ML082600578: Difference between revisions

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#REDIRECT [[L-08-241, Attachment 1, Applicable Portions of the U.S. Nuclear Regulatory Commission (NRC) and Davis-Besse Nuclear Power Station (DBNPS) Improved. Technical Specification (ITS) Conversion Website, Section 3.5 RAIs]]
| number = ML082600578
| issue date = 08/26/2008
| title = Attachment 1, Applicable Portions of the U.S. Nuclear Regulatory Commission (NRC) and Davis-Besse Nuclear Power Station (Dbnps) Improved. Technical Specification (ITS) Conversion Website, Section 3.5 RAIs
| author name =
| author affiliation = FirstEnergy Nuclear Operating Co
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000346
| license number = NPF-003
| contact person =
| case reference number = FOIA/PA-2010-0209, L-08-241, TAC MD6398, TAC MD6399, TAC MD6400
| document type = - No Document Type Applies
| page count = 512
| project = TAC:MD6398, TAC:MD6399, TAC:MD6400
| stage = RAI
}}
 
=Text=
{{#Wiki_filter:Section 3.5 RAIs NRC ITS Tracking Page I of 6 Return to View Menu Print Document RAI Screening Required:
Yes Status: Closed This Document will be approved by: Tim Regulatory Basis must be included in Comments Kobetz section of this Form This document has been reviewed and Yes information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING NRC Reviewer ID 1200710032123 Conference Call Requested?
No Category BSI -Beyond Scope Issue ITS Section: TB P.O.C:. J-FD.Number:ý Page. Number(s);:.
ITS 3.5 Ross Telson None 1 5 Information ITS.Number:
OS1.:.
Bases JFD Number:;.3.5.1 5 L.1 3 REF: Attachment 1, Volume 10, Rev. 0, Pgs 5,7,8,9,12,18,21 CTS LCO 3.5.1 b. and d. [correlating to ITS SR 3.5.1.2 and 3.5.1.3 and associated bases]---- ACTIONS NEEDED: 1. Provide analyses associated with CFT level and pressure parameters.
Include the SAFETY LIMITS, ANALYTICAL LIMITS, ALLOWED VALUES, all uncertainties, and bases. Include background discussion addressing the change in uncertainty, when it was recognized, why it changed, how the proposed uncertainty calculations differ from calculations in the current licensing basis. Explicitly identify those uncertainties which were included in the original ALLOWED VALUES in order to ensure SAFETY VALUES were not exceeded and those that would be relocated from TS to licensee-controlled surveillance procedures if the proposed ANALYTICAL LIMITS were substituted for ALLOWED VALUES.2. Revise SR 3.5.1.2 and 3.5.1.3 and associated TS bases to reflect appropriate ALLOWED VALUES (rather than ANALYTICAL LIMITS) based on the proposed licensing basis change provided above. Identify and similarly revise any other proposed ITS parameters or discussion in which ANALYTICAL LIMITS are substituted in place of ALLOWED VALUES.---- BASIS FOR REQUEST: L01 reports that the original CFT borated water level and nitrogen pressure ANALYTICAL LIMITS have not changed since issuance of CTS but that recent calculations indicate additional uncertainty between those limits and the CTS-specified ALLOWED VALUES is warranted.
The proposed change does not incorporate the warranted additional uncertainty nor does L01 provide the associated analyses included in the safety http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsfl 1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 6 analysis report that is necessary to support staff evaluation of the proposed change.The proposed ITS deviates from Babcock and Wilcox Plants, Rev. 3 STS (NUREG-1430), Vol. 1, in that it proposes to replace ALLOWED VALUES (used directly by operators and inspectors to assess OPERABILITY) with ANALYTICAL LIMITS Which would require operators and inspectors to determine the uncertainty delta and add it to- or subtract it from the proposed ITS ANALYTICAL LIMITS in order to determine CFT OPERABILITY.
This change unnecessarily increases burden on operators and inspectors and increases the likelihood of error in assessing OPERABILITY.
It is incorrectly characterized by JFD 1 and Bases JFD 3 as "The brackets have been removed and the proper plant specific information/value has been provided." Use of ALLOWED VALUES in the CTS and in the NUREG incorporates ALL uncertainties between the associated SAFETY VALUES and the TS-Specified ALLOWED VALUES, thus establishing precedent for the need to incorporate ALL these elements into the ITS. L01 states that the appropriate uncertainty is reflected in the (licensee-controlled) surveillance procedure Comment acceptance criteria.
This constitutes relocation of certain uncertainties from TS to a licensee-controlled surveillance procedure.
This relocation is not supported by a RELOCATED SPECIFICATIONS DOC.Significant discussion and disposition of philosophical and technical issues between the industry and the NRC occurred during the development of the ISTS. Therefore, a high threshold should be satisfied for deviating from the ISTS in the ITS. Language and format'preferences, unless justified on a plant-specific basis, should be avoided. [NEI 96-061---- REGULATORY REQUIREMENT:
§ 50.36 Technical Specifications (a) ... A summary statement of the bases or reasons for [technical]
specifications
... shall ... be included in the application...(b) ... Thetechnical specifications will be derived from the analyses and evaluation included in the safety analysis report...(c) Technical specifications will include..
.(3) Surveillance requirements....
to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.issue Date 10/03/2007 Close Date 01/25/2008 Logged in User: Anonymous vResponses I. .1 NRC Response by Ross Telson In a 10/15/07 email the licensee indicated the desire to further on 10/16/2007 discuss this question with the originator during a 10/17/07 conference call, indicating:
We feel our suggested approach is more in keeping with the typical approach in the Standard.
A few examples would include: -UHS Temperature, -UHS Level, -BWST Temperature, -EDG Fuel Oil and Lube Oil. -Also, our approach is identicle to what was approved as part of the Beaver Valley Conversion on their ITS 3.5.1. It makes sense to specify the http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/lfddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 3 of 6 actual limit in the Technical Specifications, and then allow the surveillance procedures to account for instrument uncertainty and readability of the available indications to assure the limit is maintained.
--------------
Reviewer requests the licensee determine which of the above examples they deem to constitute applicable precedents to the proposed BSI and submit via this database, or through formal correspondance, an appropriate evaluation showing the applicability of the precedent(s) to BSI-5.The appropriate references should be attached.
Applicable ML#'s, page numbers, paragraphs, etc. should be referenced.
This additional request does not negate the previous requests.Licensee Response by Bill Bentley on 10/17/2007 As part of a phone call between WJBentley and Ross Telson on 10/17/07, WJBentley stated thatDavis-Besse had submitted an LAR to the NRC previously to change the Core Flood Tank pressure and level limits in the Tech Specs to the analytical limits, but had withdrawn that LAR. That verbal statement was not correct. After reviewing the history more closely, an LAR package was started, but was never submitted to the NRC. A management decision was made that the requested changes would be submitted as part of the ITS Conversion amendment.
This response is posted to clearly correct the erroneous information that was verbally communicated.
Licensee Response by Bill This is a test to try and upload a file.Bentley on 10/19/2007 Licensee Response by Bill This is a test to try and upload a file.Bentley on 10/19/2007
_____________________________
Licensee Response by Jerry Jones on 10/31/2007 Action 1 Response The following documents are provided to address the information requested in Action 1. All names of individuals have been redacted in the documents.
: a. Condition report 05-00085 -problem and investigation results. b. Portion of C-NSA-064.02-036, Rev. 1 (Davis-Besse 1 LOCA Summary report) -provides the Core Flood tank (CFT) LOCA analysis input parameters for volume and pressure.
: c. C-ICE-051.01-002, Rev. 0-provides the acceptance criteria to be used for the Surveillance associated with the CFT pressure.
: d. PN(Post it Note) to C-ICE-051.01-002
-identifies that a correction needs to be made to the calculation for the correct analytical value (567.3 psig vs 567 psig), but that the final results of the calculation are not affected.
e.C-ICE-05 1.01-001, Rev. 0 -provides the acceptance criteria to be used for the Surveillance associated with the CFT level (i.e., volume). The documents are included in the first attachment (200710032123 Action 1 Information.pdf).
Furthermore, due to the conversion error described in document d above (converting from psia to psig), the SR 3.5.1.3 will be changed to reflect the nitrogen cover pressure in psia, (i.e., 582 psia to 648 psia), exactly consistent with the LOCA analysis assumptions.
This avoids placing any conversion errors in the Technical Specification in converting from psia to psig. This conversion will be addressed in the site specific calculations for the surveillance procedure.
A draft http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 4 of 6 markup regarding this change is attached (Action 2 markup.pdf provided in the next response).
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
Licensee Response by Jerry Jones on 10/31/2007 Action 2 Response ISTS 3.5.1 includes limits for CFT borated water volume (SR 3.5.1.2), nitrogen cover pressure (SR 3.5.1.3), and boron concentration (SR 3.5.1.4) (Volume 10, Page 12). The ISTS Bases, Applicable Analyses Section (Page 18), describes the basis for the limits. In the ISTS, it states that the limits are corrected for instrument inaccuracies.
However, instrument inaccuracies of parameters in Surveillance Requirements are not normally included.
The values in Technical Specifications should be the analytical limit, with the utility correcting for any uncertainties in the plant procedures performing the Surveillances.
For instance, the Borated Water Storage Tank (BWST) Technical Specification in this ECCS Section includes three Surveillances on similar parameters (SRs 3.5.4.1, 3.5.4.2, and 3.5.4.3).
The ISTS Bases, Applicable Safety Analyses Section (Page 92) specifically states that the numerical values of the parameters stated in the SR are actual values (i.e., the analytical limits) and do not include allowance for instrument errors. Thus, the utility has control of ensuring instrument error is properly accounted for during performance of the Surveillances.
Davis-Besse believes that the parameters for the two ECCS tanks (CFT and BWST) should be treated the same. Therefore, the proposed values in the Davis-Besse ITS are the analytical values, as described in Discussion of Change (DOC) L01. DOC LO 1 provides the justification for changing the current values of the CFT water volume and nitrogen cover pressure.
The NRC reviewer also commented that Justification for Deviation (JFD) 1, which provides the justification for providing the Davis-Besse volume and pressure values, only stated that the brackets were removed and the plant specific values are provided.
No mention of the values being changed from the current values to.the proposed values was provided in the JFD. Therefore, a new JFD will be provided for these two SRs (ITS SR 3.5.1.2 and SR 3.5.1.3) explaining that the proposed values are the analytical limits as justified in DOC LO 1.A draft markup regarding this change is attached (Action 2 markup.pdf).
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
Licensee Response by Bill The following was discussed verbally during a phone call today Bentley on 11/14/2007 with Ross Telson, and is posted here to describe Davis-Besse's current intent. I stated during the phone call that our plans now are to provide a change to the original submittal for the Core Flood Tank pressure and volume limits provided in ITS Specification 3.5.1. The limits placed in the Technical Specification will be limits that have been adjusted for instrument uncertainty.
Formal mark-ups of the original submittal, with appropriate changes to the CTS, DOCs, ISTS, and ISTS Bases will follow at a later date.NRC Response by Ross Telson Licensee Response on 10/31/2007 by Jerry Jones: Action 1 http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 5 of 6 on 11/15/2007 Response states that SR 3.5.1.3 will be changed to reflect the nitrogen cover pressure in psia. Please confirm that the instrumentation which is relied upon to meet SR 3.5.1.3 displays in psia (or why the TS SR is represented by a pressure scale different than that of the relied-upon instrumentation).
Licensee Response by Bill Bentley on 11/16/2007 The core flood tank pressure indications read out in psig. The previous responses postrd by Jerry Jones for Action 1 and Action 2 were based on the previous intent to use analytical values for pressure and volume. When we provide a reponse that includes the corrections for the markups to CTS, DOCs, ISTS and ITST bases such that the Tech Spec limits include applicable instrument uncertainty
-we will make it clear that the Jerry Jones responses on 10/31/07 are superceded.
Licensee Response by Jerry Jones on 12/26/2007 Based on further discussions with the NRC reviewer, the borated water volume and nitrogen cover pressure requirements will be changed to include instrument uncertainty.
The water volume limit will also be changed from gallons to feet, consistent with the actual indication provided for the operator.
Due to this change, the original Discussion of Change has been changed from less restrictive to more restrictive.
A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
The previous two responses posted by Jerry Jones on 10/31/07 are superseded by the above response.
However, the document provided with the previous response contained calculations that support the attached draft markup. The previously provided calculations are summarized below: a. Portion of C-NSA-064.02-036, Rev. 1 (Davis-Besse 1 LOCA Summary report) -provides the Core Flood tank (CFT) LOCA analysis input parameters for volume and pressure.
: b. C-ICE-051.01-002, Rev. 0 -provides the acceptance criteria to be used for the Surveillance associated with the CFT pressure.
: c. PN (Post it Note) to C-ICE-051.01-002
-` identifies that a correction needs to be made to the calculation for the correct analytical value (567.3 psig vs 567 psig), but that the final results of the calculation are not affected.
: d. C-ICE-05 1.01-001, Rev. 0 -provides the acceptance criteria to be used for the Surveillance associated with the CFT level (i.e., volume).NRC Response by Ross Telson on 01/25/2008 Staff have received and reviewed attachments to Licensee Response by JerryJones on 10/31/2007 and 12/26/2007.
Staff understand that the 12/26/2007 response superceeds two Licensee Responses by Jerry Jones on 10/31/2007 but not the attached analyses associated with CFT level and pressure parameters.
These analyses are now intended to support the 12/26/2007 response and attachment which revisesrevise SR 3.5.1.2, 3.5.1.3, and associated DOC's and TS bases to include instrument uncertainty, to establish the water volume limit in feet, consistent with operator indication, and to address this beyond-scope change as a more restrictive change. Based on confirmation from SRXB and EICB technical branch reviewers, the originator has no further questions regarding http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/
1 fddceal Od3bdbb5 85256e...
7/17/2008 NRC ITS Tracking Page 6 of 6 this question thread at this time. Should further questions arise, they will be posted in a new thread.Date Created: 10/03/2007 09:23 PM by Ross Telson Last Modified:
01/25/2008 01:33 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NOP-LP-2001-01 Site: G201ýCONDITION REPORT cR Number 1 05-00085 TITLE: CORE FLOOD TANK LEVEL TECHNICAL SPECIFICATION DISCOVERY DATE TIME EVENT DATE TIME SYSTEM I ASSET#1/5/2005 N/A N/A N/A 051-01 N/A EQUIPMENT DESCRIPTION Core Flood Tanks FLOC System FLOC DESCRIPTION OF CONDITION and PROBABLE CAUSE (if known) Summarize any attachments.
Identify what, when, O where, why, how.R During a review of calculation C-ICE-051.01-001, Uncertainty Calculation for Core Flood Tank Level I Indication (approved on 7/17/03) it was identified that the surveillance acceptance criteria that were G developed in this calculation are more restrictive than the existing Technical Specification (TS) 3.5.1 requirements for Core Flood Tank (CFT) contained volume. This is because these new acceptance criteria include the measurement uncertainties that could exist in the level indication, and the actual N uncertainty is larger than was accounted for in the TS values. These new acceptance criteria have A been implemented in procedure DB-OP-03006, Miscellaneous Instrument Shift Check, and they T ensure that the volume that was used in the LOCA analysis is actually available.
Based on this, I there is no concern about CFT operability.
However, the following issues need to be addressed; 0 N 1. The current TS limits for CFT inventory are non-conservative and do not reflect the most recent uncertainty calculation.
: 2. CR 03-02547 included an evaluation of past operability for this issue. However, based on the detailed information provided in calculation C-ICE-0511.01-001 relative to the magnitude of the uncertainty in the CFT level indication, it is recommended that past operability be re-evaluated relative to the historical use of the non-conservative TS values.3. It is recommended that an extent of condition review be performed.
IMMEDIATE ACTIONS TAKEN / SUPV COMMENTS (Discuss CORRECTIVE ACTIONS completed, basis for closure.)As stated above there is no identified concern with CFT operability.
The issue was discovered during preparation of the LOCA Analysis Input Summary for a new fuel design. Fleet Licensing personnel are evaluating the need for compensatory measures in accordance with the guidance provided by NRC Administrative Letter 98-10 (e.g., TS "pink sheet").QUALITY ORGANIZATION USE ONLY IDENTIFIED BY (Check one) [] Self-Revealed ATTACHMENTS Quality Org. Initiated
'E Yes V Individual/Work Group D Internal Oversight Quality Org. Follow-up D1 Yes D No El Supervision/Management D External Oversight D Yes W No ORIGINATOR ORGANIZATION DATE SUPERVISOR DATE PHONE EXT.DBDM 1/5/2005 1/5/2005 8567 Page 1 of 2 NOP-LP-2001.-01 Site: G201 CONDITION REPORT CR Number 05-00085 TITLE:. CORE FLOOD TANK LEVEL TECHNICAL SPECIFICATION SRO EQUIPMENT OPERABILITY ORG. IMMEDIATE ORG. !MODE CHANGE P REVIEW OPERABLE ASSESSMENT NOTIFIED INVESTIGATION NOTIFIED rRESTRAINT L 66 Yes El No W Yes ELNo [] NI El Yes W No N/A E] Yes WJ No N/A Eil Yes 0- No A I MODE ] ASSOCIATED TECH SPE 'C NUMBER(S)
ASSOCIATED LCO ACTION STATEMENT(S)
N i.7 T I2 0 DECLARED INOPERABLE REPORTABLE?
One Hour N/A APPLICABLE UNIT(S)EP II (Date N/ A Tim e) ,,RE 7 -Ye 7] N (aeN/A EightHour N/A U1 L Both R I ]Eval Required ....A T COMMENTS I The condition report documents that the level that is monitored in the Core Flood Tanks (CFT) using o DB-OP-03006 meets both current Tech Specs and the calculation referenced in the CR. The N condition report also documents that the current Tech Spec CFT capacity requirements have not S been updated to reflect the most conservative values needed. As documented in supervisor comments, Corporate Licensing has been contacted to investigate and make adjustments to the Tech Specs as needed. Therefore, Equipment Operable marked Yes. This condition needs to be evaluated for past operability and reportability therefore Reportable is marked EVAL. The supervisor of this condition report has contacted Reg. Affairs for the evaluation.
Current Mode -Uniit e e t I Current Mode -Unit 2 Power Level -Unit 2 uv1 100 N/A N/A SRO -UNIT 1 , SRO -UNIT 2 DATE__ __ _11/2005 CATEGORYSIEVAL ASSIGNED ORGANIIZATION DUE DATE R REPORTABLE?
AF DBDM 2/19/2005 E Gnr] Yes I t No os LER No.GI CP TREND CODES Comp Type /IID Cause uI CRPAL IREPORTABILITY REVIEWER II Process / Activity ICause Code(s) (if Cause T or W) Org A I11 TI SUPVI LP4 4000 NA .... NONE o0 DATE/I RI MRB I ' _ ... .Y 01/07/05 IINVESTIGATION OPTIONS .LOSED BY DATE I[] Maint.Rule
[:]OE Evaluation
[:] Generic Implications El Part 21 177/25/2006 Page 2 of 2 Attachment Site: G201 CONDITION REPORT CR Number 05-00085 REPORTABILITY DETERMINATION:
As previously described in the reportability determination of CR 03-02547 dated 4/9/03, the Technical Specification minimum allowable volume of 7555 gallons equates to an indicated level of 12.56 feet, and DB-OP-03006 specifies a minimum level of 12.6 feet. Similarly, the Technical Specification maximum allowable volume of 8004 gallons equates to an indicated level of 13.44 feet, and DB-OP-03006 specifies a maximum level of 13.3 feet. A review of the current revision DB-OP-03006 (revision 17 dated 12/15/04) shows these procedural minimum/maximum values have not changed since CR 03-02547 was initiated, and appear to have remain unchanged since at least 1999. The new values from the referenced calculation approved 7/17/03 match the current values of DB-OP-03006.
Operations typically maintains the Core Flood Tank levels at 12.8 to 13.1 feet, well away from the Technical Specification limits, so it is unlikely that a violation of Technical Specification requirements occurred as a result of this issue for at least the past three years required to be reviewed for reportability per 1 OCFR50.73(a).
The referenced calculation determines the instrument uncertainty associated with these instruments, and this CR (05-00085) identifies that when including these new instrument uncertainties, the new acceptance criteria are more restrictive than the existing Technical Specification requirements.
The actual quantities of these instrument uncertainties are unknown in the past, and there is no firm evidence that these uncertainties resulted in the subject equipment being outside of the Technical Specification, allowable values in the past. Based upon the fact that the equipment passed the Acceptance Criteria in place at the time, in accordance with the guidance of NUREG-1 022, it is assumed that no violation of the Technical Specifications occurred, so this issue is not reportable per the criteria of 1OCFR50.72 or 10CFR50.73.
Page 1 of 1 Site: G201 INVESTIGATION
 
