IR 05000280/1993015
| ML18153B296 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/06/1993 |
| From: | Merschoff E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Lainas G Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9308270118 | |
| Download: ML18153B296 (23) | |
Text
{{#Wiki_filter:AUG - 6 1993 MEMORANDUM FOR: Gus C. Lainas, Assistant Director Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation FROM: Ellis W. Merschoff, Director Division of Reactor Projects, RII SUBJECT: TASK INTERFACE AGREEMENT (TIA) - SURRY FACILITY MAIN FEEDWATER SYSTEM ISOLATION (TIA 93-016) The purpose of this request is to determine the acceptability of Surry's main feedwater isolation system, including allowed outage time, and reliance on operator actio An inspection was conducted at the Surry facility from June 6, 1993, to July 3, 199 The inspection results are documented in NRC Inspection Report Nos. 50-280/93-15 and 50-281/93-15 (Enclosure 1). An unresolved item was identified by the inspectors because methods being used by the licensee for evaluating and resolving a self identified potential unreviewed safety question involved using Probability Risk Assessment and guidance provided in NSAC-125, Guidelines for 10 CFR 50.59 Safety Evaluation Discrepancy Report (DR) S-93-0774 described a possible unreviewed safety question associated with isolating the feedwater syste The feedwater isolation function is described in section 14.3.2 of the UFSA A detailed description is not provided but the system contains a safety related protective system that is actuated by a safety injection (SI) signal. The SI signal causes the automatic closure of the air operated feedwater regulating valves (FWRVs) and the FWRV bypass valves. The SI signal also trips the main feedwater pumps {MFWPs).
With any operating unit FWRV disabled in the full open position {on its jack) to facilitate troubleshooting/repairs, there is an electrical system failure scenario that would require considerable manual operator action \\o effect feedwa~er isolatio In August 1992 to support placing a FWRV on its jack, the licensee performed a 10 CFR 50.59 safety evaluation {SE).
The SE determined that jacking the MFRV open to perform maintenance was an acceptable practice providing that MFWP redundant tripping was operable and compensatory actions were in place to close another non-safety feedwater isolation motor operator valve (1 or 2-FW-MOV-154 or -254A, B or C) just up stream of the FWRV from the control room if necessar The SE indicated that the feedwater isolation would be achieved within 30 second The above SE was subsequently reviewed by the licensee's independent safety review grou They questioned the 30 second time response as well as postulating some plant configurations where the time required to perform compensatory actions would be greater than 30 second To support future cases of operation with a disabled FWRV, the licensee developed SE 93-155 (Enclosure 2) and a Technical Specification {TS) interpretation. The new SE ,*: C* (:, r. n -. 9308270118 930806 ~; PDR ADOCK 05000280 t*i:,
PDR t\\ \\\\ IfO[
Gus C. Lainas
AUG - 6 1993 justified plant operations to continue with an administrative 72-hour allowed outage time {AOT) for an inoperable FWRV {i.e., jacked open).
The new SE was based on reliability and probability judgement that the licensee considered acceptable by NSAC-125, Guidelines for 10 CFR 50.59 Safety Evaluation The licensee also judged that manual operator action to take the FWRV off the jack and allow it to close or to close the inline MOV within a 220 second time limit would be achievable and acceptabl At the time the inspection report was generated, the licensee had developed iterative SE 93-142 to justify an eight-hour AOT with operator action within 30 seconds. This was superseded by SE 93-155 which justifies an AOT of 72 hours with 220 seconds for operator manual actio The inspector's review of SE 93-155 and supporting calculation SM-899 concluded that the licensee had evaluated the effects of the 220 second feedwater isolation time on the main steam line break {MSLB) analysis as it related to reactor vessel integrity criteria for pressurized thermal shock {PTS) only. Questionable assumptions in the calculation include; 1) dismissing the need to relock at the effects on departure from nucleate boiling {DNBR) calculations and, 2) excluding the effects of a stuck control rod from the calculation based on probability. Discussions with the licensee also indicated that containment pressurization from a MSLB was considered to be outside the licensing basis for Surry and has not been analyze The use of non-safety related equipment, which is not covered in the TS, to perform the isolation function further complicates this issu Region II requests NRR's assistance in the resolving this-issue. Of particular concern is whether this issue or similar issues can be evaluated on reliability and probability considerations based on NSAC-125 which is currently being generically utilized by the industry in evaluating safety issues. Additionally, is PTS the only event that would be sensitive to delays in isolating feedwater or should the effects on DNBR and containment pressurization be evaluated assuming worst case stuck rod in the reanalysis? Should safety-related equipment be used for this isolation function and should the equipment be covered by TS, or is the current equipment and licensee's TS interpretation acceptable? These issues have previously been discussed with members of your staf If you have any additional questions, please contact M. W. Branch on {804) 357-2101 or G. A. Belisle on {404) 331-419 }frigfoal Sir.mid by Ellis W. Merschoff Ellis W. Merschoff
Enclosures:
Excerpt from IR 50-280, 281/93-15 Licensee's Safety Evaluation 93-155
REGION 11 101 MARIETTA STREET. N.W.. SUITE 2900 ATLANTA. GEORGIA 30323-0199 JUL 3 0 m3 Docket Nos. 50-280. 50-281 License Nos. DRP-32, DRP-37 Virginia Electric and Power Company ATTN: Mr. W. L. Stewart Senior Vice President - Nuclear 5000 Dominion Boulevard Glen Allen, VA 23060 Gentlemen: SUBJECT: NRC INSPECTION REPORT NOS. 50-280/93-15 AND 50-281/93-15 This refers ta the Nuclear Regulatory Convnission (NRC) inspection conducted by Mr. M. Branch of this office on June 6 through July 3, 199 The inspection included a review of activities authorized for your Surry facilit At the conclusion of the inspection, the findings were discussed with those members of your staff identified in the enclosed repor Areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progres The enclosed Inspection Report identified activities that violated NRC requirements that will not be subject to enforcement action because the licensee's efforts in identifying and/or correcting the violation meet the criteria specified in Section VII.B of the Enforcement Polic Pursuant to 10 CFR 2.790 of the NRC's "Rules of Practicea, a copy of this letter and its enclosure will be placed in the NRC Public Document Roo Should you have any questions concerning this letter, please contact u
Enclosure:
NRC Inspection Report
REGION II== 101 MARIETTA STREET, N.W., SUITE 2900 ATLANTA, GEORGIA 30323-0199 Report Nos.: 50-280/93-15 and 50-281/93-15 Licensee: Virginia Electric and Power Company 5000 Dominion Boulevard Glen Allen, VA 23060 Docket Nos.: 50-280 and 50-281 License Nos.: DPR-32 and DPR-37 Facility Name: Surry 1 and 2 Inspection Conducted: June 6 through July 3, 1993 Inspectors: Approved by: M. ~ior Resident Inspect r a.~~ S. G. Ting~~ edent Inspector G.~~hief Division of Reactor Projects SUMMARY Scope: 7/?-~3 Da'te gned D This routine resident inspection was conducted on site in the areas of plant status, operational safety verification, maintenance inspections, safety assessment and quality verification, Technical Specification review program, Updated Final Safety Analysis Report improvement program, Level I project tracking, and licensee event revie During the performance of this inspection, the resident inspectors conducted reviews of the licensee's backshifts, holiday or weekend operations on June 20 and 2 Results: In the operations area, the following items were noted: URI 50-280,281/93-15-0l, Use of probalistic risk assessment (PRA) for Unreviewed Safety Question Determination, pending further review by the NRC (paragraph 3.c).
