regulatory Guide RHR Residual Heat Removal RMCS . Reactor Manual Control System RP&C Radiological Protection and Chemistry RPM Radiation Protection Manager RWCU Reactor Water Cleanup RWP Radiation Work Permit SCC Stress Corrosion Cracking SDC Shut Down Cooling SER Safety Evaluation Report SIL IGE] Service information Letter SLCS Standby Liquid Control System SPING System Particulate lodine Noble Gas j SRC Susquehanna Review Committee i SRO Senior Reactor Operator l SRV Safety Relief Valve )
SS Shift Supervisor SSES Susquehanna Steam Electric Station STA Shift Technical Advisor i TCM Tool Contamination Monitor l TLD Thermolurninescent Dosimeter
]
TS Technical Specification TSI Technical Specification Interpretation URI [NRC) Unresolved item US . Unit Supervisor i WA Work Authorization i
i I
)
,