==SUMMARY==
CR Number: 05-00085 NOP-LP-2001-06 Category / Eval: AF Assigned Organization:
DBDM Quality Followup Req'd: El Yes W] No~For: Fix Inv 4 tl 9~ilon ObIy.~Hardware
/Degraded Condition Resolution Required?
Y o If 4 Yes: ' Repair ~DScrapQ.IJ Acceptance of the CR Investigation signifies acceptance of the following items, as applicable:
Originator Identification Date Corrective Actions (listed below) (listed below, if any) (listed below, if any)Cause Analysis Generic Implications 10 CFR 21 Decision Checklist Acceptance of Investigation:
Date: Quality Approval:
Date: 2/17/2005 Site-VP Acceptance:
Date: Closure Comments: Problem: The Core Flood Tank Level Technical Specification values includes 75 gallons of instrument uncertainty above the analytical minimum of 7480 gallons and below the analytical maximum of 8079 gallons to establish the Technical Specification values. This results in Technical Specification values of 7555 gallons minimum and 8004 gallons maximum. However, due to the calculated instrument uncertainty in caculation C-ICE-051.01-001 being larger than the 75 gallons used to establish the Technical Specification values, the surveillance values are more restrictive than the Technical Specification values.Based on this the current TS limits for CFT inventory are non-conservative and do not reflect the most recent uncertainty calculation.
Investigation:
The Core Flood Tank Technical Specification was reviewed for consistency with other Technical Specification values. Not included in this review was Technical Specification values that have setpoints for automatic action associated with them, such as RPS, SFAS, and SFRCS. This is due to the specific method (ISA Standard 67.04.01 and Recommended Practice 67.04.02) of development of the Allowable Values and Trip Setpoints for those systems. Of the Technical Specification values reviewed, only the Core Flood Tank Volume and Core Flood Tank Pressure have Technical Specification values that include instrument uncertainty away from the analytical value. Other Technical Specification values, such as Ultimate Heat Sink Temperature, Ultimate Heat Sink Level, EDG Fuel Oil Storage Tank volume, and EDG Day Tank Volume, all are analytical, values, An analytical value was not retrieved for each Technical Specification value. However, the majority of the Technical Specification values were reviewed and the prevailing method is to not include instrument uncertainty in the Technical Specification values. The analytical value is equivalent to the Technical Specification value.A review of the B&W and Westinghouse "Standarized Technical Specifications" both had statements in the Bases for the Core Flood Tank Volume that the values included instrument uncertainty.
However, a review of the Crystal River (B&W plant) and Beaver Valley (Westinghouse plant) Technical Specifications revealed that the current approved Core Flood Tank Technical Specifications both excluded instrument uncertainty and reflected the analytical value.A discussion with the Chairman of the ISA 67.04 committee, the committee which developed the standard Page 1 of 3 Site: G201 INVESTIGATION
 
==SUMMARY==
CR Number: NOP-LP-2001-06 05-00085 for instrument uncertainty methods used at Nuclear Power Plants, stated that those types of values are reflected both ways in current Technical Specifications, but the majority are consistent with the analytical value.There is a statement in the CR that the Technical Specification values are non-conservative.
This is considered to be an incorrect statement because, if the Core Flood Tank Technical Specifications were similar to other Technical Specification values, the values would be 4680 gallons and 8079 gallons or the Analytical Values. Therefore, the tighter values currently in the Technical Specifications are considered to be conservative.
To be consistent with other Technical Specification values, the current Core Flood TankVolume and Pressure Technical Specification values should be revised to delete the instrument uncertainty and reflect the analytical values. Since the current Technical Specification values are conservative and the existing surveillance procedures reflect the correct volume and pressure'to ensure compliance with the Technical Specification values and the Analytical Values, there is no safety concern.The CR requested a review of past operability.
This was accomplished by the licensing review.The CR also requseted an extent of condition.
This is accomplished by Corrective Action #1.Corrective Actions: The MRB created a corrective action (CAF#1) for an extent of condition be accomplished.
As a part of CR 02-06407, Corrective Action 6, all Technical Specification and TRM values were evaluated for margin. If insufficient margin existed between the normal operating point and the Technical Specification value, an instrument uncertainty calculation was performed prior to startup from the long shutdown.
The Core Flood Tank Volume and Pressure were included in that review.Since inclusion of instrument uncertainty in the Technical Specification value is conservative and all Tech Spec and TRM values were previously evaluated for acceptable margin, there is no safety concern and no operational concern. In addition to the previous review in CR 02-06407, Corrective Action 6; Corrective Actions 8, 9, 10, and 11, require review of the Technical Specification and TRM instruments (in addition to many other instruments) and the development of calculations for those instruments.
As part of the calculation development, an analytical basis will be researched and documented in the calculation.
Any problems would be revealed during those calculation preparations.
Therefore, Corrective Action 1 should be closed with the review already performed considered acceptable performance of that Corrective Action.Corrective Action #2 was created for Design Basis Engineering Analysis to develop the License Amendment Request for the Core Flood Tank Volume and Pressure Technical Specifications.
This will require the Technical Specification values to be consistent with the analytical values by removing the incorrect instrument uncertainty value of 75 gallons from the core flood tank volume and 8 psig from the core flood tank pressure.Corrective Action #3 is required for Fleet Licensing to process the License Amendment Request through company and NRC approval after receipt of the input from Engineering for Corrective Action #21 Corrective Action 4 is for Fleet Licensing to determine if compensatory measures are required during the interim period of License Amendment development, review, and approval.Operability:
There is no operability concern. The current surveillance procedures correctly account for instrument error which protects the Technical Specification Values (which already include a portion of the instrument uncertainty) and the Analytical Values.Page 2 of 3 Page i~CALCULATION NOP-CC-3002-01 Rev. 01 INITIATING DOCUMENT (S) CALCULATION NO. I X ]VENDOR CALC
 
==SUMMARY==
NIA C-~NSA-064.02-036R0 N/A -6,R01 ENDOR CALCULATION No. 865006232-03 TITLE/
 
==SUBJECT:==
 
DB-1 LOCA Summary Report C] BV1 C BV2 DB PY Category 0 Active 0 Historical 0 Study Classification 0 Tier 1 Calculation 13 Safety-Related/Augmented Quality [I Nonsafety-Related Open Assumptions?
0 Yes 0 No If Yes, Enter CR Tracking Number System Number 064.02 Functional Location N/A Commitments:
N/A (Perry Only) Calculation Type: Referenced In Atlas? El Yes 0 No Referenced In USAR Validation Database [] Yes U No Computer Program(S)
Program Name Version / Revision Category Status Description WORD 2003 C Active text processing Revision Record Rev. Affected Pages' Originator/Date Reviewer/Design Verifier/Date Approver/Date 00 all JJ _ _ _ _ _ _ _ _ _ _ _ _Description of Change: Initial release.Describe where the calculation will be evaluated for 10CFR50.59 applicability.
RAD 06-00244 Rev. Affected Pages Originator/Date Reviewer/Design Verifier/Date Approver/Date 01 ijiiiiiiv, "tvc 2.3,5.8,9,10,11, , i U/M/0"7 13,44,55,56,60, ql 63,72,93.96,149, la w 150,163,166, 171,206 Description of Change: Revision 01 incorporates a sensitivity study of the delay time utilized for the Main Feedwater coastdown.
Miscellaneous editorial corrections have also been incorporated.
Describe where the calculation will be evaluated for 10CFR50.59 applicability.
RAD 06-00244 (attached) that was performed for Revision 00 remains applicable with the exception that a USAR change is not needed.Rev. Affected Pages Originator/Date Reviewer/Design Verifier/Date Approver/Date Description of Change: Describe where the calculation will be evaluated for 10CFR50.59 applicability.
Rev. Affected Pages Originator/Date Reviewer/Design Verifier/Date Approver/Date Description of Change: Describe where the calculation will be evaluated for I0CFR50.59 applicability.
Page ii CALCULATION N/A I C-NSA-064.02-036, R01[ X I VENDOR CALC
 
==SUMMARY==
VENDOR CALCULATION No. 86-5006232-03 TITLE/
 
==SUBJECT:==
 
DB-1 LOCA Summary Report TABLE OF CONTENTS SUBJECT PAGE COVERSHEET:
I OBJECTIVE OR PURPOSE iii SCOPE OF CALCULATION viii
 
==SUMMARY==
OF RESULTS/CONCLUSIONS ill LIMITATIONS OR RESTRICTION ON CALCULATION APPLICABILITY Iv IMPACT ON OUTPUT DOCUMENTS Iv DOCUMENT INDEX v CALCULATION COMPUTATION (BODY OF CALCULATION):
ANALYSIS METHODOLOGY 56 ASSUMPTIONS 13 ACCEPTANCE CRITERIA 13 COMPUTATION
-Large-Break 74 COMPUTATION
-Small-Break 142 RESULTS 13 CONCLUSIONS (page 1, paragraph no. 2) 1 ATTACHMENTS:
none N/A TOTAL NUMBER OF PAGES IN CALCULATION (COVERSHEETS
+ BODY + ATTACHMENTS) 213 Page SUPPORTING DOCUMENTS (For Records Copy Only)DESIGN VERIFICATION RECORD 0 Pages CALCULATION REVIEW CHECKLIST 3 Pages 10CFR50.59 DOCUMENTATION 2 Pages DESIGN INTERFACE
 
==SUMMARY==
1 Pages, DESIGN INTERFACE EVALUATIONS 2 Pages OTHER (Areva Design Verification Checklist) 2 Pages[] YES EXTERNAL MEDIA? (MICROFICHE, ETC.) (IF YES, PROVIDE LIST IN BODY OF CALCULATION)
[ NO[] NO
,,, ,,,Page iii NOP-CC-3002-01 Rev. 01 INITIATING DOCUMENT (S) CALCULATION NO. [ X I VENDOR CALC
 
==SUMMARY==
NIA C-NSA-064.02-036, R01 VENDOR CALCULATION No, 66-5006232-03 TITLE/
 
==SUBJECT:==
 
DB-1 LOCA Summary Report OBJECTIVE OR PURPOSE: This summary report documents results of the loss-of-coolant accident (LOCA) analyses performed for cores containing Mark-B-HTP fuel assemblies.
Both mixed-core (i.e., Mark-B12, Mark-B10K and Mark-BlOM) and full-core analyses of Mark-B-HTP fuel were performed.
A full-spectrum of break sizes and locations was analyzed.SCOPE OF CALCULATION/REVISION:
Revision 01 incorporates a correction to the delay time for the Main Feedwater coastdown.
The Revision 00 analysis utilized 4 seconds rather than the specified value of 2 seconds. The Revision 01 analysis performs a sensitivity study using Main Feedwater coastdown delay times between 0 and 12.5 seconds. Miscellaneous editorial corrections have also been incorporated.
 
==SUMMARY==
OF RESULTSICONCLUSIONS:
A summary of 10 CFR 50.46 (DIN 1) compliance for large-break LOCA analyses is provided in Table 3.1.A summary of 10 CFR 50.46 (DIN 1) compliance for small-break LOCA analyses is provided in Table 3.2.Linear heat rate limits for LOCA are listed in Tables 3-3 through 3-22.Peak clad temperatures are listed as a function of break size in Table 3-23 for small-break LOCAs.LOCA linear heat rate limits are plotted as a function of core bumup in Figures 3-1 through 3-18.The moderator temperature coefficient limit and associated power level dependent linear heat rate limit penalty are shown in Figures 3-19 and 3-20.Peak clad temperatures are plotted as a function of break size in Figure 3-21 for small-break LOCAs.The concerns of GSI-191 have been addressed by the references identified in DIN 2.Results of a sensitivity study performed for the delay time utilized for the Main Feedwater coastdown is provided on Page 150.The analyses summarized herein demonstrate that cores containing Mark-B-HTP fuel are in compliance with the five criteria of 10 CFR 50.46 (DIN 1) for both large-break and small-break LOCAs.Compliance details are discussed in Section 3.1.
i ii P age iv 6mfne CALCULATION NOP-CC-3002-01 Rev. 01 INITIATING DOCUMENT (S) CALCULATION NO.[X]VEORCLSUMY NIA C.NSA-064.02-036, R01 VENDOR CALCULATION No. 86-5006232-03 TITLE/
 
==SUBJECT:==
 
DB-1 LOCA Summary Report LIMITATIONS OR RESTRICTIONS ON CALCULATION APPLICABILITY:
DIN 2 documents the parameters and corresponding values and the operator actions that are controlled by FENOC. These data and operator actions were utilized to develop inputs to the Mark-B-HTP LOCA analysis as documented by DIN 3. Any changes to the data and/or operator actions listed in DIN 2 must be evaluated with respect the impact on the inputs to the LOCA analyses that are documented in DIN 3.IMPACT ON OUTPUT DOCUMENTS:
There have been no changes in computed data and conclusions.
Therefore, there is no impact on output documents (i.e., System Descriptions, USAR or fuel design information).
Page v CALCULATION NOP-CC-3002-01 Rev. 01 INITIATING DOCUMENT (S)NIA CALCULATION NO.C-NSA-064.02-036, R01[ X I VENDOR CALC
 
==SUMMARY==
VENDOR CALCULATION No. 86-5006232-03 TITLE/
 
==SUBJECT:==
 
DB-1 LOCA Summary Report DOCUMENT INDEX z C: Z Document Number/Title Revision, S-Edition, Of Date 1 Title 10, Code of Federal Regulations, Part 50.46, "Acceptance Criteria for 11-3-97 9 0 El Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors." 2 C-NSA-064.02-035, "Design Inputs for Accident Analyses." ROO 0 El 13 3 51-5053743-01, "DB-1 LOCA AIS for Mk-B-HTP Fuel," AREVA RO0 0 El E](ACT 06-0007).
._
20897-10 (3/3/6)A CALCULATION
 
==SUMMARY==
SHEET (CSS)AR EVA Document Identifier 88-5006232-03 Title DB-1 LOCA Summary Report PREPARED BY: REVIEWED BY: METHOD: DETAILED CHECK 0 INDEPENDENT CALCULATION NAME NAME SIGNATU A---- SIGNATURE TI.TE Supervisor DATE TITLE Engineer Ill DATE tI/o/6 COST REF. 1TM STATEMENT-E_CENTER 41306 PAGE(S) 204-208 REVIEWER IN DEPENNDENCE NAME; URPOSE AND
 