5 Feedwater Isolation System Review During a station deviation review, the inspectors became aware of DR S-93-0774 which described a possible unreviewed safety question associated with isolating the feedwater syste Feedwater isolation is needed to protect against the consequences of a steam line failure that could cause a pressurization of the containment or a RCS cooldown with a loss of reactor shutdown margi The feedwater isolation function is described in the UFSAR, section 14.3.2. A detailed description is not provided but the system consists of a safety related protective system that is actuated by an SI signal which causes automatic closure of all three pairs of air operated FWRVs and bypass FWRVs and the tripping of the MFWP Additionally, the tripping of the MFWPs through non-safety related breakers causes the non-safety related discharge valves (1 or 2~FW-MOV-150 or 250A&B) which are powered from non-safety electrical busses to clos In the past, Surry has disabled (jacked) a FWRV in the open position to a specific SG; therefore, it would not automatically close on an SI signal. The licensee performed SEs (i.e., 10 CFR 50.59) reviews for this evolution since 198 Recent SEs determined that jacking the valve open to perform maintenance on the control or feedback circuitry was an acceptable practice as long as the redundant tripping of the MFWPs was operable and compensatory actions were in place to close another non-safety feedwater isolation MOV (1 or 2-FW-MOV-154 or 254A, B, C) just up stream of the FWRV, from the control room if necessar DR S-93-0774 was issued because of the licensee's independent review group's assessment of the SE 92-173, dated September 24, 1992, which addressed placing a FWRV on its jack. The package reviewed included administrative control form (AC S2-92-807) and TS interpretation (TSI-014), including safety evaluation (SE 92-173B).
The independent review group's report to the station, attached to the DR, indicated that for certain postulated - cases, termination of feedwater flow during a steam line break accident may not be achievable with current compensatory actions and within the time frame assumed in the steam break accident analysi The following information was taken from that report: The safety evaluation (SE 92-173B) assumes that the feedwater isolation will occur within 30 seconds through closure of the feedwater isolation MOV For a total LOOSP 1 the MFWP and condensate pumps will shutdown due to a loss of electrical power thereby terminating feedwater within the required tim However, the SE did not address the impact of a partial loss of station service electric powe A potential scenario is that a loss of a 4160V station service bus with one FWRV jacked open could result in continuous feedwater flow to the SG that has experienced a MSL The combination of a feedwater isolation MOV and MFWP
e
discharge MOV loosing power, MFWPs tripped on the SI signal, and the condensate pump continuing to run (from the unaffected station service busses}, results in continued feedwater flo This scenario assumes a MSLB upstream of the MSTVs and the single failure being a loss of a 4160V station service bu Feedwater flow would continue until the operators locally closes the feedwater isolation MOV or trips the condensate pum The time required to perform these compensatory actions would be greater than the 30 seconds assumed in the safety analysi After being informed of the above potential unreviewed safety question item, the plant initiated actions to evaluate the condition and determine reportabilit The station's review resulted in generating another SE (93-142) and a TS interpretation that allows plant operations to continue with an administrated 8-hour AOT for an inoperable FWRV (i.e., jacked open).
The new SE was based on reliability and probability considerations which the licensee considered acceptable by NSAC-125, Guidelines for 10 CFR 50.59 Safety Evaluation The licensee also considered that manual operator action to trip the FWRV off the jack and allow it to close within the 30 second time limit would be achievable and acceptabl The inspectors questioned the use of probability assessment in making the decision that the modification (i.e., loss of automatic isolation of main feedwater) does not constitute an unreviewed safety question. It is clear that the inability to automatically terminate feedwater flow was not considered in the safety analysis described in the UFSA Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increase The licensee's calculation (S-896) reviewed by the inspectors indicated that the combined increase in probability for the event described was insignificant if an AOT of 8 hours is use The licensee, therefore, considered it acceptable to operate with the FWRV on the jack if an AOT did not result in an increase in probability of occurrenc The licensee's recent SE (93-142) also indicated that the operators could close the jacked open FWRV within the 30 seconds needed to terminate feedwater flow for the steam break acciden However, the SE contained statements that the basis was limited risk associated with operating on the jack and not dependent on strict operator performance to close the FWRV within the 30 seconds assume If it could be demonstrated that manual action was addressed by a procedure and could be reliably performed within the values bounded by the accident analysis, it appears that the guidance of GL 91-18 could apply for future occasion The inspectors held discussions with NRC staff and were informed that the use of NSAC-125 has not been endorsed by the NR Therefore, the use of a probability assessment is in questio *