==SUMMARY==
OF RESULTS: Revised to address corrections in the SBLOCA analyses.FirstEnergy Corporation operates the B&W-designed plant Davis-Besse Unit 1 (DB-1). FirstEnergy Corporation has scheduled to transition DB-1 to AREVA NP Inc. (AREVA) Mark-B-HTP fuel. As part of this effort, AREVA has performed new loss of coolant accident (LOCA) linear heat rate (LHR) limit analyses for DB-l with the Mark-B-HTP fuel. These analyses consider a full core of Mark-B-HTP fuel and a mixed core of Mark-B-HTP, Mark-B12, Mark-B10K, and Mark-BIOM fuel to address the transition to the new fuel design. The purpose of this document is to summarize the results of these analyses and demonstrate conformance to the five criteria of 10 CFR 50.46.The DB-1 plant has been shown to be in compliance with the five criteria of 10 CFR 50.46 for both the large and small break loss-of-coolant accident (LOCA) analyses.
Compliance with the first three criteria of 10 CFR 50.46 has been demonstrated based on analyses with the LOCA evaluation model (EM) described in BAW-10192P-A (Reference 1). Compliance with the remaining two criteria of 10 CFR 50.46 have been demonstrated through a combination of evaluations, analyses, monitoring and testing. The analyses considered U0 2 fuel as well as gadolinia fuel in a variety of concentrations.
The LOCA analyses were performed at a core power level of 102, percent of 2966 MWt, or 3025 MWt, to cover a potential future power uprate. With the CALDON LEFM CheckPlusTM system in operation, the core power level is 100.37 percent of 3014 MWt, or 3025 MWt. Additional analyses were performed to determine LHR limits applicable to a power level of 102 percent of 2772 MWt, or 100.37 percent of 2817 MWt with the CALDON LEFM CheckPlus' T M system in operation.
THE DOCUMENT CONTAINS ASSUMPTIONS THAT MUST BE VERIFIED PRIOR TO USE ON THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:
SAFETY-RELATED WORK CODENERSION/REV COOENERSIONREV YES None Z NO I AREVA NP Inc.. an AREVA and S/emmns company Page*_1. of 208 A AREVA 86-M,6232-03 RItcrd of Revision Revision Date Pages/Sections Changed Description/Change Authorization 00 April 2000 Original Release.Revised to incorporate new LOCA 01 September Marked with change ars limits for Mark-B12, Mark-B12 2002 MGad, PSC 2-00 resolution, reduced HPI flow, and peak local oxidation.
02 January 2006 Replace entire document.
RVised to add Mark-B-HTP LOCA... results.Revised to address corrections in the Complete revision.
SBLOCA analysis.Technical changes Framatome ANP, Inc, updated to 03 April 2006 marked with change bars. AREVA NP Inc.Company name change FANP a t AREVA.not marked. u Typos corrected.
I 2 A AR EVA 86-&)6232-03 Table of Contents I Introduction and Purpose ...........................................................
j ..... .. ........ 12 2 Key Assumptions
.......................................
13 3 Summary of Results .........................................................................................
13 3.1 Adherence to 10 CFR 50.46 Criteria ...............................................................
13 3.1.1 Peak Cladding Temperature
.............................................................................
13 31.2 Local Cladding Oxidation
..............................................................................
14 3.1.3 Whole-Core Oxidation and Hydrogen Generation
...........................................
14 3.1.4 Coolable Core Geometry ................................................................................
15 3.1.5 Long-Term Core Cooling .................................................................................
16 3.2 Summary of LBLOCA Results ......................................................................
18 3.3 Summary of SBLOCA Results .......................................................................
21 4 Analytical Methodology
.................................................................................
56 4.1 LBLOCA Analyses .........................................................................................
56 4.2 SBLOCA Analyses .......................................................................................
57 5 Plant Parameters and Inputs ..........................................
o ..................................
59 6 LBLOCA Sensitivity Studies and Analyses .....................................................
74 6.1 LBLOCA Sensitivity Studies ..........................................................................
74 6.1.1 LBLOCA Evaluation Model Generic Sensitivity Studies ...............................
74 6.1.2 LBLOCA Evaluation Model Plant-Specific Sensitivity Studies ...........................
77 6.1.3 DB-1 Plant-Specific LBLOCA Sensitivity Studies ........................................
81 6.1.4 LBLOCA Topics ...............................................................................................
85 6.2 LBLOCA Analyses .........
.............................
87 6.2.1 LBLOCA Base Model .........................
...........................................................
87 6.2.2 Transient Progression
......................
................................................................
88 6.2.3 Sequence and Cases for Analysis ............................................
..............................
89 6.2.4 Mark-B-HTP LOCA LHR Limits .............................
90 6.2.5 Mark-B 10K LOCA LHR Limits .....................................................................
93 6.2.6 Mark-B 12 LOCA LHR Limits ..........
I .....................
97 6.2.7 Mark-BIOM .LOCA LHR Limits ......................................................................
98 7 SBLOCA Sensitivity Studies and Analyses ........................................................
142 7.1 SBLOCA Sensitivity Studies ...............................................................................
142 7.1.1 SBLOCA Evaluation Model Generic Studies ......................................................
142 7.1.2 SBLOCA Evaluation Model Plant-Specific Sensitivity Studies ........................
144 7.1.3 SBLOCA DB-1 Plant-Specific Sensitivity Studies .............................................
147 7.1.4 SBLOCA Topics ..............................................................................................
150 7.2 SBLOCA Analyses ...... ......................................................................................
152 7.2.1 SBLOCA Base Model ........................................................................................
152 7.2.2 SBLOCA Transient Progression
.............................
154 7.2.3 Interdependencies of ECCS and AFW Used in SBLOCA Mitigation for B&W Plants ........................................................................................................
156 7.2.4 Break Spectrum Analyses ...................................................................................
162 8 RELAPS/MOD2-B&W EM SER Restrictions
...................................................
199 9 References
...........................................................................................................
204 3 A ARE VA 86-5006232-03 5 Plant Parameters and Inputs This section provides a summary of the specific plant parameters and inputs used in the LBLOCA and SBLOCA analyses.
The plant parameters and inputs for the Mark-B-HTP analyses are discussed in detail in Reference 33 and listed in Table 5-1 through Table 5-14. The inputs were developed based on the customer-supplied plant-specific data documented in Reference
: 32. Customer approval of the Reference 33 plant parameters and methods used for the LOCA analyses is contained in Reference
: 34. Select plant parameters and boundary conditions are summarized in this section.The original Mark-B10K and Mark-B12 LBLOCA and SBLOCA were based on a slightly different set of input parameters.
The Mark-B10K and Mark-B12 LBLOCA analyses were adjusted to the new plant parameters, with the exception of TAVE, in Reference 33 via the evaluations performed in Reference
: 47. A reduced TAVE of 580 F was selected for the Mark-B-HTP analyses as a conservatism to envelope a potential future power uprate that includes a reduction in TAVE. The SBLOCA spectrum for the Mark-B-HTP has been determined to be applicable to all fuel assembly types at DB- 1. Therefore, the plant parameters listed in Table 5-1 through Table 5-14 are applicable to the LOCA analyses for the Mark-B-HTP, Mark-B12, Mark-B 10K, and Mark-BIOM fuel assemblies.
59 A A t'. 8OL9fi"232-ft-A Table 5-1: LOCA Plant Parameters and Inputs for DB-I [Reference 33]Parameter LBLOCA SBLOCA RCS Conditions Core Power (MWt) 3,025 (Note 1) 3,025 (Note 1)RCP Energy Contribution to RCS (MWt) 16 (Note 2)Decay Heat Standard 1.2 ANS 1971 Actinide Coefficients B&W Heavy Actinides Primary Side TAvE (F) 580 Mark-B-HTP, 582 others (Note 5)RCS Pressure @ HL Tap (psia) 2170 Total RCS Mass Flow Rate (gpm) 380,000 Core Bypass Flow (%) 7.5 Pressurizer Parameters Indicated PZR Level (in) 220 (on 320 in scale)Pressurizer Heaters & Sprays Not Modeled PORV Not Modeled PSV Not Modeled MFW System Parameters and SG Tube Plugging MFW Temperature (F) 464 MFW Flow Rate per SG (Ibm/s) 1792 Ibmlsec (Note 3)MFW Isolation LOOP Reactor Trip Main Feedwater Coastdown 2 sec delay, 12.5 sec coastdown (Note 11)SG Tube Plugging 25% BL/15% IL 175% in AFW wetted region Turbine and Main Steam System Parameters Turbine Header Pressure (psia) 885 (Note 4)Turbine Trip Criteria on Reactor Trip Turbine Trip Delay Time (s) 0.5 Turbine Stop Valve Stroke Time (s) 0.5 MSIV Trip Setpoint (psia) 606.3 MSIV Trip Delay (s) 1.0 MSIV Stroke Time (s) 6.0 MSSV Capacity & Setpoints Table 5-8 AVV Capacity (lbm/hr) Not Modeled AFW Parameters and SG Level Control AFW Temperature (F) Not Modeled 120 AFW Delay Time (s) Not Modeled 120 Table 5-1 continued on next page.60 A A ~C'.IA Table 5-1 (continued):
LOCA Plant Parameters and Inauts for DB-1 [Reference 331 Parameter LBLOCA SBLOCA AFW Parameters and SG Level Control (continued)
AFW Actuation Not Modeled SFRCS (loss of RCPs)AFW Flow per SG (gpm) Not Modeled Table 5-9 AFW Steam Demand (lb/br) Not Modeled Table 5-10 Fill to 2.2 ft on SFRCS SFRCS SG Level Control (ft) Not Modeled' signal plus AFW delay time SFAS SO Level Control (ft) Not Modeled Fill to 11.1 ftonSFAS_______________________signal RPS and SFAS Parameters Low RCS Pressure Reactor Trip (psia) 1885 (Note 10)Reactor Trip Delay (s) Note 10 0.6 SFAS Low RCS Pressure Trip (psia) i515 HPIDelay Time after SPAS Low Pressure Trip (s) 40 SFAS Low-Low RCS Pressure Trip (psia) 384.4 LPI Delay Time after SFAS Low-LoW RCS Note 9 Pressure Trip (s)LSCM (F) 20 LoopTurbine Trip (modeled'as LOOP Brek OpningReactor Trip + 0.6 sec)ECCS Parameters BWST Liquid Temperature (F) 90 BWST Minimum Volume (ft3) .360,000 CFT Liquid Volume (fl 3/tank) 1000-1080 CFT Liquid Temperature (F) 120 CFT Gas Pressure (psia) 582-648 Nominal CFT Surge-line losses (form & friction) 7700 CLPD, F IT 7 CFT Area of Surge Line (f2) 0.7213 Average Length of Surge Lines (ft) 70.1 Average Elevation Change of CFT Surge Line (ft) 0.4479 (Note 7)Mark-B-HTP:
Not credited for core cooling HPI Flow Mark-B OK, Mark-B 12, Table 5-2, Table 5-3, Mark-B IOM: Credited for Table 5-4 core cooling. See Section 6.1.2.3.LPI Flow 2*Table 5-5 l*Table 5-5 Table 5-1 continued on next page.61 Page i CALCULATION NOP-CC-3002-01 Rev. 01 INITIATING DOCUMENT (S) CALCULATION NO, [ VENDOR CALC
 
==SUMMARY==
C-ICE-051.01-002, Revision 0 VENDOR CALCULATION NO.TITLE/
 
==SUBJECT:==
 
Core Flood Tank Pressure[I BVI I EIBV2 0DB '[EPY Category ' Active C Historical I- Study, Classification 0 Tier 1 Calculation 0 Safety-Related/Augmented Quality E] Nonsafety-Related Open Assumptions?
0 Yes 0 No If Yes, Enter CR Tracking Number System Number: 051-01 DB-PTCF4AI, DB-PTCF4A2, DB-PTCF4BI, DE'-PTCF4B2, DB-PTCF4Ai (IBI), DB-PTCF4A2 (lB 1), DB-PTCF4BI (1B2), DB-PTCF4B2 (1B2), DB-PTCF4AI
(-K+E), DB-PTCF4A2
(-K+E), DB-PTCF4BI Asset Number: (-K+E), DB-PTCF4B2
(-K+E), DB-PTCF4A2(EB I), DB-PTCF4A2(EBI), DB-PTCF4BI(EB3), DB-PTCF4B2(EB3), DB-PICF4AI, DB-PICF4A2, DB-PICF4BI, PICF4B2, DB-PSCF4A 1, DB-PSCF4A2,....____ _ DB-PSCF4BI, DB-PSCF4B2, P079, P080, P089, P090 Commitments:
None (Perry Only) Calculation Type: Referenced In Atlas? [I Yes C] No Referenced In UISAR Validation Database 0 ]Yes [] No Computer Program(S)
Program Name Version f Revision Category Status Description Microsoft Word 97 C Active Word Processor Revision Record Rev. Affected Pages Originator/Date Reviewer/Design Verifier/Date Approver/Date Descpion of Change: Initial re sion establishes thdeinstrument und rtainty for Core Floods Describe where the calculation has been evaluated for 10CFR50.59 applicability.
05-00482 Rev. Affected Pages Originator/Date Design Verifier/Date Approver/Date Description of Change: Describe where the calculation has been evaluated for 10CFR50.59 applicability.
Rev. Affected Pages Originator/Date Design Verifier/Date Approver/Date Description of Change: Describe where the calculation has been evaluated for 1OCFR50.59 applicability.
Page ii CALCULATION
.... NOP-CC-3002-01 Rev. 01 INITIATING DOCUMENT (S) CALCULATION NO. ] VENDOR CALC
 
==SUMMARY==
C-ICE-051.01-002, Revision 0 VENDOR CALCULATION No.TITLE/
 
==SUBJECT:==
 
Core Flood Tank Pressure TABLE OF CONTENTS SUBJECT PAGE COVERSHEET:
OBJECTIVE OR PURPOSE iii SCOPE OF CALCULATION iii
 
==SUMMARY==
OF RESULTS/CONCLUSIONS ii LIMITATIONS OR RESTRICTION ON CALCULATION APPLICABILITY ii IMPACT ON OUTPUT DOCUMENTS lii DOCUMENT INDEX Iv CALCULATION COMPUTATION (BODY OF CALCULATION):
ANALYSIS METHODOLOGY 1 ASSUMPTIONS 4 ACCEPTANCE CRITERIA 6 COMPUTATION 7 RESULTS 26 CONCLUSIONS 28 ATTACHMENTS:
Attachment I -Copy of applicable page of Framatome 88-5006232-01 (DIN 59% 1 Page TOTAL NUMBER OF PAGES IN CALCULATION (COVERSHEETS
+ BODY + ATTACHMENTS) 35 Pages SUPPORTING DOCUMENTS (For Records Copy Only) .DESIGN VERIFICATION RECORD 1 Pages CALCULATION REVIEW CHECKLIST 3 Pages 10CFR50.59 DOCUMENTATION 4 Pages DESIGN INTERFACE
 
==SUMMARY==
1 Pages DESIGN INTERFACE EVALUATIONS 11 Pages OTHER -Comment sheet from Barteck 1 Pages 0] YES EXTERNAL MEDIA? (MICROFICHE.
ETC.) (IF YES, PROVIDE LIST IN BODY OF CALCULATION)
[ NO 1 0 NO
...Page iii jCALCULATION NOP-CC-3002-01 Rev. 01 INITIATING DOCUMENT (S) CALCULATION NO. I] VENDOR CALC
 
==SUMMARY==
C-ICE-051.01-002, Revision 0 VENDOR CALCULATION No.TITLEI
 
==SUBJECT:==
 
Core Flood Tank Pressure OBJECTIVE OR PURPOSE: This calculation establishes the Instrument uncertainty associated with the measurement of Core Flood Tank Pressure using indicators PICF4A1, PICF4A2, PICF4B1, PICF4B2 or computer points P079, P080, P089, P090.This calculation also validates the pressure switch setpoints for annunciation of pressure being either too high or too low.SCOPE OF CALCULATION/REVISION:
Initial revision.
This calculation only addresses Instrument uncertainties associated with normal plant environmental conditions
-those present during the surveillance testing.
 