This item is identified as URI-50-280, 281/93-15-01, Use of PRA for Unreviewed Safety Question Determination, pending further review by the NR Housekeeping In an effort to upgrade station housekeeping, the Assistant Station Manager is walking down areas of the station with the supervisors responsible for housekeeping in the are On June 9, the inspectors accompanied the Assistant Station Manager and I&C Supervisor on a housekeeping tour of the auxiliary building 45 foot level. During this tour, general material condition and cleanliness were monitored and out of the way places such as under and behind equipment were inspecte Examples of items noted during the tour were red tape on cable tray covers, coat hanger wire dangling from conduit, damaged label plates, loose junction box covers, and loose fasteners on ventilation equipmen The inspectors noted that during the inspection period, the Vice President of Nuclear Operations conducted a similar walkdown of the auxiliary building basemen The inspectors concluded that housekeeping has improved and management walkdowns have contributed to the improvemen Within the area inspected, one URI was identifie. Maintenance Inspections {62703) {42700) During the reporting period, the inspectors reviewed the following maintenance activities to assure compliance with the appropriate procedure Leak Repair at Mechanical Joint On June 14, the inspectors witnessed the licensee repairing a leaking mechanical joint in the piping/tubing to charging pump 2-CH-P-lB suction pressure gage 2-CH-39 The mechanical joint was disassembled, inspected and reassembled. This maintenance was considered minor maintenance and accomplished in accordance with deficiency card IC-93-017 The maintenance was isolated utilizing operator standby and the joint was leak tested when placed back into service. The inspectors reviewed Attachment 13 of VPAP-2002, Work Request and Work Order Tasks, dated January 1, 199 The attachment describes criteria for determining which tasks are considered minor maintenance and can be performed with deficiency cards. Maintenance accomplished per deficiency cards is generally simple in nature and does not require a WO or written instruction Item {d) of Attachment 13 states that minor maintenance shall not effect the integrity of safety related components and disassembly of a safety-related component is not a minor maintenance activity. The inspectors concluded that disassembling the mechanical joint in the
ENCLOSWRE 2 I Memoran m NSA 93146
VIRGINIA POWER NORTHCAROLJNAPOWER To Mr. W. R. Benthall - SPS From Mr. K. L. Basehore - IN/3SW Innsbrook Technical Center July 13, 1993 SAFETY EVALUATION TSI FOR ADMINISTRATIVE CONTROL OF FEEDWATER REGULATING VALVES SURRY POWER STATION UNITS 1 AND 2 The evaluation of the 72 hour allowed outage time (AOT) for the main feedwater regulating valves has been complete The 10 CFR 50.59 Safety Evaluation, which supports this extension to 72 hours, is attache The evaluation of the steam line break impact and the human reliability for manual actions to close the main feedwater regulating valves and the isolation MOV has been documented in Calculation SM-89 If you have questions, please call Nuclear Safety Analysi i~ D. S,i-K. L. Basehore Attachments: 1) Safety Evaluation 2) supporting Calculation SM-899 cc: Mr. R. M. Berryman - IN/3SW Mr. R. L. Blount - SPS Mr. J. o. Erb - IN/3SW Mr. W. M. Oppenhimer - IN/3SW NAF File 2.5.2 - IN/3SW NSA File 21.1 - IN/3SW ~ P~e,vr:o f Jtn.". f 5 1993
- >tJS Sena, #---
SURRY LICENSJNG-- Fonn No. 720003(Feb 901 (Fonneny 970240201
93025 * VIRGINIA POWER Safety Evaluation Number q3 _( s-s-e Safety Evaluation Page I of 12 VPAP-3001 Rev. 2 Applicable Station [ J North Anna Power Station [XJ Surry Power Station List the governing doc1.J11ents for which this safety evaluation was performe. Applicable Unit [ J Unit [X] Unit [ ] Unit 2 [Xl Unit 2 GOV Engineering Calculation SM-899, Assessment of Adninistrative Allowed outage Time for Main Feeed Regulating Valves and I so lat ion Valves, 7 /9/9 <<. x-1 "4C:: Surrmarize the change, test, or experiment evaluate Placement of any Unit 1/2 FRV on its full open jacK as require This condition will prevent the valve from performing intended safety functio The proposed administrative controls will locate an operator in the MCR to close the FW line isolation HOV, and an operator in the affected unit's MER to remove the valve from the jacK to allow the valve to close demand if require In addition, the allowed time to place a FRV on the jacK is limited to 72 hour. State the purpose for this change, test, or experimen The purpose of the adninistrative controls is to ensure that the safety function of the Unit 1/2 MFRV is preserve Placing the MFR\\/ on its jacK is used on limited occasions to stabilize a Main Feedwater Regulating Valve (MFRV) for maintenance, typically the replacement of failed control cards or control system corrponents, or troubleshootin. List the limiting conditions and special requirements identified or assuned by this safety analysi These are identified in Calculation SM*899 and in the respose to Question 1. Will the proposed activity/condition result in or constitute an unreviewed safety question or require a licensing amenanent? Preparer Name (Print) 1 Date N. A. Smith, J ~/fj 1 Cognizant Supervisor Name (Print) K. L. Basehore [ l Yes 15. Disposition [~pproved [ l Disapproved C l Approved As Modi f i ed [ l Requires Further Evaluation 1 SNSOC Chairman Signature 1 Date 1 -{(o _q3 Comnents =- [Xl No I I J
.Safety Eval Page 2 of 12 VIRGINIA POWER VPAP-3001 Rev. 2 GOV 18. SLillll8rize from Part D, Unreviewed Safety Question Determination, the major issues considered; state the reason the change, test, or experiment should be allowed; and state why an 1..nr.eviewed safety question does or does not exist (a si~le conclusion statement is insufficient).
MAJOR ISSUES: IJhen a MFRV malfunction occurs, the resultant level oscillations can easily result in a severe secondary transient anct eYe-a unit tri In order to perform certain corrective maintenance (e.g., control circuit troubleshooting), the valve ii stabilized by being placed on its jacking mechanis However, this action defeats the autanatic closure of the valve on a SI or a SG hi-hi level (P*14) signal. The use of the jack nust c~ly with the safety analysis requirement for FW isolation ard the potential for secondary overcooling during a Hain Steamline Break (HSLB) acciden PROPOSED SOLUTION: The 1,/QG STS (Reference 2) i""'°se a 72 hour AOT for one or more HFRV's or HFRV bypass valves out of service, a 72 hour AOT for one or more isolation HOV's out of service and an 8 hour AOT for more than one of the above valves out of service in the same feedwater flow path. The STS is based on a plant designed with redundant closure of fast-acting MOV's and MFRV's on deman Since Surry does not have HOV autoclosure on an isolation signal, placing a HFRV on its jack at Surry is an acceptable alternative to the condition addressed with an 8 hour AOT in the Standard Specifications. Use of a longer AOT (e.g., 72 hours) should therefore be acc0ff1)8nied by other aaninistrative controls for redundant manual feedwater isolatior An aaninistrative AOT is needed because one is not defined in the Surry Technical Specifications ard it is periodically necessary to place a HFRV on its jac The proposed controls for performing work on a HFRV at power include the following: 1) No more than one HFRV will be on its jack at a tim ) The AOT will be no more than 72 hour ) During the AOT, the associated isolation HOV will be stroked partially closed to ensure it can be 111&,..,.lly closed if require ) During the AOT, an operator in the control room will be dedicated to stroking the isolation HOV closed in the event of safety injection signal or a SG hi*hi level signa ) During the AOT, an operator will be dedicated at the HFRV to close the valve as require JUSTIFICATION: Engineering calculation SH*899 showed that, on a risk basis, these proposed controls form an appropriate basis for performing maintenance of these valves at Surr This was done by showing that: a: the overall reliability of the valves is not significantly irrpacted and therefore the probability of a malfl.SlCtion of equipment i~rtant to safety is not increased; ard b: the probability of a main steam line break inside contairment coincident with excess feedwater flow to the faulted generator (i.e., failure of automatic or manual FW isolation) is negligibly small and therefore the consequences of a design basis HSLB are not increased for the proposed AOT 1s. Specifically, it was concluded that: (CONTINUED)
- e Safety Evalu!on Supplemental Page 2A of 12 VIRGINIA POWER VPAP-3001 Rev. 2 1 (Continued)
~) A transient assessment of MSLB with failure to isolate MFW prior to 220 seconds shows that excessive cooldown (as measured by the reactor vessel integrity criteria) does not resul GOV 2) The estimated frequency of a MSLB with failure to isolate main feedwater prior to 220 seconds is< the cutoff fr~ 1.ox10** yr*' for sequences which emergency operating procedures rrust cover (see Ref. 3).
3) The expected reliability of the proposed operator actions in isolating MFW when the MFRV is on its jack is c~rable the expected reliability of the MFRV during normal operations, given that at least 220 seconds are available to achiev manual isolatio ) The !PE concluded that MSLB in containment was not a significant contributor to core damag However, even if the extremely conservative assurption is made that a MSLB with a MFRV on its jack leads to core damage with a probability unity, the resulting freqency is less than 0.01!1: of the !PE point estimate COF of 7.4 x 10** yr*' (Ref. 4 Executive s llTlllll ry) * The MFRV is placed on the jack only because the system is unstable and maintenance is require The alternative to performing maintenance with the valve in OVERRIDE is to leave the MFRV in its degraded condition. It is plainly preferab to inmobilize the erratic MFRV for the short repair period than to continue to operate with deteriorated SG level control Only one valve at a time will be on its jac Finally, the following plant and procedural features provide r~nt assurance of FW isolatio ) The main feedwater l)U1')S trip on an SI or High/High SG leve ) The main feedwater ~ discharge MOV's, 1(2)-FW-MOV-150(250)A/B, auto close on their respective feedwater ~ trfp: 3) Dedicated operators are stationed at both the control board and the MFRV to ensure that the valve will be closed when feedwater isolation is require ) lnmediate actions in 1(2)-E-O require verification of feedwater isolation following an SI, with explicit instructions verify closure of the FRV's, the FW bypass valves, and removal from service of all main FW JlU1l>S Cpurp trip and discharge MOV closure).
5) If a reactor trip occurred without an SI, 1(2)*ES-0.1 directs the operator to verify that the MFRV's close at 554 degrees F and to also close SG FW isollation MOV's 154A/B/ ) Procedure 1(2)-FR-H.3 (Response to Steam Generator High Level) can be entered if SG levels exceed 75!1: N The proce verifies FW isolation (with explicit instructions to verify that main feed l)U1')S are stopped; the MFRV's are closed: t FW bypass valves are closed; and the feedwater ~ discharge MOV's are closed).
(CONTINUED)
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~afety Evalu _ Supplemental Page 2B of 12 VIRGINIA POWER VPAP-3001 Rev. 2 GOVO ~-,.,*::.* *.-.**<*;:,*~~.. *~**_',*.. -\\*,~.:... *..... ~*~:*;~~-:.*_,* .
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.. 1 (Continued) UNREVIEWED SAFETY QUESTION ASSESSMENT: T~rery placement of a MFRV on its jack does not pose en unreviewed safety question for the following reasons: 1) Accident probability is not increased. Placing a MFRV on its jack is sometimes a necessary part of stabilizing and restoring en erratic SG water level control system from which feedwater malfunctions can originat No other Chapter 14 accident precursors have been affecte ) Accident consequences are not increase The redundant accident mitigation systems remain fully operable; the i~t upon feedwater isolation has been shown to be negligible, and all other mitigation systems remain unaffecte ) Unique accident probability is not created. Feedwater system malfunctions remain the only credible events that could occur as a consequence of placing a valve in OVERRID ) Equipment malfunction probability has not increase The proposed limitations on the duration of placement of the MFRV on its jack do not significantly i~ct the total failure probability of a MFR Further, the maintenance which requires the use of the jack restores the feedwater control system, including the MFRV itself, from a degraded state to fully OPERABLE conditio (Alternately, a power downramp will ccxrmence to isolate the MFRV.)