==SUMMARY==
OF RESULTS/CONCLUSIONS:
This calculation determined the instrument uncertainty associated with the measurement of Core Flood Tank Pressure indication and pressure switches.The error for CFT Pressure Indicators PICF4A1 (A2), PICF4BI(B2) is: +/-1.93% span or +/- 13.51 psig. To protect the analytical value, the surveillance must be between 580.51 pslg and 619.49 psig. Due to readability of the indicators, the indicated pressure must be between 590 and 610 psig.The error for computer points P079, P080, P089, P090 is: +/-1.39% span or +/-9.73 psig. To protect the analytical value, the surveillance must be between 576.73 psig and 623.27 psig. With additional margin added, the surveillance should be established between 580 psig and 620 psig.The error for CFT Pressure Switches DB-PSCF4A1 (A2), DB-PSCF4B1 (B2) is: +1.34% span or +9.38 psig.The Analytical Limit is 615 psia + 33 psi (600 psig +/- 33 psi) (DIN 59) and the Technical Specifications (DIN 2)require a nitrogen cover-pressure of between 575 and 625 psig. This provides 8 psi of instrument uncertainty between the Analytical Limit and the Technical Specification values. The instrument uncertainty calculated is larger than the 8 psi assumed in the current analysis.
However, with alarm setpoints of 585 psig decreasing and 615 psig increasing and an instrument uncertainty of only + 9.38 psig, the alarms will protect both the Technical Specification and the Analytical Limit values.LIMITATIONS OR RESTRICTIONS ON CALCULATION APPLICABILITY:
Conclusions are applicable only to surveillance testing and normal operating conditions.
IMPACT ON OUTPUT DOCUMENTS:
The results of this uncertainty evaluation should be used to adjust the acceptance criteria for the Core Flood Tank Pressure measurements in procedure DB-OP-03006 (DIN 47) (See CR 05-00381 -DIN 61). It is recommended that the computer points be used in lieu of the indicators due to a smaller uncertainty associated with those indications.
The pressure switches provide adequate protection to ensure the operator is alerted prior to reaching the Technical Specification value. In addition, the analytical limits are protected even further based on instrument uncertainty already being included in establishing the Technical Specificationvalues.
Page iv CALCULATION NOP-CC-3002-01 Rev. 01 INITIATING DOCUMENT (S) CALCULATION NO. [)VENDOR CALC
 
==SUMMARY==
C-ICE-0511.O1
-002. Revision 0 VENDOR CALCULATION No.TITLE/
 
==SUBJECT:==
 
Core Flood' Tank Pressure DOCUMENT INDEX 6 0 -z Document Numberf'itle Revision, Edition, ý( -0.Date .: 1. Updated Safety Analysis Report Rev. 24 0 El 6/04 2. Technical Specifications 4.5.1 .a. 1, Core Flooding Tanks Includes [ [] [--"__Amendment 207 3. DBI-100, Davis-Besse Nuclear Station Unit 1 Environmental Rev. 10 0 1] 1]Qualification of Electrical Equipment 4. M-034, "Piping & Instrumentation Diagram, Emerg. Core Cooling Rev. 56, [1 []System Ctmt. Spray & Core Flooding Systems 5. OS-0006, Operational Schematic, Core Flooding System Rev. 16 0 E2 [E]6. M-530-329, Core Flooding control Loop CF3, C4 Schematic (Sheet I Rev. 6 f] l fl___of I) B&W Nuclear Steam Supply System 7. ANSI/ISA-67.04.01-2000, Setpoints for Nuclear Safety-Related Approved I January .] l [Instrumentation 2000 8. ISA-RP67.04.02-2000, Methodologies for the Determination of Approved 1 January [ []Setpoints for Nuclear Safety-Related Instrumentation, Approved 1 2000 January 2000 9. ANSIIISA-S51.1, Process Instrumentation Terminology 1979" [] [j 10. SAMA Standard PMC 20.1, Process Measurement and Control 1973 11 0.Terminology R1 Reg Guide 1.105, Setpoints for Safety-Related Instrumentation Rev. 3 0 [1 12. SAP N/A 10 El'13. SD-040, System Description for Core Flooding Rev. 3 [ [] []14. SD-05 1, System Description for Non-Nuclear Instrumentation Rev. I [ [] []15. Vendor Manual, M-530-354, B&W Integrated Control and Non- Rev. 4 E] z n Nuclear Instrumentation, Data Sheet 34162-1, Westinghouse Veritrak Products Model 56PM Series, Gauge Pressure Transmitter, Medium__ Range, Issued 11-71 16. Vendor Manual M-536-118, B&W Module Instructions Book for NNI Rev. 4 El 0 0 and JCS, Volume 5-Bailey Controls Company Product Instruction E92-79, Buffer module, Pt. No. 6624610P Bailey Control Systems Product Instruction E92-60-3, Summer Plus Bias Action Unit, Pt. No. 6623695--Bailey Controls Company Product Instruction E92-74, Signal Monitor module, Pt. No. 6623819-1, 17. Vendor Manual, M-530-353, Integrated Control Systems and Non- Rev. 6. [ LI Nuclear Instrumentation, Vol. 1 El Page v CALCULATION NOP-CC-3002-01 Rev. 01 INITIATING DOCUMENT (S) CALCULATION NO. VENDOR CALC
 
==SUMMARY==
C-ICE-051.01-002, Revision 0 VENDOR CALCULATION No.TITLEI
 
==SUBJECT:==
 
Core Flood Tank Pressure 6 -, Z c " Document Number/Title Revision, Edition, 0 -Date 2 Bailey Product Instruction El 2-9, Edgewise Indicator Type RY " 18. Vendor Manual G-CS-406-2, MODCOMP Vol. XIV, "Process Rev. 2 Li [ LI Input/Output Subsystem" 19. String Work Package 5 IA-ISPCF4AI Rev. 2 Li [ L]20. Instrument Information Sheet PT-CF4AI Rev. I -] El .21. Instrument Information Sheet PT-CF4Al (I131) Rev. I .[Ell 22. Instrument Information Sheet PT-CF4A I (-K+E) Rev, I El 23. Instrument Information Sheet PT-CF4A I (EB 1) Rev. I [] E) , 24. Instrument Information Sheet PS-CF4AI Rev. 2 Li 0 I]25. Instrument Information Sheet PI-CF4AI Rev. 2 [] E[]26. String Work Package 5 1A-ISPCF4A2 Rev. 1 El 0 El 27. Instrument Information Sheet PT-CF4A2 Rev. 0 [] [ []28. Instrument Information Sheet PT-CF4A2 (IB1) Rev'. 0 [] [ I]29. Instrument Information Sheet PT-CF4A2 (-K+E) Rev. 0 Li ' []30. Instrument Information Sheet PT-CF4A2 (EBI) Rev. 0 EL [ I]31. Instrument Information Sheet PS-CF4A2 -Rev. 2 Li , Li 32. Instrument Information Sheet PI-CF4A2 .Rev. 2 Li E Li 33. String Work Package 51A-ISPCF4BI Rev. 2 Li [ LI 34. Insmmment Information Sheet PT-CF4B I Rev.1 E Li E L 35. Instrument Information Sheet PT-CF4BI (1B2) Rev. 1 Li .E ]36. Instrument Information Sheet PT-CF4B I (-K+E) Rev. I [] 0 []37. Instrument Information Sheet PT-CF4B I (EB3) Rev. I Li [ F]38. Instrument Information Sheet PS-CF4BI Rev. 4 El 0 EL 39. Instrument Information Sheet Pl-CF4BI Rev. 4 El 0 El 40. String Work Package 5 1A-ISPCF4B2 Rev. I [] [ El 41. Instrument Information Sheet PT-CF4B2 Rev. 0 Li] []42. Instrument Information Sheet PT-CF4B2 (1B2) Rev. 0 El[] [43. Instrument Information Sheet PT-CF4B2 (-K+E) Rev. 0 E' 0 []44. Instrument Information Sheet PT-CF4B2 (EB3) Rev. 0 C] .LI 45. Instrument Information Sheet PS-CF4B2 Rev.] 0 []46. Instrument Information Sheet PI-CF4B2 Rev. 1 F] [ Li Page vi FtEnen: ,CALCULATION NOP-CC-3002-01 Rev. 01 INITIATING DOCUMENT (S) CALCULATION NO. [] VENDOR CALC
 
==SUMMARY==
C-ICE-051.01-002, Revision 0 VENDOR CALCULATION No.TITLE/
 
==SUBJECT:==
 
Core Flood Tank Pressure ICl)6,)0- "5 z 0" , Document Number/Title Revision, Edition, 0 O Date 47. Surveillance Test Procedure DB-OP-03006, Miscellaneous Instrument Rev. 17 j] [] [Shift Check 48. Procedure DB-OP-02003, ECCS Alarm Panel 3 Annunciators Rev. 06 U] El 0 49. Procedure DB-OP-06014, Core Flooding System Procedure Rev. 09. U-" 0 0 50. Instrumentation and Control Procedure DB-MI-04250, String Check of Rev. 1 0 El []51A-ISPCF4A I Core Flooding Tank 2 Pressure 51. Instrumentation and Control Procedure DB-MI-0425 1, String check of Rev. I a 0 El 5 1 A-ISPCF4A2 Core Flooding Tank 2 Pressure 52. Instrumentation and Control Procedure DB-MI-04252, String Check of Rev. 1 0 " : U U 51 A-ISPCF4B I Core Flooding Tank 1 Pressure 53. Instrumentation and Control Procedure DB-MI-04253, String Check of Rev. 1 0 U U 5 1 A-ISPCF4B2 Core Flooding Tank 1 Pressure 54. DB-MI-05026, Motorola/Westinghouse Veritrak Model 56P Series Rev. 1 0 El U Gauge Pressure Transmitter Calibration
: 55. DB-MI-05040, Bailey 820 Buffer Module Calibration Rev. 0 N E U 56. DB-MI-05007, Bailey Type RY Edgewise Indicator Calibration Rev. 0 " D]57. NOP-WM-5001, Measuring and Test Equipment Calibration .Rev. n E U 58. NOP-CC-3002, Calculations Rev. 1 0 E] El 59. Framatome Calculation 86-5006232-01, LOCA Analysis Rev. 1 0] Z U 60. Condition Report 02-05959, SHRR: EVALUATION OF INDUSTRY N/A 19 Ul U1 EVENTS NEED FURTHER INVESTIGATION
: 61. Condition Report 05-00381, INSUFFICIENT INSTRUMENT N/A ' Ul Ul UNCERTAINTY IN PROCEDURE FOR CFT TS COMPLIANCE I
Page I fs CALCULATION COMPUTATION
!NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE I
 
==SUBJECT:==
Core Flood Tank Pressure 1.0 ANALYSIS METHODOLOGY This calculation determines the instrument uncertainty for the measurement of Core Flooding Tank Pressure.
The results of this calculation may be used when obtaining measurements for the surveillance test procedure that provide verification of adequate pressure in the Core Flood Tank. This calculation determines instrument uncertainties associated with pressure indications under plant normal environmental conditions, representative of those experienced during the surveillance testing.This calculation also calculates the uncertainty associated with the pressure switches that actuate annunciators 3-4-F, 3-4-G, 3-2-F, and 3-2-G, for Core Flood Tank High and Low pressures to validate the setpoints will alert the Operator prior to violating a Technical Specification.
These also are evaluated under normal plant environmental conditions.
ANSI/ISA-67.04.01-2000 (DIN 7) develops a basis for establishing setpoints for nuclear safety-related instrumentation.
This document was prepared by the Instrument Society of America (ISA) with a goal of providing uniformity in the field of instrumentation.
ISA-RP67.04.02-2000 (DIN 8) presents guidelines and examples of methods for the implementation of ISA-67.04.01-2000.
Regulatory Guide 1.105 (RG 1.105, DIN 11) endorses the use of the ISA standard as an acceptable method for determining safety-related setpoints.
While RG 1.105 is specifically applicable to safety-related setpoints, it also recognizes that the standard "is also appropriate for non-safety system instrumentation for maintaining design limits described in the Technical Specifications".
1.1 Affected Instrument Strings Core Flooding Tank Pressure Pressure Transmitter:
DB-PTCF4Al, DB-PTCF4A2, DB-PTCF4B1, DB-PTCF4B2 (Motorola-Westinghouse 56PM142K-0030)
Current Buffer: DB-PTCF4A 1 (B1 1), DB-PTCF4A2 (I31)(Bailey 6624610-2222)
Current Buffer: DB-PTCF4B1 (IB2), DB-PTCF4B2 (IB2)(Bailey 6628999-1)
Summer+Bias+Inverter:
DB-PTCF4AI
(-K+E), DB-PTCF4A2
(-K+E), DB-PTCF4BI
(-K+E), DB-PTCF4B2
(-K+E)(Bailey 6623695-2)
Page 2* /CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE I
 
==SUBJECT:==
Core Flood Tank Pressure Voltage Buffer: Pressure Indicator:
Pressure Switch: Computer Point: DB-PTCF4AI (EB I), DB-PTCF4A2(EBI), DB-PTCF4B 1 (EB3), DB-PTCF4B2(EB3)(Bailey 6624609-1)
DB-PICF4A 1, DB-PICF4A2, DB-PICF4B 1, DB-PICF4B2 (Bailey RY-21 IX)DB-PSCF4AI, DB-PSCF4A2, DB-PSCF4BI, DB-PSCF4B2 (Bailey 6623819-1)
P079, P080, P089, P090 Page3~CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE I
 
==SUBJECT:==
Core Flood Tank Pressure 1.2 Functional Description/Design Basis (DIN 1, 13, 14, 19, 26, 33, 40)Pressure transmitters DB-PTCF4A 1 (A2), DB-PTCF4B 1 (132) sense Core Flood Tank pressure and provide indication to the Control Room (C5716), signals to computer points, and high and low alarms in the Control Room. The alarms indicate leaks into and out of the Core Flood Tanks.Technical Specifications Technical Specification 3.5.1 .d requires that each reactor coolant system core flooding tank be operable with a nitrogen cover-pressure of between 575 and 625 psig. The surveillance requirement 4.5.1 .a. 1 is to verify the contained borated water volume and nitrogen cover-pressure in the tanks at least once per 12 hours. Surveillance Test Procedure DB-OP-03006 (DIN 47) satisfies the surveillance by verifying the core CFT pressure using indicators DB-PICF4B 1, DB-PICF4B2, DB-PICF4A2, and DB-PICF4A 1.If the CFT pressure indicators are not available, computer points P079, P080, P089, and P090 may be used as alternatives.
Proper Core Flooding Tank Pressure The proper Core Flooding Tank pressure is 600 +/-25 psig, which allows for variations during operation.
The LOCA analyses assumed the Core Flooding Tank pressure was initially at 567 psig (DIN 59).
Page 4 Ekst ~CALCULATION COMPUTATION NOP-CC,3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 2.0 ASSUMPTIONS 2.1 Reading Error ISA RP67.04.02 (DIN 8) does not address uncertainties attributable to the effects of Readability.
The Reading Error is assumed to be bounded by Y/ of a minor division on the scale, typical of industry practice when determining this error term.2.2 Drift Effects Drift will be assumed to be equivalent to accuracy in cases where instruments do not have specified effects and in lieu of performance information.
2.3 Temperature Effects Temperature effects that are not specified by the Vendor are assumed to be included in the Vendor provided accuracy if the instrument is operating within the referenced temperature range.2.4 Containment Temperature Calibration temperature and operating temperature is assumed to be 60-100 degrees F for the Core Flooding Tank pressure transmitters.
This is a reasonable assumption based on the fact that the transmitters are located on the 565-foot level in the containment building.
This area is generally cooler than the average containment temperature because the Containment Air Coolers discharge to the lower elevations.
This airflow arrangement tends to moderate any extreme temperatures in the lower containment elevations.
Using the worst case 55-110 degrees F (DIN 3) range is an overly conservative assumption that is not appropriate for these transmitters.
The lower value of 60 degrees F is based on calibration temperature.
Per CR 02-05959 (DIN 60), a review of the temperatures in containment revealed a worst case of 60 degrees during plant shutdown, the period during which the transmitter would be calibrated.
Page 5 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 2.5 Power Supply Voltage Variation The variation in power supply voltage for the pressure transmitters will be assumed to be +/-10% or 2.4 volts for the 24 Vdc power supply. A 10% variation is considered very conservative.
2.6 Definition of Accuracy Accuracy is defined in SAMA PMC 20.1 (DIN 10). It is also defined in ANSI/ISA S51.1 (DIN 9), which uses the SAMA PMC 20.1 as the reference for the definition.
In the SAMA Standard, accuracy is defined as "reference accuracy" for performance specifications unless otherwise noted.Reference accuracy is defined in SAMA PMC 20.1 as including the combined conformity, hysteresis, and repeatability errors. Reference accuracy is defined in ANSI/ISA S5 1.1 as equivalent to "accuracy rating." Accuracy rating is defined as including the combined effects of conformity, hysteresis, dead band, and repeatability.
Therefore, it is assumed that if accuracy is stated in addition to conformity, hysteresis, dead band and/or repeatability, the only error that must be included in the final calculated uncertainty is accuracy.It also should be noted that linearity, per SAMA PMC 20.1 and ANSI/ISA S51.1, is identical to conformity with the exception that linearity is only for straight lines, where conformity would allow the line to also be a curved line. Therefore, conformity includes and bounds linearity.
2.7 Control Room Normal Temperature The control room temperature is defined as being controlled at 75 degrees F and 50 percent relative humidity in the summer and 75 degrees F and 30 percent relative humidity in the winter per DIN I, Section 9.4.1.1. Also in DIN 1, Table 7.2-3 defines the range as 60 -.80 degrees F. It is assumed that the calibration is done at 75 degrees F since that is the normal control room temperature.
Based on this, the largest variation would be between a normal calibration at 75 degrees F and the low of 60 degrees F or 15 degrees F. It is considered overly conservative to assume the calibration was accomplished at either the high or low temperature and then operated at the other extreme since the control room temperature does not normally vary by more than a few degrees.
Page 6 RCALCULATION COMPUTATION NOP-CC-3002-.01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 3.0 ACCEPTANCE CRITERIA The acceptance criteria for the calculation of measurement uncertainties for Core Flood Tank Pressure is that the instruments will operate with margin to the Analytical Limits and that work is done in accordance with the methodology of ISA RP67.04.02-2000, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation.
There are no other specific numerical criteria associated with this calculation.
I Page 7 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 TITLE I
 