5) The consequences of equipment failure have not increase No other event mitigation equipment has been affected. The potential consequences of neglecting an erratic MFRV are RUCh more severe than the negligibly small consequences of briefly placing the valve on its jac Further, diverse isolation redundancy exists in various plant systems and EOP's to protect against secondary overcoolin ) Unique equipment failure probability has not been created. Spurious MFRV re-positioning or inmobilHy remain the only credible consequences of MFRV maintenance; these events have already been analyze ) Margin of Safety is maintaine The safety analysis requirement for FW isolation is met by diverse and redlrdant lll8,-JII and automatic isolation feature Although some scenarios have been postulated under which the analysis requirement to have diverse isolation features is not met (e.g., partial loss of station service resulting in no power to the isolatio HOV but the main feedwater ~ continue to ri.n) calculations have shown these to be statistically insignifican Therefore the current UFSAR conclusions regarding margin of safety remain vali * e - Safety Evaluation Page 3 of 12 VIRGINIA POWER VPAP-3001 Rev. 2 Pert. I - "R)l:icabte.,Refe~\\'::i.', :.. :\\:.. :*...... ***:-.*:. -:*
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.*:-*.*.*: 1 Identify applicable Updated Final Safety Analysis Report (UFSAR) section.2.7, Excessive Heat Removal Due to Feedwater System Malfunctions; 14.3.2, Rupture of a Main Steam Pipe; 10.3.5, Condensate end Feedwater System; 14.2.11, Loss of Normal *Feedwater-t Identify applicable Technical Specifications section.7, lnstrunentation Systems (Table 3.7*2) 2 Identify any other references used in this revie GOV 1) Engineering Calculation SH*899, Assessment of Acininistrative Allowed outage Times for Main Feedwater Reg. Valves and Isolation Valves, 7/9/9 ) WOG Standard Technical Specification, Rev. 0, 9/28/92, Section 3.7.3, MFTVs and MFRV ) Letter from R. W. Jurgensen, WOG, to s. H. Hanauer, USNRC, Westinghouse Owners Group Update of Item I.C.1 of NUREG*0737 Activiti.es, March 18, 198 ) Probabilistic Risk Assessment for the Individual Plant Examination, Final Report, Surry Power Station Units 1 and 2, August 199 ) 1(2)*E*O, Reactor Trip or Safety Injection: 1(2)*ES*0.1, Reactor Trip Response; 1(2)-FR*H.3, Response to Ste1111 Generator High Level * . Pert::c:-:*:-.... rteas*::COl)ll~~*:*av:,:n-til:,~~tv::~~~~l;~r,, **,,.:wt::@:i\\I\\i@?t:::j::J}ttNJMtfW:i.tJ::t::*:;::::,,:\\@:J:t\\tW{IttJ:=tf:tJ:f/t=W%ifoi.t\\i/@f@'fa@IINAWl@
- ~~:~!~~i=~it~0:;;~~~ll:i~--;-lf8iil.1lll~lilll1llllilllil:lll!l!ll:j 2 Will the operation of any system or component as described in the Safety Analysis Report be altered? This includes abandonnent of equir:xnent or extended periods of equipment out of service. Explai CX] Yes
[ l No The FSAR notes that the MFRV's auto-close on a SG hi-hi level, provide reo..ndant isolation during a MSLB and close at 554F following a reactor trip (P*4), all of which are defeated by placing the valve on its jac However, the duration in OVERRIDE is t~rary and in fact negligibly affects the total failure probability of the MFRV' Further, the use of OVERRIDE to stabilize the plant for SGWLCS repairs is greatly preferable to leaving the plant with an erratic SGWLC In addition, redundant hardware and procedural features provide additional assurance that a required isolation will oce&. Will the activity alter the performance characteristics of any safety related system or c~ent? Action statements, j~rs, and ten.,orary modifications should be reviewe Explai [X] Yes
[ ] No While the valve is on its jack, the MFRV will not auto-close on an SI, Hi-Hi SG level or a (Rx Trip/Tavg< 554F) signa However, the t~rary defeat is a negligible perturbation of total MFRV reliability. Further, OVERRIDE fs used only during maintenance which restores an already*degraded MFRV to OPERABLE conditio If the maintenance is not c~leted in 72 hours an assessment of valve operability will be performed and either the MFRV will be removed from OVERRIDE or the plant will be placed in a condition where the valve can be isolate. Will the ability of operators to control or monitor the plant be reduced in any way? CX] Yes [ J No Explai Auto-closure of the MFRV is t~rarily blocked while the valve is on its jack. However, this is done only to perfora maintenance which restores a deteriorated SG Level Control system to fully operational status. The brief loss of auto* closure capability fs greatly out-weighed by the restoration of secondary stability. Further, reo..ndant hardware and procedural features, including dedicated operators on the board and at the MFRV, provide additional assurance that a required isolation will occu * 2 Is a t~rary modification involved? [~itlasl1: 3.2.15] Are testing requirements as stated for the t~rary modification adequate to ensure operability after installation, as well as after removal of the t~rary modification? Explai [ l Yes [ ] Yes CX] N/A 00 No [ ] No The jack is an integral part of the MFRV designed to permit continued (stable) operation while maintenance is being performe e Sat ety Evaluation
Page 4 of 12 VIRGINIA POWER VPAP-3001 Rev. 2 GOV Part c -* ru.s,,Cclnsidered.*,ay:,:cJMs * .. Safety.Evetuetion==<Ccntiftled>.;,;:;:.;::,*
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.. *.*::\\i;(i':=.<:::* *.. :*.**. *.. <<:' \\i>*: *;,,* 2 could the proposed activity affect reactivity? Reactivity is affected by such CX] Yes [ ] No items as: RCS ten-perature, dilution, or flow; boric acid concentrations or .... volL.mes; RWST or accurulator boron concentration; main steam flow or instrunents that measure main steam flow; main steam pressure; nuclear instrunentation; - calorimetric power monitoring; rod control system; fuel, and fuel c~nent Explai The Reactor Engineer rrust approve the explanation for "Yes" answer [Ccaaitaents 3.2.9 & 3.2.14] A Safety Injection or Hi-Hi SG level signal will trip the main FW ~; redundant operator actions will close the MFRV's end isolation MOV' These actions will protect against overcooling the primar MFRV isolation at 554F will not occur for the valve which is blocked ope The valve will be manually closed upon e Rx trip~ Remote closure of the isolation MOV or local closure of the effected MFRV by dedicated operators is available to manage a reactor trip recovery with respect to stabilizing T-ave et no-loa The net result may bee slightly larger cooldown on a reactor trip. This reactivity effect will be minima A. Reactor Engineer Signe~ " 268. Date Pa) 7 (;'.J/13 /~I 2 IJi ll the activity significantly increase the potential for personnel injury or equii:ment [ ] Yes [Xl No damage? No system or c~nt will be subjected to any conditions which exceed its design parameter The MFRV 1s are provided with the jacking mechanism as a part of the valve design. As such, neither the MFRV and its connecting piping, nor the dedicated (local) operator, are subject to any increased ris. Will the activity create or increase the levels of radiation or airborne radioactivity? [ ] Yes [Xl No IJill that change result in a significant unreviewed envirorrnental impact, a significant increase in occupational exposure, or significant change to the dose to operators [ ] Yes CX] No performing tasks outside the filtered air bol..ndary during a design basis accident (GOC-19)? Explai The Superintendent-Radiological Protection lll.lst approve the explanation for "Yes" answer The secondary system is not radioactive; neither will the use of MANUAL OVERRIDE threaten the integrity of any systems which interface with the secondary system (specifically, the RCS).
28A. Superintendent Radiological Protection Signature 288. Date 2 Could the activity change or decrease the effectiveness of the emergency plan? Explai [ l Yes [X] No The Coordinator-Emergency Preparedness nust approve the explanation for 11Yes 11 answer This is a normal maintenance activity that does not affect the Emergency Pla A. Emergency Preparedness Coordinator Signature 298. Date 3 Will the consequences of failure for this activity affect the ability of systems or C l Yes CX] No c~nents to perform safety functions? Describe the modes and consequences of failure considered during this evaluation * . Failure of this activity would result in failure of early feedwater isolation following a main steam line brea However the presence of redundant operator actions to isolate the MOV and MFRV, combined with the limited AOT serve to preserve t overall reliability of the feedwater isolation ft.nCtio Calculation SM-899 shows that the ill1)8ct on feedwater isolatio,, reliability due to terrporary placement of the valve on the jack is negligibl e Safety Evaluation
Page 5 of 12 VIRGINIA POWER VPAP-3001 Rev. 2 GO\\ Part C - * I tflllS Considered* Sy Thi 8.safety Ewiluation (C<<JtiNJedl:\\'... 31. - Will the activity cause equipment to be exposed (or potentially exposed) to adverse conditions, including those created by t~rature, pressure, hunidity, radiation or meteorological conditions? [Colaitllent 3.2.11) If Yes, could these conditions lead to equiJ:CIIE!nt failure or a dangerous atmosphere? Explai [ l Yes [Xl No
C l Yes [Xl No
No system will be subject to any conditions outside of its design parameter Inmobilizing the MFRV's will not affect th anbient or accident conditions in the Mechanical EquiJ:CIIE!nt Roo Excessive cooldown of the RCS and contairment mass addition will be prevented via FW isolation in the form of the MFRV's and FW isolation MOV's closing on operator action e the main feed purps tripping on an SI or SG hi*hi level signa The MFRVs are designed for use of a MANUAL OVERRIDE functio. Could failure of the activity feed back into protective circuitry? Explai C l Yes [XJ No
Placement of the valve on its jack has no interface with any protective circuitry. It is a mechanical manipulation which supports the recovery of stable feedwater contro. could failure of the activity feed back into control circuits irrportant to stable plant operation (e.g., feedwater control, control rods)? [Colaitaent 3.2.12) C l Yes
[XJ No Placement of the valve on its jack has no interface with any control circuitr However, it does prevent operation in automatic SG level control mode for the affected S Operation with manual feed flow control to one SG does not preclude stable plant operatio Placing the valve on the jack is a mechanical manipulation which supports the recovery of stable feedwater contro. could the activity affect emergency diesel generator sequencing logics (including testing logics), or other logics irrportant to safet [Colllitaent 3.2.8) [ J Yes
[XJ No Placement of the valve on the jack has no interface with any safety-related logic It is a mechanical manipulation onl) 35. Could the act1v1ty cause a loss of separation of instr1.111ent channels/trains or electrical power supplies? Explai Placement of the valve on the jack has no interface with any instrunents or power Sl.4)1)lie. Will the act1v1ty involve the addition or deletion of any loads on the Class 1E electrical distribution system? Explai No electrical loads will be added or deleted. The electrical distribution is not affecte C J Yes [XJ No
[ J Yes [XJ No
e Safety Evaluation
Page 6 of 12 VIRGINIA POWER VPAP-3001 Rev. 2 GOVC
- .*,i_**.-.ic" 3 Will the activity adversely affect the ability of a system or c~nent to maintain its integrity or code requirements?
Explai C ] Yes CX] No - Placing a MFRV on its jack is within the design parameters for the Feedwater syste FW isolation capability will still be maintained via manual closure of the MFRV and the feedwater isolation MOV's and manual or automatic feed~ trip; this ensures that the RCS cooldown and contairment mass addition (during an inside-contairment MSLB) are limite. Will the activity reconfigure, eliminate, or add c~nents end/or piping to the single or two-phase erosion/corrosion piping inspection program? Explai [ ] Yes
[Xl No No piping changes will be mad. Will additional surveillance requirements, es defined in the Technical Specifications, be necessitated by the activity? Explai C J Yes [)(] No Existing surveillance procedures acc~lish testing of main FW l)U1'J trip; Partial stroking of the isolation MOV as part of the procedure serves to verify its operabilit. Will the applicable Technical Specification basis description be altered by the activity? [ l Yes CX] No Explai The overall reliability of the Feedwater Isolation c~t of the RPS/ESF system is not degraded by placing a MFRV on its jack, subject to the constraints and limitations set forth in this evaluation (as demonstrated in Calculation SM-899).
MFW isolation can still be obtained via re<iradant automatic and manual actions. There is adequate time to perfon11 the manual actions, as shown in Calculation SM-89 Scenarios do exist in which isolation may not occur within the TS requirec time; however, calculations have shown these occurrences to be statistically insignifican. Will the activity result in a violation of any Limiting Condition for Operation (LCO), as defined in the Technical Specifications? Explai C l Yes [)(] No The requirement for feedwater isolation is established by the safety analysis but not specifically addressed in the Technical Specifications (except for initiating instrumentation).