==SUBJECT:==
Core Flood Tank Pressure 4.0 COMPUTATION 4.1 Loop Diagram For CFT Pressure Indication The instrument loop shown on Figure 1 depicts the CFT Pressure Indication addressed in this calculation.
Channel CF4B 1 is indicative of the other channels.
Refer to P&ID for layout (DIN 4)FIGURE 1 Motorola-Westinghouse 56PM 142K-0030 0-700 psig Bailey 6628999-1 Bailey Voltage Buffer 6624609-1 U.0 Bailey Summer+Dias+Inverter 6623695-2 Bailey Voltage Buffer 6624609-I High Pressure Annunciator Low Pressure Annunciator Bailey Pressure Indicator RY-21 IX Output Range: 0-700 psig Page 8 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 4.2 Instrument Calibration Uncertainties The calibration standard will not be included in the calculation.
There is at least an 8:1 accuracy ratio between the calibration standard and the installed instrument, based on controls in procedure NOP-WM-5001 (DIN 57). This calibration standard accuracy component, when included as a component in the Square Root Sum of the Squares, results in a final calculated value that is less than 1% larger than the accuracy of the installed instrument.
This is two orders of magnitude less, which is considered negligible.
The surveillance test procedures in conjunction with the IC Data Packages specify the calibration equipment to be utilized.
The error components are noted below.4.3 Process Measurement Effects for CFT Pressure Indication None Page 9 CALCULATION COMPUTATION
.__. NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 V0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 4.4 Pressure Transmitter (DIN 15, 20, 27, 34, 41)Component ID: Manufacturer:
Model: Input Range: Output Range: Range Limits: Accuracy: Deadband: Repeatability:
Ambient Temperature Effect Zero Error: 100% Error: (Zero plus Span)Relative Humidity: Indicating Meter Accuracy: Frequency Response: Power Supply Voltage Effect: DB-PTCF4A I, DB-PTCF4A2, DB-PTCF4B I, DB-PTCF4B2 Motorola-Westinghouse 56PM 142K-0030 0-700 psig 4-20 mAdc (1-5 Vdc across a 250 ohm resistor)0-200/800 psig (continuously adjustable)
+/-0.5% of calibrated span (includes Linearity, Hysteresis) not more than 0.01% of calibrated span+0.1% of calibrated span+/-1% of range / 100IF from 0 to 200'F+/-1% of range / 50°F from 0 to 200'F transmitter is operable up to 95%+/-2%flat to 20 Hz less than 0.1 % change for I volt power supply change Page 10 , CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 4.4.1 Pressure Transmitter Accuracy Effect Vendor Accuracy Effect is specified as 0.5% of calibrated span including Linearity and Hysteresis.
The vendor has also specified a Repeatability and a Dead Band value. Per Assumption 2.6, the repeatability and dead band errors are included with accuracy and will be ignored.Accuracy Effect +/- (Vendor Acec)0.5% span 4.4.2 Pressure Transmitter Drift Effect Drift is considered equivalent to the reference accuracy of +/-0.5% span in lieu of any performance data.See Assumption 2.2.4.4.3 Pressure Transmitter Temperature Effect Temperature Effect is +/-1-% of range / 100'F for zero range error and +/-1% of range / 50'F for 100%error (Zero plus Span). The effect will be based on the 100% error and an assumed temperature variation of 40'F. See Assumption 2.4 Temperature Effect =+/- 1% of range / 50'F= + 1% (800 psig/700 psig)
* 40*F/ 501F=+0.91% span 4.4.4 Pressure Transmitter Power Supply Effects Power Supply Effects are less than 0.1 % change for a 1 volt power supply change. The variation in power supply voltage will be assumed to be 2.4 volts for the 24 Vdc supply. See Assumption 2.5.
Page 11 Frs EnerCALCULATION COMPUTATION NOP-CC--3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 4.4.5 Pressure Transmitter Transmitter M&TE Pressure gauge, 0-800 psig, +/-0.1% or equivalent 250 ohm precision resistor, +/-0.05% or equivalent Digital Multimeter (DMM), 0.03% or equivalent (for conservatism, 0.05% will be used)The calibrated span of the transmitter is less than the range of the test gauge so that the effect is: Pressure gauge = +/-0. 1%(800 psig/ 700 psig)= +/-0.114% span The equipment uncertainties are random and independent from each other. Therefore, they will be combined using the SRSS method.Transmitter M&TE = SRSS (0.114, 0.05, 0.05)= +/-0.134 % span 4.4.6 Pressure Transmitter Calibration Tolerance Calibration tolerance for the transmitters is +/-0.5% span. The larger of the effect for Calibration Tolerance or Accuracy will be included in the uncertainty calculation.
In this case the terms are the same, thus Accuracy will be used to determine the total uncertainty. (DIN 8, Section 6.2.6.2)4.4.7 Pressure Transmitter Uncertainty Pressure Transmitter Uncertainty
= SRSS (Acc, Drift, Temp Effect, Pwr Sup Eft, M&TE)= SRSS (0.5, 0.5, 0.91, 0.24, 0.134)= +/-1.18% span Page 12 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 4.5 Current Buffer (DIN 16, 21, 28, 35, 42)Component ID: Manufacturer/Model No.: Buffer Module Pt. No.: Current Buffer Pt. No.: Input Signal Range: Output Signal Range: Accuracy: Ambient Temp. Range: Temperature Effect: Power Supply: DB-PTCF4A 1 (11 1), DB-PTCF4A2 (1131)DB-PTCF4BI (112), DB-PTCF4B2 (I]2)Bailey / 6624610-2222 6624610-2222 6628999-1 1-5,Vdc #'+/-10 Vdc+/-0.2% span 40 to 1400 F+/-1.82% span (Dev. from Ref. Over Normal Range)+/-24 Vdc +/-0.5% (Normal: 22.8 to 25.2 Vdc)#14 volt span (within +/-10 Vdc); any current range which produces a 4 V span across an external range resistor.4.5.1 Current Buffer Accuracy The accuracy is specified as + 0.2% of span.4.5.2 Current Buffer Drift Effect Drift is considered equivalent to the reference accuracy of+ 0:2% span in lieu of any performance data.See Assumption 2.2.4.5.3 Current Buffer Temperature Effect The temperature change is assumed to be 15 degrees F (Assumption 2.7). As the temperature error for the instrument is assumed over a 100 degree range (40 to 140 degrees F), the error could be accounted for as 15% of that error. For conservatism, a value of 20% will be used.Temperature Effect = +1.82% span
* 20%= +0.36% span Page 13 ,Y CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C.ICE-051.0"O02 o TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 4.5.4 Current Buffer MTE Effect The calibration of the module is with a DMM having an accuracy of 0.03% or equivalent.
For conservatism, 0.05% will be used.4.5.5 Current Buffer Calibration Tolerance (DIN 21, 28, 35, 42)Calibration Tolerance for the Current Buffer Module is : 0.15% span. The larger of the effects for calibration tolerance or reference accuracy will be used to determine the device uncertainty.
In this case the reference accuracy is larger than the calibration tolerance. (DIN 8, Section 6.2.6.2);4.5.6 Current Buffer Total Uncertainty Current Buffer = SRSS (Accuracy, Drill, Temp Effect, M&TE)= SRSS (0.2, 0.2, 0.36, 0.05)= +/-0.461% span I. Page 14 F _n y CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE I
 
==SUBJECT:==
Core Flood Tank Pressure 4.6 Calculation of Summer+Bias+Inverter Unit (DIN 16, 22, 29, 36, 43) -Component ID: Manufacturer:
Model No.: Input Signal Range: Output Signal Range: Accuracy: Ambient Temperature Range: Power Supply: DB-PTCF4AI
(-K+E), DB-PTCF4A2
(-K+E), DB-PTCF4B1
(-K+E), DB-PTCF4B2
(-K+E)Bailey Controls, Inc.6623695+/-10 Vdc+/-10 Vdc+/-0.15% span (per RFA 90-0304, DIN 22, 29, 36, 43)40 to 140 0 F+/-24 Vdc Normal: 22.8 to 25.2 Vde 4.6.1 Sum+Bias+Inv Unit Accuracy Accuracy is equal to +0.15% span.4.6.2 Sum+Bias+Inv Unit Drift Drift is considered equivalent to the reference accuracy of +0. 15% span in lieu of any performance data. See Assumption 2.2.4.6.3 Sum+Bias+lnv Unit Temperature Effect Temperature effects are assumed to be included in the reference accuracy since no temperature effect is given and the unit is operated within the reference operating conditions.
See Assumption 2.3 4.6.4 Sum+Bias+Inv Unit MTE Effect The calibration of the Summer+Bias+Inv module is with a DMM having an accuracy of 0.03% or equivalent.
For conservatism, 0.05% will be used.
Page 1'CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 R0ON TITLE
 
==SUBJECT:==
Core Flood Tank Pressure 4.6.5 Sum+Bias+lnv Unit Calibration Tolerance The Calibration Tolerance for Summer+Bias+lnv module is ' 0.15% span. The larger of the effects for calibration tolerance or reference accuracy will be used to determine the device uncertainty.
In this case the calibration tolerance is equal to the reference accuracy, thus the accuracy uncertainty will be used.(DIN 8, Section 6.2.6.2)4.6.6 Sum+Bias+Ilnv Unit Uncertainty Sun+Bias+lnv Unit Uncertainty
= SRSS (Accuracy, Drift, Temp Effect, M&TE)= SRSS (0.15, 0.15, 0, 0.05)= +/-0.218% span
, Page 16 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 4.7 Voltage Buffer (DIN 16, 23, 30, 37, 44)Component ID: Manufacturer:
DB-PTCF4A I (EB 1), DB-PTCF4A2(EB 1)DB-PTCF4BI (EB3), DB-PTCF4B2(EB3)
Bailey Controls, Inc.Buffer Module Pt. No.: 6624610-2222 Voltage Buffer Pt. No.: 6624609-1 Input Signal Range: +/-10 Vdc Output Signal Range: +/-10 Vdc Accuracy:
+/-+0.1% span Ambient Temperature Range: 40 to 140* F Temperature Effect: +/-0.25% span (Dev. from Ref. Over Normal Range)Power Supply: +/-24 Vdc +/-0.5%Normal: 22.8 to 25.2 Vdc 4.7.1 Voltage Buffer Accuracy Effect Accuracy is specified as +/- 0.1% span.4.7.2 Voltage Buffer Drift Effect Drift is considered equivalent to the reference accuracy of+/- 0.1% span in lieu of any performance data.See Assumption 2.2.4.7.3 Voltage Buffer Temperature Effect The temperature change is assumed to be- 15 degrees F (Assumption 2.7). As the temperature error for the instrument is assumed over a 100 degree range (40 to 140 degrees F), the error could be accounted for as 15% of that error. For conservatism, a value of 20% will be used.Temperature Effect = +/-0.25% span
* 20%= +0.05% span Page 17 yCALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0.TITLE I
 
==SUBJECT:==
Core Flood Tank Pressure 4.7.4 Voltage Buffer MTE Effect The calibration of the voltage buffer module is with a DMM having an accuracy of 0.03% or equivalent.
For conservatism, 0.05% will be used.4.7.5 Voltage Buffer Calibration Tolerance The Calibration Tolerance for the Voltage Buffer Module is + 0.1% span. The larger of the effects for calibration tolerance or reference accuracy will be used to determine the device uncertainty.
In this case the calibration tolerance is equal to the reference accuracy, thus the accuracy uncertainty will be used.(DIN 8, Section 6.2.6.2)4.7.6 Voltage Buffer Uncertainty Voltage Buffer 7 SRSS (Accuracy, Drift, Temp Effect, M&TE)= SRSS (0.1, 0.1, 0.05, 0.05)= +/-0.158% span Page Is~CALCULATION COMPUTATION, NOPw-C .C3002-O11 Rev. 01 CALCULATION NO.: ... -EV)SION: C-4CE-051.01-02 0 TITLE I
 
==SUBJECT:==
Core Flood Tank Pressure 4.8 CFT Pressure Indicator (DIN 17, 25, 32, 39, 46)Component ID: Manufacturer:
Model No.: Input Signal Range: Output Signal Range: Accuracy: Linearity:
Repeatability:
Ambient Temp.: Temperature Effect: Power Supply Effect: Minor Scale Divisions:
DB-PICF4A 1, DB-PICF4A2, DB-PICF4BI, DB-PICF4B2 Bailey Controls, Inc.RY21IX 0-10 Vdc 0-700 psig (indication)
+1.0% span+/-1.0% span+0.5% span 40 to 1400 F+0.01% output span/degree F (Deviation from calibrating conditions within normal span)+/-0.01 % output span/volt AC 1I18 +/-1.0 Vac (Normal: 107 -127 Vac)2 psig x10(C5716) 4 4.8.1 Pressure Indicator Accuracy Effect Vendor Accuracy Effect is specified as 1.0% of calibrated span. The vendor has also specified a Repeatability and Linearity value. Per Assumption 2.6, the repeatability and linearity errors are included with accuracy and will be ignored.Accuracy Effect= +/- (Vendor Acc)= + 1.0% span 4.8.2 Pressure Indicator Drift Effect Drift is considered equivalent to the reference accuracy of +/-1.0% span in lieu of any performance data.See Assumption 2.2.
Page 19 CALCULATION COMPUTATION
__________NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE I
 
==SUBJECT:==
Core Flood Tank Pressure 4.8.3 Pressure Indicator Temperature Effect The temperature change is assumed to be 15 degrees F (Assumption 2.7). As the temperature error for the instrument is assumed over a 100 degree range (40 to 140 degrees F), the error could be accounted for as 15% of that error. For conservatism, a value of 20% (20 degrees F) will be used.Temperature Effect = -0.01% span / degree F* 20 degrees F= 0.20% span 4.8.4 Pressure Indicator Power Supply Effect The reference is specified as I I8v AC + 1.0v AC. The normal is specified a 107 -127 v AC. A worst case of 11 v AC (118 -107 v AC) will be used to establish the error.Power Supply Effect = +0.02% output span/volt AC
* I v AC= +0.22%4.8.5 Pressure Indicator Readability The CFT pressure indicators are located in the control room on panel C5716. The vertical scale indicates 0-700 psig of pressure indication.
There are minor divisions for every two psig x 10. The reading error is typically Y2 of the minor division or 10 psig. The readability will not be included in the uncertainty calculation.
It will be addressed in the Results section by rounding the final calculated value to a value able to be read on the indicator.
Indicator Reading Error = +/-0% span.4.8.6 Pressure Indicator MTE Effect The calibration of the Level indication meter is with a DMM having an accuracy of 0.03% or equivalent.
For conservatism, 0.05% will be used.4.8.7 Pressure Indicator Calibration Tolerance (DIN 25, 32, 39, 46)The Calibration Tolerance for the CFT Level Indication
+ 1.0% span. The larger of the effects for calibration tolerance or reference accuracy will be used to determine the device uncertainty.
In this case the calibration tolerance is equal to the reference accuracy. (DIN 8, Section 6.2.6.2)
Page 20.CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 4.8.8 Pressure Indicator Total Uncertainty Pressure Indicator= SRSS (Accuracy, Drift, Temp Effect, Power Supply Effect, Readability, M&TE)= SRSS (1.0, 1.0, 0.20, 0.22, 0, 0.05)= +1.446% span Page 21 YCALCULATION COMPUTATION NOP.-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 4.9 CFT Pressure Switches (DIN 17, 24, 31, 38, 45)Component ID: Manufacturer:
Model No.: Input Signal Range: Output Signal Range: Accuracy: Hysteresis:
Repeatability:
Ambient Temp.: Temperature Effect: Power Supply: DB-PSCF4A1, DB-PSCF4A2, DB-PSCF4B I, DB-PSCF4B2 Bailey Controls, Inc.6623819-1 0-10 Vdc Contact output+0.25% span about 0.05% span+0.1 % span 40 to 1400 F+0.25% output span over temp range 24 Vdc 4.9.1 Pressure Switch Accuracy Vendor Accuracy Effect is specified as 0.25% of calibrated span. The vendor has also specified a Repeatability and Hysteresis value. Per Assumption 2.6, the repeatability and hysteresis errors are included with accuracy and will be ignored.Accuracy Effect+/- (Vendor Ace)+/- 0.25% span 4.9.2 Pressure Switch Drift Effect Drift is considered equivalent to the reference accuracy of+0.25% span in lieu of any performance data. See Assumption 2.2.
Page '22 CALCULATION COMPUTATION NOP-CC-3002-0i Rev. 01 CALCULATION NO.: REVISION: C-iCE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 4.9.3 Pressure Switch Temperature Effect The temperature change is assumed to be 15 degrees F (Assumption 2.7). As the temperature error for the instrument is assumed over a 100 degree range (40 to 140 degrees F), the error could be accounted for as 15% of that error. For conservatism, a value of 20% will be used.Temperature Effect = +/-0.25% span
* 20%= +0.05% span 4.9.4 Pressure Switch MTE Effect The calibration of the Pressure Switches is with a DMM having an accuracy of 0.03% or equivalent.
For conservatism, 0.05% will be used.4.9.5 Pressure Switch Calibration Tolerance (DIN 24, 31, 38, 45)The Calibration Tolerance for the CFT Level Switches -0.25% span. The larger of the effects for calibration tolerance or reference accuracy will be used to determine the device uncertainty.
In this case the calibration tolerance is equal to the reference accuracy, thus the accuracy uncertainty will be used. (DIN 8, Section 6.2.6.2)4.9.6 Pressure Switch Total Uncertainty Pressure Switch= SRSS (Accuracy, Drift, Temp Effect, M&TE)= SRSS (0.25, 0.25, 0.05, 0.05)= +0.361% span Page 23 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C- CE-051.01.o02 0 TITLEI
 
==SUBJECT:==
Core Flood Tank Pressure 4.10 Computer Point- Multiplexer and AID Conversion Card (DIN 18)Manufacturer:
Model:
 
== Description:==
 
Input Range: Output Range: Accuracy: Manufacturer:
Model:
 
== Description:==
 
Input Range: Output Range: Accuracy: Modcomp 1873-1 Multiplexer Autoranging
-: 10mV A: 0.05%Modcomp 1870-1 Modacs Basic Card Autoranging 12 bit digital+/- 0.05%4.10.1 MUX -A/D Conversion Card Reference Accuracy (DIN 18)A/D-Multiplexer
= SRSS (A/D, Mux)= SRSS (.05, .05)= +/- 0.07% span Since the accuracy of the A/D-Mux (+/-- 0.07%) is an order of magnitude less than the calibration tolerance of the string, it will be neglected.
The 12-bit A/D is auto ranging which results in a quantizing error equal to V 2 of the Least Significant Bit (LSB). This is similarly a small error that will not be considered.
4.10.2 MUX -A/D Conversion Card Drift The Multiplexer/A-D pair is fully autoranging and not subject to drift.
Page 24 FCALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure 4.10.3 MUX -A/D Conversion Card Power Supply Effects The multiplexer and A-D card (1870/1873) are capable of a withstanding common mode voltage of up to +/- 200 Vdc or up to 120 Vac, and common mode noise rejection up to 132dB at up to 60 Hz, without a loss of accuracy.
Thus, variations in power supply voltage and frequency are estimated to have an effect bounded by the accuracy in terms of minor board signal variations.(DIN 18).4.10.4 MUX -A/D Conversion Card Temperature Effects (DIN 18)During normal conditions, the Computer Room Temperature is controlled at or very near 75 F.As such, no additional drift will be assumed due to temperature drift. It should be noted that the output of these cards changes relatively little due to temperature variation (approximately 0.01%/ C) and so even a large variation of+ 10 F would not significantly change their output.4.10.5 MUX -A/D Conversion Card Humidity Effects There is no published humidity effect data on the Multiplexer/A-D card pair. The cards are normally maintained in a temperature and humidity controlled environment.
Humidity effects are estimated to be negligible.
4.10.6 MUX -A/D Conversion Card Pressure Effects The Multiplexer/A-D card pair is installed in the computer room, which is maintained at atmospheric pressure, or slightly higher. There is no pressure effect.4.10.7 MUX -A/D Conversion Card Vibration Effects The Multiplexer/A-D card pair is installed in the computer room, which is not subject to vibration during normal operation.
This instrument is not required to operate during a seismic event. There is no vibration effect.
Page 25 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE I
 
==SUBJECT:==
Core Flood Tank Pressure 4.10.8 MUX -A/D Conversion Card Calibration Accuracy (DIN 18)These Multiplexer/A-D modules are "string" calibrated together, and will be treated as a single module. The larger of the effect for Calibration Tolerance or Accuracy will be included in the uncertainty calculation. (In this case, the Calibration Tolerance is larger.) (DIN 8, Section 6.2.6.2).
From the data sheet: tolerance
= 0.5% span.4.10.9 MUX -A/D Conversion Card M&TE Since the M&TE (DMM) accuracy is < 1/10 of Calibration Accuracy it will be neglected.
4.10.1 OMUX -A/D Conversion Card Indicator Readability The digital indication is readable to the tenths digit, which is a negligible contribution.
4.10.11MUJX
-A/D Conversion Card Multiplexer/A-D Card Uncertainty Computer Loop Uncertainty
= Calibration Tolerance+/- 0.5% span
, ~Page 26.RmtE .CALCULATION COMPUTATION.
NOP.;CC-3002..01 Rev. 01 CALCULATION NO.: ,RVSO: , C-ICE-051.01-002 R TITLE I
 
==SUBJECT:==
Core Flood Tank Pressure 5.0 RESULTS CFT Pressure Uncertainties Instrument Instrument ID (Module #) Uncertainty Section Pressure Transmitter DB-PTCF4Ai, DB-PTCF4A2
+/-1.18% span 4.4.7' _ _ DB-PTCF4B1, DB-PTCF4B2 Current Buffer DB-PTCF4A1 (IBI) (5-1-2) +/-0.461% span 4.5.6 DB-PTCF4A2
(]B31) (6-4-11)DB-PTCF4B1 (1B2) (5-1-2)DB-PTCFnB2 (1B2) (6-4-11)Summer+Bias+Inverter DB-PTCF4AI
(-K+7,) (5-1-7) +/-0.218% span 4.6.6 DB-PTCF4A2
(-K+7) (6-4-8)DB-PTCF4B1
(-K+E) (5-1-8)DB-PTCF4B2
(-K+E) (6-4-9)Voltage Buffer. DB-PTCF4AI,(EB1)'(5-1-1)
+/-0.158% span 4,7.6 DB-PTCF4A2 (EBI) (6-4-10)DB-PTCF4B1 (EB3) (571-1)DB-PTCF4B2 (EB3) (6-4-10)Pressure Indicator DB-PICF4A1, DB-PICV4A2
+/-1.446% span 4.8.8 DB-PICF4B1, DB-PICF4B2 Pressure Switches DB-PSCF4AI, DB-PSCV4A2
+/-0.361% span 4.9.6 DB-PSCF4B1, DB-PSCF4B2.
Computer Point P089, P090, P079, P080 +/-0.5% span 4.10.11 Total Uncertainty for Core Flood Tank Indication DB-PICF4A1, DF~-P1CF4A2 DT~-PTCF41~
1 DB-PICF4A2 B-PICF4B I DB-P1CF4B2 Indicator
= SRSS (Xmtr, Current Buffer, Sum+Bias+Inv,.
lnd)+ Process Measurement Effect=SRSS (1.18, 0.461,0.218, 1.446)+ 0-+/- 1.93% of 700 psig span or +/- 13.51 psig To protect the Analytical Limits in DIN 61, the indicators must read between 580.51 psig (567 psig + 13.51 psig) and 619.49 psig (633 psig -13.51 psig).Due to readability of the indicators, the indicated pressure must be between 590 and 610 psi g.
Page 27 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 V0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure Total Uncertainty for Computer Point P089, P090, P079, P080 Computer SRSS (Xmtr, Current Buffer, Sum+Bias+Inv, Voltage Buffer, Computer Pt)+Process Measurement Effect= SRSS (1.18, 0.461, 0.218, 0.158, 0.5)+ 0= 1.39% of 700 psig span or + 9.73 psig The computer point indications are accurate to Within +/- 9.73 psig. To protect the Analytical Limits in DIN 61, the computer points must read between 576.73 psig (567 psig + 9.73 psig) and 623.27 psig (633 psig -9.73 psig).With margin, the surveillance of the computer points should be between 580 psig and 620 psig.Total Uncertainty for Core Flood Tank Switches DB-PSCF4Al, DB-PSCF4A2.
DB-PSCF4B I.DB-PSCF4B2 Switches = SRSS (Xmtr, Current Buffer, Swn+Bias+lnv, Switch)+ Process Measurement Effect= SRSS (1.18, 0.461, 0.218, 0.361)+ 0+ 1.34% of 700 psig span or +/- 9.38 psig Page 28 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 01 CALCULATION NO.: REVISION: C-ICE-051.01-002 0 TITLE /
 
==SUBJECT:==
Core Flood Tank Pressure
 
==6.0 CONCLUSION==
S This calculation determined the instrument uncertainty for measurement of Core Flood Tank Pressure by indicator DB-PICF4A1, DB-PICF4A2, DB-PICF4B1, DB-PICF4B2 or computer point P089, P090, P079, P080. The instrument errors were determined for normal conditions.
The calculation also determined the instrument uncertainty associated with the Core Flood Tank Pressure switches, which provide annunciation to the operator when pressure is too high or too low.The setpoints are 615 and 585 psig (DIN 12). With a 10 psig difference between the alarm setpoints and the Technical Specification values of 625 and 575, the uncertainty of 9.38 psig will ensure the Operator is alerted to a high or low pressure prior to reaching the Technical Specification values. As there is an additional 8 psi between the Technical Specification values and the Analytical Limits, there is additional protection of the Analytical Limits.The measurement uncertainties determined above are provided for use in the surveillance testing used to verify adequate pressure for the Core Flood Tank. The current procedural (DIN 47)requirement for Technical Specification compliance is a maximum of 625 psig and a minimum of 575 psig as provided by indicators DB-PICF4A1, DB-PICF4A2, DB-PICF4BI, DB-PICF4B2.
Since the error for the indicators is 13.51 psig and the margin between the Analytical Limit and the Technical Specifications is only 8 psig, verification of the Core Flooding Tank pressure with an acceptance criteria equal to the Technical Specification value is considered non-conservative.
This is documented in Condition Report 05-00381 (DIN 61). This is currently considered acceptable due to the alarms alerting the operator prior to reaching the Technical Specification values. It is recommended that the procedure be revised to change the acceptable values and use the computer points as the preferred pressure indications.
AN 0 p8f)42 I OF0I 88 5006232 01 Frm tm ANPVpb .. ... ....Large Break Small Break (1080 used in analysis)
(1080 used In analysis)CFT Liquid Temperature (F) 120 120 CFT Gas Pressure (psia) 615+/-33 615+/-33 (582 used in analysis)
(582 used in analysis)CFT Surge Line K Factor (based on 7 0 700 0 (Note 5)an area of 0 7213 ft 2) ..........
Volume of Surge Line (ft'/line) 40 40 Elevation Change of Surge Line (in) 5375 5375 HPI Flow 2 x Table 3 6 (Note 7) Table 3 6 Table37 &Table 3 8 LPI Flow 2 x Table 3 5 (Note 7) Table 3 5 RCP Parameters Reactor Coolant Pump Trip On LOOP (Note 6) 1n.OOP or LSCM+2 Min OLO(Note 8)Reactivity Control Parameters (Note _ __Control Rod Drop Time (s) Not Modeled 1 4 to 2/3 Insertion Full Insertion Rod Worth (%&k/k) Not Modeled 2 26 Control Rod Insertion vs Time Not Modeled Table 3 9 Delayed Neutron Fraction 0.0 0007102 0 007102 Prompt Neutron Generation Time (s) 0 248x10' 0 248x10'4 Doppler Coefficient 2 0x10"* @1420 F 2 0xI0"* @1420 F Moderator Temperature Coefficient at 0 0 0 0 100% FP(pcm/F) (Table 3 10) (Table 3 10)SS Fuel Pin Energy Deposition As stated in Results As stated in Results Transient Fuel Pin Energy Deposition As stated in Results As stated In Results Fuel Parameters
_" FuelType Mark B10A Mark B10K Mark B10A Mark B10K Mark B12 -U0 2 & Gadolinia Mark B12 -U0 2 & Gadolinla U0 2 Enrichment
(%) < 5 1 .. 5 1 Gadolinia Enrichment
(%) 2, 3,4, 6, & 8 NA Containment Parameters Containment Pressure (psia) Calculated w/ CONTEMPT 70 using parameters on Table 3 11 NOTES I After reactor trip the decay heat generation rate calculation will be based on 1 2 times the ANS 1971 standard for fission products plus B&W heavy actinides 2 RCP thermal input to the RCS is NOT an input value It is calculated by the computer code and depends on the pump component input and fluid properties 3 Initial MFW flow will be adjusted during steady state Initialization to obtain the appropriate heat balance 4 The turbine header pressure may be adjusted during steady state initialization to obtain the appropriate heat balance and T 5 The CFT line resistance is Increased by a factor of 100 for all SBLOCAs except the CFT line break (Ref 3)6 LOOP occurs coincident with break opening for LBLOCA 7 The use of maximum ECCS is conservative in the context of the evaluation model and the effect that additional ECCS has on the minimum calculated containment pressure (see Section 4 1 2 3 for details) By no means should this analysis choice be construed as justification for limiting or reducing the actual plant ECCS flows 8 The operators have up to 2 minutes after loss of subcoollng margin to trip the RC pumps if a LOOP does not occur For cases that consider LOOP LOOP occurs coincident with reactor trip 33 CALCULATION NO. j CALCULATION REVISION NO.C-ICE-051.01-002 -I 0 I I INITIATING DOCUMENT ORIGINATOR/DATE NIA I N 1 10/18/07 DESCRIPTION:
During the next update of the calculation, revise the conversion from absolute pressure to atmospheric pressure from 15.0 psi to 14.7 psi. The current analytical limits are in psia. Since a higher pressure will inject more water into the core, use of 15.0 psi for the conversion (converting 615 psla to 600 psig) is non-conservative compared to the 14.7 psi conversion factor (converting 615 psia to 600.3 psig).There is no change in the final calculated results. The calculated values for low and high pressure for the indicators are 580.51 psig and 619.49 psig, respectively.
Due to readability, thevalues are revised to 590 and 610 psig. With this additional margin, increasing the conversion value by 0.3 psi will not result In a change to the final value in the calculation.
The calculated values for low and high pressure for the computer points are 576.73 psig and 623.27 psig, respectively.
Additional margin is added with resulting values of 580 and 620 psig. With this additional margin increasing the conversion value by 0.3 psi will not result in a change to the final value In the calculation.
I -,_ ........ ... iii P ag e I QyCALCULATION NOPCyCC-3002-01 Rev. t0 INIeIATIN
? YOCUMENT () Y.s, CALCULATION NO. NVENOR CALC SUMMAeY StR 03-02N54b -ICE-051.01-001
'rLEfLF/
 
==SUBJECT:==
 
Uncertainty Calculation for Core Flood Tank Level Indication r ] Bva Nm BV2 Re DB Er i p Y cosegory 173 Active Wo HdPtoic'e s3 s tody Classification Safety-Related/Augmented Quality DeNonsigety-Related Open Assumptions?
o Yes h No If Yes, Enter CR gacking Number System Number: 51-01 Asset Number; LICF3AJ (A2, B1, & B2)Commnitments:
0755"'(Perry Only) Calculation Type: Referenced In Atlas? 0_ Yes El No e Ae ae/ j Referenced In USAR Validatin Database AYes DaNo Computer Program(S)
Program Name Version /Revision Category Status Description Microsoft Word 97 C Active Word Processor Revision Record Rev. Affected Pages Onginator/Date j Reviewer/Date Design Verifier/Date Approver/Date Description of Ch dran-go: Establishes khe iFRAnero uncertainty, astocirwt Core- [tTWO k Wle, dication and alanns..Describe where the calculation has been evaluated for I10CFR50.59 applicability.
03-01316 Ple'. Affected Page3 Originator/Date Reviewer/Date
-Design Verifier/Date Approver/tte Description of Change Describe where the calculation has been evaluated for 10CFR50.59 applicability.
Rev. Affected Pages Originator/Date Reviewer/Date ,Des ign Verffier/Date Approver/Date Desedption of Change: Deciewhere the calculation has been evaluated for IOCFR50.59 applicability.
Rev. Affected Pages Originator/Date Reviewer/Date Design Verifier/Date
* Approver/Date Description of L-hangv., Describe where the calculation has been evaluated for 10CFR50.59 applicab/iliy.
C i Page ii Ff_ trW CALCULA11ON NOP-CG-3002-01 Rev. 00 INITIATING DOCUMENT (S) CALCULATION NO.IVEDRCLSUMY CR 03-02547 C-ICE-051.01-001 TITLE/
 
==SUBJECT:==
 
Uncertainty Calculation for Core Flood Tank Level Indication TABLE OF CONTENTS SUBJECT PAGE COVERSHEET:
OBJECTIVE OR PURPOSE iii SCOPE OF CALCULATION iii
 
==SUMMARY==
OF RESULTS/CONCLUSIONS iii LIMITATIONS OR RESTRICTION ON CALCULATION APPLICABILITY iii IMPACT ON OUTPUT DOCUMENTS iii DOCUMENT INDEX iv CALCULATION COMPUTATION (BODY OF CALCULATION):
ANALYSIS METHODOLOGY I ASSUMPTIONS 2 ACCEPTANCE CRITERIA 3 COMPUTATION 4 RESULTS 18 CONCLUSIONS 19 ATTACHMENTS:
ATTACHMENT 1: 0 Pages ATTACHMENT 2: 0 Pages TOTAL NUMBER OF PAGES IN CALCULATION (COVERSHEETS
+ BODY + ATTACHMENTS) 24 Pages SUPPORTING DOCUMENTS (For Records Copy Only)DESIGN VERIFICATION RECORD I Pages CALCULATION REVIEW CHECKLIST 2 Pages I 0CFR50.59 DOCUMENTATION 3 Pages DESIGN INTERFACE
 
==SUMMARY==
I Pages DESIGN INTERFACE EVALUATIONS 21 Pages OTHER 0 Pages o YES EXTERNAL MEDIA? (MICROFICHE, ETC.) (IF YES, PROVIDE LIST IN BODY OF CALCULATION) 11 -NO 1.
Page Hil ,. CALCULATION NOP-CC-3002-01 Rev. O00, INITIATING"DOCUJMENT (S) CALCULATION NO. [ EDRCALC
 
==SUMMARY==
CR 03-02547 C-ICE-051.01-001 TITLE/
 
==SUBJECT:==
 
Uncertainty Calculation for Core Flood Tank Level Indication OBJECTIVE OR PURPOSE: This calculation determines the instrument uncertainty associated with the Core Flood Tank (CFT) level indication, high/low alarms and plant computer level indications.
The results of this calculation will be applied to control room indications that are used to satisfy TS 3.5.1.SCOPE OF CALCULATI-N/RE VISION: This calculation only analyzes the core flooding tanks control room level indications, control room high/low alarm, and the plant computer level indications.
This analysis determines the total string uncertainty for 51 -ISLCF3AI, A2, B1, & B2.
 