The safety analysis requirement is satisfied by the operability of the FW isolation MOV's and tripping of the main FW ~- The l)U1'J discharge MOV auto-closure following e ~ trip is not tested to a specifc time requirement; however, it does provide additional assurance of positive closur. Were any other concerns or items identified during this review? If 11Yes," explai [ l Yes [)(] No
- Safety Evaluation Page 7 of 12 VIRGINIA POWER VPAP-3001 Rev. 2 GO' Pert**.c.. - 1u.s,:conei<lered*.ay,,,This:1lafety:*.Evaluation;::(CantfraJed>,.,,* Items 43 through 62 consider potential irrpact VPAP-3001 provides engineering review guidelines for these item If the answer to any of the questions for these items is "Yes," a detailed engineering review nJst be performe The results of the detailed review should be docunented one supplemental page, identified by this safety evaluation nllli:,er and Part C It, l"IUIDe *.~/:: StatlavSec:uMty{::f Will the activity deactivate a security-related system or breach a security barrier? A. Will the activity add or eliminate a significant amount of conixJstible material frOIII plant areas? e. Will the activity change or affect any plant structure or barrier that acts as a fire berri er? c. Will the activity ill'P8ct the performance of an existing fire protection or detection system? 0. Will the activity involve modifying any COll'p()nent required for Appendix R, or any Appendix R support system such as emergency lighting or emergency power supplies? E. Will the activity change or affect system flow paths shown on Appendix R flow diagrams (North Anna Power Station - 11715/12050-DAR-Series and Surry Power Station - 11448/11548-DAR-Series)? F. Will the activity change station equii:ment arrangement drawings that show Appendix R equipment (North Anna Power Station - 11715-FAR-Series and Surry Power Station - 11448-FAR*Series)? A. Will the activity adversely affect any Class 1E electrical equipment located in a potentially harsh envirorment (as designated by the Environnental Zone Description)? B. Will the activity have the potential to alter any of the environnental parameters identified in the Envirormental Zone Description? C. Will the activity have the potential to affect any of the Class 1E electrical distribution systems (e.g., 4KV, 480V, 120V(AC))? D. Will the activity add, eliminate, or have the potential to affect ASME Section XI equipment adversely? E. Will the activity change a setpoint in the Precautions, Limitations, and Setpoints (PLS) Docunent? F. Will the activity adversely affect equipment on the EQML or a-List? Could the activity be adversely affected by a seismic event, or could the activity affect surrounding equipment during a seismic event? C l Yes [ ] Yes [ ] Yes
[ ] Yes [ ] Yes
[ ] Yes [ l Yes [ ] Yes
[ ] Yes
[ ] Yes
[ ] Yes [ ] Yes C l Yes [ ] Yes
[Xl No [Xl No [Xl No [Xl No CXl No [X] No [X] No [X] No [X] No CX] No [X] No [)(] No CX] No OCl l\\lo 471'Muimn,,:faciofi:.<:::::'::::::: *==:=
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A. Will the activity change instrunentation or controls in the Control Room or on the [ ] Yes [X] No auxiliary shutdown panel? B. Will the activity alter the Control Roan or the auxiliary shutdown panel? [ l Yes [Xl No
e Safety Evaluation Page B of 12 VIRGINIA POWER VPAP-3001 Rev. 2 ~- ** Safety,=*~a~:=,J)ispl~=,,"$)1$tai/C.rgesir:y,Response,,FaciH Will the activity change any of the equipment associated with the SPDS/ERF, including SPDS/ERF c~ter inputs?
- '* Statjon c~rs=t'::::::
Will the activity have a significant potential to modify or add software to station c~ters? A. Will the activity irrpact more than one-fourth of an acre of land, work in navigable waters, wells, dams, or wetlands, and/or involve any wastes or discharges? B. Will the activity involve changes to site terrain, features, or structures? c. Will the activity have a significant potential to expose safety related equipment to flooding via fluid system equipment/piping malfunction or failure? Will the activity have a significant potential to modify equipnent and/or instrunentation associated with Regulatory Guide 1.97 variables? C l Yes
C l Yes
C ] Yes C ] Yes C ] Yes
C l Yes
GOV CX] No CX] No CX] No CX] No CX] No CX] No
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- entHai1.on11iif ):oniii tfomriEJ',::=:;? : /:=:::::* ;:: *=:':? '?(?>':::???':???\\='??????: ??'?'???::==:?,:?:???:::t????>'??'?\\?t??ttE711?=8tM!Wt"!"f='?
A. Will the activity have a significant potential to increase the heating or cooling loads Cl Yes D<l No in plant areas and/or to plant equipment?
8. Will the activity change the existing ventilation system in any way? C. Will the activity change any building walls, ceilings, windows, doors, or floors, in a way that may affect existing HVAC systems? C l Yes
C l Yes
D<l No D<l No >~"\\.:*:Heew't.oads:(:::::::.:*:::*:.*:*)=:,::*::* :.:.*:.. :(:,,.:* ',,<}: ::)/\\.:**::,,,:'===,:=,*::: ::::::,::..... :*,:.::::::*:::..* Ji =.:: '.*.** :::.... ::.::... :.:...*. )... \\.\\....... :.:......:........,..,.*.,*...:..,..,.....,................... :~'.:_\\,;,.,...,:.::.:::::: Will the activity involve heavy loads (including the transfer of heavy loads in areas housing safety related equipment)? Will detrimental materials be introduced into the contairment or other plant areas? C l Yes r l Yes 00 No 00 No $.$~:*:*:=:~* t'*'.M~t,:::Mi.:t,~e::~,::*= ::*::::=:::::=::= :::=::'::=:::-:=:::::::=:::::::::::::::::::::::::::::::::::::::::::::::::::::::::::,;::::::=:::::::::::::::::::::::::::::::::=:::=::::::::::':::::::::::::::::::::::==::=:::::=:=:::::::::=:::::::::::=::::::::::=::::::;:::::::::::::::::::::::;:::::::::::::::,Mlttt:n:::r::::: Have ALARA concepts been included? (Detailed explanation not required.)