==SUMMARY==
OF RESULTS/CONCLUSIONS:
Based upon the total uncertainties calculated for the strings, the existing instrumentation can provide acceptable performance of the surveillance with the procedure and data package changes identified under"Impact on Output Documents." No field changes are required as a result of this calculation.
The following are the surveillance values/settings based on the uncertainties associated with each of the instrument strings: Indicators:
-13.25, > 12.75 feet Computer:
< 13.3, > 12.6 feet Alarm Setting: < 13.3, >- 12.7 feet LIMITATIONS OR RESTRICTIONS ON CALCULATION APPLICABILITY:
None.IMPACT ON OUTPUT DOCUMENTS:
The calculation of the Core Flooding Tank level indicators uncertainties will be used to verify the.adequacy of surveillance criteria and alarm settings that support verification of TS 3.5.1. At this time, no field setting changes are required since the calculated values are supported by existing settings.Section 4.6 of DB-OP-03006 and the associated Data Sheet will require revision to record the Computer points as the preferred input, backed up by the control board indicators.
The indicator criteria will require a minor revision to reflect the new values.I&C Data Packages for the indicators will require modification to implement a reverse calibration.
I Page 1v~CALCULATION N O P-C C .3002-0 1 R ev. ODIF0N O C L U M R INITIAT'ING DOCUMENT (S) .....CALCULATION NO. }VEDRALSUAY CR 0302547 C-ICE-051.01-001 TITLE/
 
==SUBJECT:==
 
Uncertainty Calculation for Core Flood Tank Level Indication DOCUMENT INDEX 6 z 0Document Number/Title Revision, Edition, Date I Updated Safety Analysis Report, Section 6.3.3.2.4, Minimum Conditions of Rev. 23, 11/02 0 0- El ECCS 2 Technical Specifications, Section 3/4.5, Emergency Core Cooling Systems Rev. 257, 2/26/03 9 El 1:]3 M-0034, Piping & Instrumentation Diagram, ECCS Containment Spray and Rev. 55 El] E L]Core Flooding System 4 J-1I I Sh. ISA, Loop Diagram Core Flood Tank 2 Level (LT-CF3A I) Rev. 0 0 [- El 5 J-I I ] Sh. 15B, Loop Diagram Core Flood Tank 2 Level (LT-CF3AI)
Rev. 0, 0 Dl []6 J- I I I Sh. 16A, Loop Diagram Core Flood Tank 2 Level (LT-CF3A2)
Rev. 0' 0] El]7 J-1 II Sh. 16B, Loop Diagram Core Flood Tank 2 Level (LT-CF3A2)
Rev. 0 D 0 8 3-I 11 Sh. 17A, Loop Diagram Core Flood Tank I Level (LT-CF3B 1) Rev. 0 El D ]9 J-III Sh. 17B, LoopDiagramCoreFloodTank 1 Level(LT-CF3B1)
Rev. 0 0 [D] E 10 J- Ill Sh. 18A, Loop Diagram Core Flood Tank 1 Level (LT-CF3B2)
Rev. 0 El El []11 IJ-I I Sh. 1 8B, Loop Diagram Core Flood Tank I Level (LT-CF3B2)
Rev. 0 0 El [0 12 M-5 18-00040, Assembly of Vessel for the Core Flooding Tank Rev. 0 z E" I 13 M-518-00035, Instruction Manual -Core Flood Tank Rev. 2 ElQ El]14 C-ME-51.01-086, Tank Level Curve Calculation
-Core Flood Tanks (T9-1, Rev. 1 0 T9-2)15 ISA-RP67.04.02-2000, Methodologies for the Determination of Setpoints for Approved I 0 El E]I Nuclear Safety-Related Instrumentation January 2000 16 DBI-100, Davis-Besse Nuclear Station Unit I. Environmental Qualification Rev. 10 -of Electrical Equipment 17 Asset Data Base Version 7.6, 0 [] El]Release 17 18 SD-040, Core Flooding System Rev. 2 23 0 0 19 SD-05 1, System Description for the Non-Nuclear Instrumentation System Rev. 1 0 El El]20 C-NSA-05 1.01-002, Minimum CFT LB LOCA Assumptions Rev. 0 El 9 0 1-Page v Fir tlneiv CALCULATION NOP-CC-3002-01 Rev. 00 INITIATING DOCUMENT (S) CALCULATION NO. [] VENDOR CALC
 
==SUMMARY==
CR 03-02547 C-ICE-051.01-001 TITLE/
 
==SUBJECT:==
 
Uncertainty Calculation for Core Flood Tank Level Indication z 0 Document Number/Title Revision, Edition, Date a)Ii I.21 C-NSA-05 1.01-001, Core Flood Tank Technical Specification Level Range Rev. 0 0 N 0 22 DB-OP-02003, 3-I-F, 3-1-G, 3-3-F, 3-3-3 Rev. 3 0 0 9 23 DB-OP-03006, Miscellaneous Instrument Shift Check Rev. 8 0] "1 24 DB-MI-04254, String Check of 5 1A-ISLCF3AI Core Flooding Tank 2 Level Rev. 02 12 [] 0 25 DB-MI-04255, String Check of SIA-ISLCF3A2 Core Flooding Tank 2 Level Rev. 02 ] 0'0 26 DB-MI-04256, String Check of 5IA-ISLCF3B I Core Flooding Tank I Level Rev. 02 [ E..27 DB-MI-04257, String Check of 51A-ISLCF3B2 Core Flooding Tank 1 Level Rev. 03 0 0 0 28 DB-MI-09026, Backing Filling Process Sensing Lines and Reference Legs Rev. 02 0 0 0 29 DB-I, Electrical Equipment Qualification File Number DBI-100, Tab-2.1 Rev. 9 [] 0 0l 30 ICDP 51-JSLCF3AI, A2, BI, B2 Rev. 2 El N 31 M-536-00118, Module Instruction Book for NNI and ICS Vol 5 Rev4 [] 0 11 32 M-530-00353, Integrated Control & Non-Nuclear Instrument Peripheral Rev. 6 [Equip. Vol. 1 Rev. 6 33 M-7201, Instrument Index Rev. 50 0 El El A Vendor Manual G-CS-406-2, MODCOMP Vol. XIV, "Process Re. 0, 2/84 Input/Output Subsystem" 35 System Description SD-029A, "System Description for Control Room Rev. 3, 8/25/95 0 0 Normal Ventilation" 36 Framatome Calculation 86-5006232-01, LOCA Analysis (including Rev. 1, 3/03 0 assumptions related to CFT performance) 37 NES-87-10018,
 
==Subject:==
Maximum Specific Gravity of Water in 11/24/86 0 BWST & BAAT I I Page 1 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 00 CALCULATION NO.: C-ICE-051.01-001 TITLE I
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 1.0 ANALYSIS METHODOLOGY ISA RP67.04.02-2000 defines the uncertainties that should be addressed in the calculation.
Although other uncertainties may be included, the following list from the standard includes the uncertainties to be accounted for in this calculation.
Instrument calibration uncertainties Calibration standard Calibration equipment Calibration method Instrument uncertainties during nolnmal operation Accuracy including linearity, hysteresis, dead band and repeatability Power supply voltage changes Power supply frequency changes Temperature changes Humidity changes Pressure changes Vibration Radiation effects Analog to digital conversion Digital to analog conversion Instrument drift Design basis event effects Temperature effects Radiation effects Seismic effects Process dependent effects Calculation effects Dynamic effects Calibration and installation bias accounting i
Page 2 Ftr _rlt'jj CALCULATION COMPUTATION
~NOP-CC-3002-01 Rev. 00 CALCULATION NO.: c-IcE-051
.01 -001 TITLE I
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication The uncertainties will be discussed below. Each uncertainty that is zero (0) will not be included in the forrrmulas.
Any effects that are for abnormal or accident conditions will not be addressed in this calculation.
The calibration standard will not be included in the calculation.
There is at least an 8:1 accuracy ratio between the calibration standard and the installed instrument based on controls in procedure NOP-WM-5001. This calibration standard accuracy component, when included as a component in the Square Root Sum of the Squares (SRSS), results in a final calculated value that is less than 1% larger than the accuracy of the installed instrument.
This is two orders of magnitude less, which is considered negligible.
The surveillance test procedures in conjunction with the IC Data Packages specify the calibration equipment to be utilized.2.0 ASSUMPTIONS 2.1 Ambient Temperature
-The area temperature (Containment 565', Zone 217) is normally at 55 to 110&deg;F. (Ref. 29)2.2 Process Measurement Errors -The only Process Measurement Error for the CFT is the boron contribution to the level indication.
The transmitters are calibrated for a boron concentration of 3500 ppm. This is the maximum allowed by Tech Specs for the tank. The minimum allowed by Tech Specs is 2600 ppm. The potential level difference associated with the boron concentration difference will be calculated.
The only other error could be temperature effect on the reference leg. The CFT level transmitters are differential pressure transmitters and compare a reference leg to that of the level within the CFT. There are no temperature effects or significant changes in water densities to induce any stratification effects that could impact measurement of the process. Since density changes, within the reference leg and tank would counteract each other, this effect is considered negligible.
2.3 Drift Effects -Drift will be assumed to be equivalent to accuracy in cases where the manufacturer has not specified any drift value.
* .Page 3 ,L f RY CALCULATION COMPUTATION" ~~~NOP-CC-3002-01 Rev. O00...CALCULATION NO.: C-lC'E-051.01-001 TITLE /
 
==SUBJECT:==
* Uncertainty Calculation for Core Flood Tank Level Indication 2.4 Temperature effects -Temperature effects that are not specified by the Vendor are assumed to be included in the Vendor provided accuracy if the instrument is operating within the referenced temperature range.* 2.5 Calibration temperature and operating temperature is assumed to be 60-90 degrees for the Core Flooding Tank level transmitters.
This is a reasonable assumption based on the fact that the transmitters are located on the 565-foot level in the containment building.
This area is generally cooler than the average containment temperature because the Containment Air Coolers discharge to the lower elevations.
This airflow arrangement tends to moderate any extreme temperatures in the lower containment elevations.
Using the worst case 55-110 degrees (Ref.16) range is an overly conservative assumption that is not appropriate for these transmitters.
2.6 The specification for Rosemount 1153 transmitters are 3-sigma values. Based on section Annex J. I in reference 15, 3-sigma could be considered unnecessarily conservative.
Therefore, the appropriate transmitter instrument uncertainties will be adjusted from 3- sigma to 2-sigma.(Note -the higher confidence level does not apply to the drift uncertainty).
3.0 ACCEPTANCE CRITERIA The acceptance criteria for the calculation of measurement uncertainties for CFT level indication and alarms will be thatthc work is done in accordance with the methodology of ISA RP67.04.02-2000, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation.
: i.
Page 4 , F CALCULATION COl PUTATION NOP-CC-3002-01 Rev. 00 CALCULATION NO.: C-lCE-05i.01-001 TITLE /
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication
 