CX] Yes [ ] No
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Will the proposed change adversely i"'PSct the current systemtcoopo11ent capacities [] Yes 00 No or design performance? /57~.*.*.*:oesign *a.s *.;**~, .*... **.**::*;*.*.***)/{L***:***.**::::(:,;*::.*::: ****,:*.**:****-::***:*******:,.**..,*.*.**..,.*'.. :.,.. *.. *.*.*.,.:.:.. :.. :... ',.*.*,.. *... *.:.:.. *.*.*..... *.*.*... *.*.*:.*.*, *. *.:.*.*:.,.:.*..... *........* :.:.....:.. :.. :... :. b>.:.:i.:.;;fitirdtL{. Will the activity change applicable sections of the station design basis doc~t? [] Yes [X] No If a change to the Control Roan or Safe Shutdown Panel is considered, will the change need to be replicated in the silllJlator? r l Yes 00 No
e Safety Evaluation
Page 9 of 12 VIRGINIA POWER VPAP-3001 Rev. 2 Part C - ltem**Considered*Sy**This*Safety Evaluation*(Contirued)*(
59.**.* llluclear-**Naterials*.-Contr.ot.. :.... *.- Will the activity result in the procurement of special nuclear materials or change the handling or storage of special nuclear materials? Will the activity affect a masonry block wall in any way, either through addition, removal, mounting of equipment, or location of safety related equipment within the vicinity of a block wall? Will the activity create a potential hazard/chemical release? Will the activity affect station labeling? Is Management oversight of infrequent tests or evolutions (as defined by VPAP-0108, Infrequently Conducted or Corrplex Tests or Evolutions) reconrnended? GOVc C l Yes [Xl No C l Yes [X] No C ] Yes [Xl No C l Yes [X] No C l Yes CX] No
. t, 0 - e Safet}t Evaluation Page 10 of 12 VIRGINIA POWER VPAP-3001 Rev. 2
- .:::/,=**
.*,.*,****** 6 Which accidents previously evaluated in the Safety Analysis Report were considered? ~ 14.2.7, Excessive Heat Removal Due to Feedwater System Malfunctions 14.2.11, Loss of Normal Feedwater 14.3.2, Rupture of a Main Steam Pipe Other SI-design basis events (LOCA, SGTR, Rod Ejection) A. Could the activity increase the probability of occurrence for the accidents identified above? State the basis for this conclusio GOVC C l Yes [X] No Placing the MFRV on its jack eliminates the possibility of control system malfunctions that can cause a feedwater syster transien No other precursors to any Chapter 14 accidents have been affecte The use of MANUAL OVERRIDE is perfo!'ffle( only to support required maintenance on a degraded SGWLC system; the net effect is the recovery from an adverse conditio B. Could the activity increase the consequences of the accidents identified above? C l Yes CX] No State the basis for this conclusio Accident mitigation systems remain unaffected. Because performance of required SGWLCS maintenance iq,roves plant stability (alternately, a power downreq> may occur to isolate the HFRV) and a wide variety of redundant c~t and procedural features exist to assure FW isolation (required for MSLB mitigation, the most limiting accident), the consequences of any Chapter 14 event are not increased by the use of MANUAL OVERRID c. Could the activity create the possibility for an accident of a different type than was previously evaluated in the Safety Analysis Report? State the basis for this conclusio C l Yes [Xl No Feedwater system malfunctions remain the only credible events that could occur as a consequence of placement of a MFRV on its jac The redundant and diverse administrative controls on the isolation features provide adequate c~atory capability for the limiting overcooling event (i.e., the MSLB).
6 What malfunctions of equipment related to safety, previously evaluated in the Safety Analysis Report, were considered? Loss of Feedwater system control or auto-Isolatio A. Could the activity increase the probability of occurrence of malfii,ctions identified above? State the basis for this conclusio C l Yes [X] No Enginering Calculation SM-899 has shown that placing a MFRV on its jack for up to 72 hours negligibly i~cts the probability of a MFRV failure. *The maintenance restores SGWLCS operabil f ty, thus reducing the probability of any accident due to the previously degraded plant condition. Finally, extensive FW isolation red\\rdancy exists in the various plant systems and emergency procedure.. ( e -
Safety Evaluation Page 11 of 12 VIRGINIA POWER - VPAP-3001 Rev. 2 GOVC B. Could the activity increase the consequences of the malfunctions identified above? C l Yes CX] No State the basis for this conclusio No other event mitigation equipment has been affecte The potential consequences of leaving the MFRV in an unstable condition are much more severe than the negligibly small, potential consequences of briefly placing the valve on its jac Further, considerable isolation redundancy exists in various plant systems and EOP's to protect against secondary overcoolin c. Could the activity create the possibility for a 11111lfunction of equipment of a different type than was previously evaluated in the Safety Analysis Report? State the basis for this conclusio C l Yes [X] No MFRV inmobilization or a spurious auto-closure or opening are the only plausible failures associated with the OVERRIDE proces These failures bound any other consequences of MFRV maintena_nce events, including accident. Has the margin of safety of any part of the Technical Specifications as described in the bases section been reduced? Explai [ ] Yes CX] No The safety analysis requirement on feedwater isolation protects against 1) reactivity addition via generator overcooling and 2) mass and energy addition to the containnen The diverse and redundant isolation capability, as well the limited allowed outage time ensure that the margin of safety is maintaine. Does the proposed change, test, or experiment require a change to the Technical Specifications? Explai [ ] Yes CX] No TS c~liance is not c~romised by placing a MFRV on its jac The feedwater isolation requirement is met by marual MOV closure, manual MFRV closure and feed J:IU1'> trip. Scenarios have been postulated in which early isolation might not be 111tt; however, these have been analyzed in Calculation SM-899 and shown to be statistically insignifican :~tiiln&':#l,,:anci='6?:,,#Pty,:~1y:,,~11;::,f).:~f!:)~~~:,;~7t,'-':=:!~ll~tj~:if~rfa~:1Y#~~J/\\==Fttt=tS=t~:,:i:i:j,j,,,:jt,{{ttt,:i:~:j,::~iii::,~,:{{:tii@::,,:fa@fafa 6 Does the proposed change, test, or experiment involve a significant unreviewed [] Yes [] No envirormental i~ct? Explai CX] N/A 6 Does the proposed change, test, or experiment involve a significant increase in occupational exposure? State the basis for this conclusio [ l Yes CX] N/A [ ] No
~ .. * e Safety Eva.an Page /2 of 12 VIRGINIA POWER VPAP-3001 Rev. 2 "GO' j Part O * Unrevieweo SafetyQuest:fon 0etenirinat'io11 (Continued),,,.,. I !f all resoonses are 'No" to Questions 64 througn 69. the propose<:! activity may De implemented following SliSOC apl)l'OVII All reiatea aocumentation must be retaine i !fa resoonse 1s "Yes* ~~ any part of Questions 64 through 67. an operating license amendment nust be approved by the~ before tne cnange, test. or experiment may De 1mplemente If a response is "Yes* to Question 68 or 69. an application for an ISFSI license amendment 111JSt be approved before the change. test. or exoerimem: may be irnolemente. Rev1ewer Name (Print> 7 Reviewer Title I J. O. Erb System Engineer 17 Reviewer Signature Q ('4 I G, I 7 Date 7 /13 /93
- 7 Ces1gn Autnor1 ty Reviewer Name (Print>
? Design Autnority Reviewer Title ~\\IV - ~vc..le.u s~~ l\\u,~..,~~ 7 n. Date }}