==4.0 COMPUTATIONS==
 
==4.1 INTRODUCTION==
 
This calculation determines the instrument uncertainty associated with the Core Flood Tank (CFT)level indication and high/low alarms. The results of this calculation will be applied to control room indications that are credited in satisfying TS 3.5.1. This uncertainty will be used as inputs for the associated procedures to establish the limits to meet the required levels as specified in TS 3.5.1 for Core Flooding Tank Levels.There are four independent level transmitters (2 for each CFT) that provide a 4-20 mA signal, which is ultimately converted, to 0-1 0 volts DC for level indication and alarms. The high/low level alarms are only to alert control room operators that level in the CFT(s) is changing.
The CFT high/low alarm function does not provide any automatic actuation.
CFT levels are monitored on control room panel C5716. The CFT level indicators are dual purpose in that each indicator shows pressure as well as level. Additionally, the CFT transmitters provide an input to the plant computers as an alternate method to monitor CFT levels and control room alarm panel.The affected instruments for this instrument loop: Core Flooding Tank Level (Ref. 4-11)LT-CF3A1, LT-CF3A2, LT-CF3B1, LT-CF3B2 LT-CF3AIA, LT-CF3A2A, LT-CF3B1 A, LT-CF3132A LY-CF3AI, LY-CF3A2, LY-CF3B1, LY-CF3B2 LI-CF3A1, LI-CF3A2, LI-CF3B1, LI-CF3B2 LS-CF3A1, LS-CF3A2, LS-CF3BI, LS-CF3B2 LT-CF3AIB, LT-CF3A2B, LT-CF3BIB, LT-CF3B2B Computer Points L079, L080, L085, L090-CFT Level Transmitters
-CFT Current Buffer (1B3/IB4)-CFT Signal Conditioner
(-K &#xf7;E)-CFT Level Indicator-CFT Level Switch Monitor-CT Voltage Converter (EB4/EB2)-Multiplexer/
A-D Converter Note: EMPAC assigns the "A" and "B" designator to differentiate between the transmitters, Current Buffer and Voltage Converters.
This is done because the drawing identifies all three components with the same asset number.It
~Page 5" F lrst- CALCULATION COMPUTATION j; NOP-CC-3002-01 Rev. 00 CALCULATION NO..C-ICE-051.01-001, TITLE /
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 4.2 Loop Diagram for Core Flooding Tank (CFT) Level The instrument loop shown below is provided as a layout of the components for CFT level indication and alarms. (Ref 4-11)CFT LEVEL (Typical)Rosemount 1I 53DD5 LT-CF3Al 0-750 inwc Input 168-0 inwc 4-20 mAmp 250 Ohm' ' 1-5 VDC IBailey IB3 6624610-2222
_ I-10 to +10 VDC-K +E Bailey 6623695 0-10 VDC Bailey 6623819-1/
LS-CF3A1 LI-CF3A1 LT-CF3A1B RY-211X/I 6624610-1111 L079, LOS0, L085, L090 i ,11Page 6CALCULATION COMPUTATION
~~~NOP-CG-3002-01 Rev. 00.....CALCULATION NO.: C-ICE-051.01-001 TITLE /
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 4.3 FUNCTIONAL DESCRIPTION/DESIGN BASIS The Core Flooding System functions as part of the Emergency Core Cooling System. The CFT level instrumentation provides indication of level in the control room via LI-CF3A 1, A2, B1, and B2. The CFT level instrument string also provides input to the plant computer and a high/low alarm. DBNPS TS 3.5.1 states the Mode 3 requirement for CFT level is 7555 and 8004 gallons. CFT level instruments are located on panel C-5716 in the main control room. These instruments (LICF3A1, A2, B1, and B2)are read in feet (0-14). The control room meters serve a dual purpose in that they use two inputs; the first input is CFT level and the second is CFT pressure.The surveillance criteria have been derived from consideration of the Tech Spec values and the Analytical Liriits that form the basis of the spec. The LOCA analysis (Ref. 36) provides the values assumed and evaluated for the performance of the CFTs during a design basis accident.
Included in the analysis is an allowance for instrument uncertainty, equal to the difference between the Analytical Limit (AL) and the Tech Spec. This value is equal to 75 gallons or 0.148 ft of level based on the tank curve of Ref. 14. Though this 0.148 ft is not sufficient to bound all of the applicable uncertainties, it does amount to the majority of the error. For the purposes of this calculation, the AL will be used as the starting point for uncertainty evaluation, and the alarm/surveillance criteria will be based on the AL adjusted for the calculated uncertainty.
The Tech Spec values are, therefore, considered nominal.The level transmitters use a 24 VDC power supply and provides a 4-20 mAinp output. The 4-20 mAmp output is converted to a 1-5 VDC signal across a 250-Ohm resistor.
The 1-5 VDC signal is inputted to a current buffer (1B3/4) which changes the signal to -10 to +10 VDC. The -10 to +10 VDC signal is converted to 0 to +10 VDC by a Signal Conditioner
(-K +E). This 0 to +10 VDC signal is the input for the control room level indicators, plant computer, and alarms.The CFT Level loop includes instrument LY-CF3A/B and its associated signal conditioning components and output modules. The following device uncertainties are calculated for each component.
The devices are identified as shown in the EMPAC database and/or on Drawing M-034 (Ref. 3). The module numbers are shown to aid in identification of the devices.
Page 7 CALCULATION COMPUTATION 3002-01 Rev. 00 CALCULATION NO., C-ICE-051.01-001 TITLE /
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 4.4 Level Transmitter Uncertainty
-(LT-CF3A1, LT-CF3A2, LT-CF3B1, LT-CF3B2)(Ref. 32)Component I.D.: Manufacturer/Model No.: Range Limits: Calibrated Range: Input: Output Signal: Calibration Period: Baseline Accuracy: Drift: Temperature Effect: Humidity Effect: Overpressure Effect: Static Pressure Effect: Power Supply Effect: LT-CF3A 1, LT-CF3A2, LT-CF3B 1, LT-CF3B2 Rosemount/
1153DD5 0-125 to 0-750 ini-H20 168-0 inwc 4-20 mAdc 30 months (Refueling cycle plus 25%)+/- 0.25% Span @ 3 Sigma, +/-0.167% Span @ 2 Sigma+ 0.2% (30 months)+ (0.75% URL + 0.5% Span)/100IF 0% (0-1 00%RH)-1% of URL > 3,000 psi+ 0.2% URIJ1000 psig (Zero Eff.) +/-0.5% of reading/1000 psig (Span Eff.)+/- 0.005% of output span per volt 4.4.1 Specified Drift The transmitters are calibrated on an 24-month interval and the Technical Specifications allow for a 25% extension (TS 4.0.2). Therefore, the calibration period can be extended to 30 months.Drift= +/- 0.2 (750 inH20/168 in}1 2 0) (30/30)+/- 0.89% Span 4.4.2 Temperature Effect The transmitters are wall mounted in the containment building El 565'. (Assumption 2.5 and Ref. 29).'Temperature Effect = (0.75%(URIJSpan)
+ 0.5% Span) AT/100 @ 3 Sigma+/- ([0.75% (750 inH 2 O/168 inHz0) + 0.5% Span] (90-60/100)
(2/3)+0.77% Span @ 2 Sigma 1
~Page 8: ~CALCULATIN COMPUTATIN
: ~NOP.-OC-3002-01, Rev. DO CALCULATION NO.:* C-ICE-051.01-001 TITLE /
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 4.4.3 Static Pressure Effect (Range 5)The Static Pressure Zero Effect and Span Effect is associated with differential pressure transmitters.
The Static Pressure Zero effect has been calibrated out in the ICDP packages (Ref. 30). However, there remains a span effect correction uncertainty which cannot be calibrated out and this must be included in the determination of the total loop uncertainty.
The maximum operating pressure for this process is 600 psi.Span Effect 1 0.5% input reading/1000 psi @ 3 Sigma= +/- [0.5% (600/1000)]
(2/3)=+/- 0.20 % Span @ 2 Sigma 4.4.4 Transmitter M&TE (Ref. 24-27)Pressure gauge, 0-200 inwc, +1-0.05% or equivalent 250 ohm precision resistor, +/-0.01% or equivalent Digital Multimeter (DMM), 0.03% or equivalent The equipment uncertainties are random and independent from each other. Therefore, they will be combined using the SRSS method. Pressure gauge uncertainty
= +/-0.05% span (dp) The calibrated span is 168 inches, so that the pressure gauge uncertainty is 0.06% span in units of the instrument loopr.The resistor converts a 4-2OmA signal to a 1-5 Vdc signal. The accuracy of the 250 ohm resistor is+/-+0.01%: Resistor = +/-0.01% span (dp)Transmitter M&TE = SRSS (0.06, 0.01, 0.03) @ 3 Sigma= SRSS (0.06, 0.01, 0.03) (2/3)= +/-0.045 % span dp @ 2 Sigma 4.4.5 Calibration Tolerance (Ref. 24-27)i L.V I, CALCULATION NO C-ICE-051.01-0 TITLE /
 
==SUBJECT:==
Page 9 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 00)01 Uncertainty Calculation for Core Flood Tank Level Indication Calibration tolerance for the transmitters is +/-0.25% span. The larger of the effect for Calibration Tolerance or Accuracy will be included in the uncertainty calculation.
In this case the calibration is larger.4.4.6 Process Measurement Effects The transmitter is calibrated to compensate for a boron concentration of 3500 ppm. The Technical Specifications allow for a boron concentration between 2600 and 3500 ppm. If the boron concentration is less than 3500, the effect is that the transmitter will detect a lower differential pressure than if there is 3500 ppm boron in the tank. Therefore, the level will appear lower. The transmitter provides input to both the high and low values. The only non-conservative effect will be with the high level value.The process measurement effect will be subtracted from the high limit., The information is based upon Memo NES-87-10018 (Ref. 0). The specific gravity will be determined at the temperature of 68&deg;F and 120*F and boron concentrations of 2600 and 3500 ppm.p H 3 B0 3 = 1.436 g/cm 3 p H20 @ 68&deg;F = 0.99820 g/co 3 p H 2 Og 120OF = 0.99418 g/eam'M.W. H 3 B0 3 = 61.81 g/mole M.W. B = 10,81 g/mole Cone. B 3 B0 3 = 2600 * (61.81/10.81)
= 14866.420 ppm or 1.487 % sol Cone. H3aB0 3 = 3500 * (61.81/10.81)
= 20012.488 ppm or 2.001 % sol S.,G., 2 6 0 o = [((1.436 g/em 3)(0.01487))
+ ((0.9982 g/cm 3)(0.98513))]/0.9982 g/cm 3= 1.007 S.G.6J3500
= [((1.436 g/cm 3)(0.02001))
+ ((0.9982 g/cm 3)(0.97999))]/0.9982 g/em 3= 1.009 The effect at 68&deg;F in ft is (1.009/1.007) x 22 ft)- 22 fR = 0.044 ft.S.G.120/260
= [((1.436 g/cxn 3)(0.01487))
+ ((0.99418 g/cm 3)(0.98513))]/0.99418 g/cm,=1.007 S.G.1 2 0 6 0 0  = [((1.436/g/cm 3)(0.02001))
+ ((0.9418 g/cra 3)(0.97999))]/0.99418 g/cm 3= 1.009 The effect at 120&deg;F in ft is (1.009/1.007) x 22 ft)- 22 ft 0.044 ft.As can be seen, the temperature does not have an effect on the value. The calculated value will be rounded to 0.05 ft. This value will be added as a bias in the final calculated high values.
.. .........' ... ..P age 10 Fs nq CLCULATION COMPUTATION NOP-CC-3002-01 Rev. 00 CALCULATION NO.: C-ICE-051.01-001 TITLE I
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Jndication 4.4.7 Reference Leg Effects There is no reference leg effect for this instrumentation.
The CFT temperature will be the same as the reference leg because both temperatures are dependent on ambient temperature.
Since both the tank and reference leg temperatures are the same then there is no reference leg temperature effect to influence transmitter level operation.
4.4.8 Transmitter Uncertainty (LT-CF3A1, LT-CF3A2, LT-CF3BI, LT-CF3B2)(Assumption 2.6)Transmitter
= SRSS (Accuracy, Drift, Temp Effect, Static Press Effect, M&TE)= SRSS (0.167, 0.89, 0.77, 0.2, 0.045)= +/-1.206% span dp p 4.5 Current Buffer (LT-CF3AIA, LT-CF3A2A, LT-CF3BIA, LT-CF3B2A)(Ref. 31)Component I.D.: Device Type: Manufacturer:
Model No.: Input Signal Range: Output Signal Range: Accuracy: Ambient Temp. Range: Temperature Effect: Power Supply: LT-CF3AIA(EB3)
Current Buffer Bai]ey Controls, Inc.6624610-2222 1-5 Vdc+/-10 Vde+/-0.2% span 40 to 1400 F+01.82% span (Dev. from Ref. Over Normal Range)+/-24 Vdc +/-0.5% (Normal: 22.8 to 25.2 Vdc)4.5.1 Drift Effect Drift is considered equivalent to the reference accuracy of +/-0.2% span in lieu of any performance data.See Assumption 2.3.I,
... .....Page CALCULATION COMPUTATIO1N NOP-CC-3002-01 Rev. ,00 CALCULATION NO.: C-ICE-051.01
-001 TITLE I
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 4.5.2 Temperature Effect The temperature effect is only applied when the Current Buffer is beyond the normal temperature range. Since the current buffer is located in a controlled environment (e.g. the control room) there is no correction required for temperature effect for this instrument.
4.5.3 MTE Effect The calibration of the voltage converter module is with a DMM having an accuracy of 0.03% or equivalent.
4.5.4 Calibration Tolerance (Ref. 24-27)Per channel calibration procedures, Calibration Tolerance for the CFT Level Current Buffer Module is+ 0.2% span. The larger of the effects for calibration tolerance or reference accuracy will be used to determine the device uncertainty.
4.5.5 Current Buffer Total Current Buffer= SRSS (Accuracy, Drift, Temp Effect, M&TE)= SRSS (0.2, 0.2, 0.0, 0.03)= +/-0.284% span 4.6 Signal Conditioner (LY-CF3AlB)
(-K +E)(Ref. 31)Manufacturer:
Model No.: Input Signal Range: Output Signal Range: Accuracy: Ambient Temp. Range: Temperature Effect: Power Supply-Bailey Controls, Inc.6623695-2+/-10 Ydc 0-10 Vde+/-0.1% span 40 to 1400 F+/-0.25% span (Dev. from Ref. Over Normal Range)+24 Vdc +/-0.5% (Normal: 22.8 to 25.2 Vdc)
Page 12" Lf" CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 00 CALCULATION NO.:* C-ICE-051.01-001 TITLE /
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 4.6.1 Drift Effect Drift is considered equivalent to the reference accuracy of 10.1% span in lieu of any performance data.See Assumption 2.3.4.6.2 Temperature Effect The temperature effect is only applied when the Signal Conditioner is beyond the normal temperature range. Since the Signal Conditioner is located in a controlled environment (e.g. the control room) there is no correction required for temperature effect for this instrument.
4.6.3 MTE Effect The calibration of the Signal Conditioner module is with a DMM having an accuracy of 0.03% or equivalent.
4.6.4 Calibration Tolerance (Ref. 24-27)Per channel calibration procedures, Calibration Tolerance for the CFT Level Signal Conditioner Module is + 0.15% span. The larger of the effects for calibration tolerance or reference accuracy will be used to determine the to determine the device uncertainty.
4.6.5 Signal Conditioner Total Signal Conditioner
= SRSS (Cal Tol, Drift, Temp Effect, M&TE)= SRSS (0.15, 0.1, 0.0, 0.03)= 0.183% span 4.7 CFT Level Indicator (LI-CF3A1)(Ref. 31)Manufacturer:
Bailey Controls, Inc.Model No.: RY211X Input Signal Range: 0-10 Vde\,
Page 13 yCALCULATION COMPUTATION NOP-CC-300M-01 Rev. 00 CALCULATION NO.: C-ICE-051.01-001 TITLE I
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication Output Signal Range: Accuracy: Linearity:
Repeatability:
Readability:
Ambient Temp.: Temperature Effect: Power Supply;0-14 ft.+/-1.0% span+/-1.0% span+/-0.5% spanr 1.5 inches 40 to 140 0 F+/-0.01% output span/degree F (Deviation from calibrating conditions within normal span)118 +/-1.0 Vac (Normal: 107 -127 Vac)4.7.1 Drift Effect Drift is considered equivalent to the reference accuracy of +/-1.0% span in lieu of any performance data.See Assumption 2.3.4.7.2 Temperature Effect The temperature effect is only applied when the Level Indication is beyond the normal temperature range. Since the Level Indication is located in a controlled environment (e.g. the control room) there is no correction required for temperature effect for this instrument.
 
====4.7.3 Readability====
The CFT level indicators are located in the control room on panel C-5716. The vertical scale indicates 0-14 feet of level. Each major division is 2 feet and there are 3 minor divisions, each represent 6 inches. The reading error is typically Y2 of the minor division or 3 inches. Readability during the calibration introduces error in that the indicator can be as much as +/-3 inches from the actual input signal and still be within the calibration tolerance.
The calibration procedure wil specify performance of a reverse calibration, the indication is adjusted based on a specific level indication.
This removes the readability error from the uncertainty analysis.4.7.4 MTE Effect The calibration of the Level indication meter is with a DMM having an accuracy of 0.03% or equivalent.
 
COMPUTATION NOP-CC-3002-01 Rev. 00......CALCULATION NO.: C-ICE-051.01-001 TITLE !
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 4.7.5 Calibration Tolerance (Ref. 24-27)Per channel calibration procedures, Calibration Tolerance for the CFT Level Indication
+/- 1.0% span.The larger of the effects for calibration tolerance or reference accuracy will be used to determine the device uncertainty.
4.7.6 Level Indicator Total Level Indicator= SRSS (Accuracy, Drift, Temp Effect, Readability, M&TE)= SRSS (1.0, 1.0, 0.0, 0.0, 0.03)= +/-1.415% span 4.8 Voltage Converter (LT-CF3A1B, LT-CF3A2B, LT-CF3B1B, LT-CF3B2B) (EB2/EB4)(Ref. 31)Manufacturer:
Model No.: Input Signal Range: Output Signal Range: Accuracy: Ambient Temperature Range: Temperature Effect: Power Supply: Bailey Controls, Inc.6624610-1111 0-10 Vdc 0-10 Vdc+/-0.1% span 40 to 1400 F-0.25% span (Dev. from Ref Over Normal Range)+/-24 Vdc +/-0.5% (Normal: 22.8 to 25.2 Vdc)4.8.1 Drift Effect Drift is considered equivalent to the reference accuracy of +0. 1% span in lieu of any performance data.See Assumption 2.3.4.8.2 Temperature Effect The temperature effect is only applied when the Voltage Converter is beyond the normal temperature range. Since the Voltage Converter is located in a controlled environment (e.g. the control room) there is no correction required for temperature effect for this instrument.
Page 15 COMPUTATION ,NOP-CC-3002-01 Rev. 00 CALQULATION NO.: C-ICE-051.01-001 TITLE /
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 4.8.3 MTE Effect The calibration of the Voltage Converter module is with a DMM havingan accuracy of 0.03% or equivalent.
4.8.4 Calibration Tolerance (Ref. 24-27)Per channel calibration procedures, Calibration Tolerance for the CFT Level Voltage Converter Module is +/- 0.1% span. The larger of the effects for. calibration tolerance or reference accuracy will be used to determine the device uncertainty.
4.8.5 Voltage Converter Total Voltage Converter
= SRSS (Accuracy, Drift, Temp Effect, M&TE)= SRSS (0.1, 0.1, 0.0, 0.03)= +/-0.145% span 4.9 CFT Level switches (LS-CF3Al, LS-CF3A2, LS-CF3B1, LS-CF3B2)(Ref. 31)Manufacturer:
Model No.: Input Signal Range: Output Signal Range: Accuracy: Repeatability Hysteresis (switching)
Ambient Temperature Range: Temperature Effect: Power Supply: Bailey Controls, Inc.6623819-1 0-10 Vdc (0-14 ft.)Contact Logic+/-0.25% span+/-0.1% span+/-0.05% span 40 to 1400 F+/-0.25% span(Over ambient temperature range)+/-24 Vdc +/-0.5% Normal: 22.8 to 25.2 Vdc 4.9.1 CPT Level Switches LS-CF3AI, LS-CF3A2, LS-CF3B1, LS-CF3B2; Drift Effect Drift is considered equivalent to the reference accuracy of +0.25% span in lieu of any performance data. See Assumption 2.3..1 Page 16 CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 00 CALCULATION NO.: C-ICE-051
.01-001 TITLE&#xfd; I
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 4.9.2 Temperature Effect The temperature effect is only applied when the Level Switch is beyond the normal temperature range.Since the Level Switch is located in a controlled environment (e.g. the control room) there is no correction required for temperature effect for this instrument.
4.9.3 MTE Effect The calibration of the Level Switch module is with a DMM having an accuracy of 0.03% or equivalent.
4.9.4 Calibration Tolerance (Ref. 24-27)Per channel calibration procedures, Calibration Tolerance for the CFT Level Switch Module is :: 0.1%span. The larger of the effects for calibration tolerance or reference accuracy will be used to determine the to determine the device uncertainty.
4.9.5 Level Switch Total CFT Level Switches = SRSS (Accuracy, Drift, Temp Effect, M&TE)= SRSS (0.25, 0.25, 0.0, 0.03)= +/-0.355% span 4.10 Computer Point -Multiplexer and A/D Conversion Card (Ref. 34)Manufacturer:
Model:
 
== Description:==
 
Input Range: Output Range: Accuracy-Manufacturer:
Model:
 
== Description:==
 
Input Range: Modcomp 1873-1 Multiplexer Autoranging
* 10mV* 0.05%Modcomp 1870-1 Modacs Basic Card Autoranging
~Page 17 EneW CALCULATION COMPUATION~N.O.P-CC-3002-,01 Rev. 00, CALCULATION NO.: C-ICE-051.01-001 TITLE /
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication Output Range: 12 bit digital Accuracy:
+ 0.05%4.10.1 Reference Accuracy (Ref 34)AID-Multiplexer SRSS (A/D, Mux)= SRSS (.05, .05)= 0.07% span Since the accuracy of the A/D-Mux (+ 0.07%) is an order of magnitude less than the calibration tolerance of the string, it will be neglected.
The 12-bit A/D is auto ranging which results in a quantizing error equal to V 2 of the Least Significant Bit (LSB). This is similarly a small error that will not be considered.
4.10.2 Drift The Multiplexer/A-D pair is fully autoranging and not subject to drift. (Ref 34)4.10.3 Power Supply Effects The Multiplexer and A-D card (1870/1873) are capable of a withstanding common mode voltage of up to +_ 200 Vdc or up to 120 Vac, and common mode noise rejection up to 132dB at up to 60 Hz, without a loss of accuracy.
Thus, variations in power supply voltage and frequency are estimated to have an effect bounded by the accuracy in terms of minor board signal variations.(Ref. 34)4.10.4 Temperature Effects (Ref. 35)During normal conditions, the Computer Room Temperature is controlled at or very near 75 F.As such, no additional drift will be assumed due to temperature drift. It should be noted that the output of these cards changes relatively little due to temperature variation (approximately
.01%/C) and so even a large variation of :h 10 F would not significantly change their output.
,, l i l i 1l1Page 18 F tne" CLCULATION COMPUTrATION NOP-CC-3002-01 Rev. 00 CALCULATION NO.: C-ICE-051.01-001 TITLE
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 4.10.5 Humidity Effects There is no published humidity effect data on the Multiplexer/A-D card pair. The cards are normally maintained in a temperature and humnidi ty controlled environment.
Humidity effects are estimated to be negligible.
4.10.6 Pressure Effects The Multiplexer/A-D card pair is installed in the computer room, which is maintained at atmospheric pressure, or slightly higher. There is no pressure effect.4.10.7 Vibration Effects The Multiplexer/A-D card pair is installed in the computer room, which is not subject to vibration during normal operation.
This instrument is not required to operate during a seismic event. There is no vibration effect.4.10.8 Calibration Tolerance These Multiplexer/A-D modules are "string" calibrated together, and will be treated as a single module. The larger of the effect for Calibration Tolerance or Accuracy will be included in the uncertainty calculation. (In this case, the Calibration Tolerance is larger.) (Ref. 15, Section 6.2.6.2).
From the data sheet: tolerance
-0.5% span.* 4.10.9 M&TE Since the M&TE (DMM) accuracy is < 1/10 of Ref. accuracy it will be neglected.
4.10.10 Indicator Readability The digital indication is readable to the tenths digit, which is a negligible contribution.
4.10.11 Multiplexer/A-D Card Uncertainty Computer Loop Uncertainty
= Calibration Tolerance 0- I).5% span 4 Page 19 CACUATON COMPUTATIN NOP-CC.-3002-0`1 Rev. 00" CALCULATION NO.: C-ICE-051,01-001 TITLE /
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 5.0 RESULTS 5.1 Overall level channel uncertainty evaluation Per the ISA error combination methodology (Ref. 15) the appropriate combination technique for independent random uncertainties in % span is a square-root-sum-of-the-squares.
The uncertainties will be determined below for CFT level indication and alarms during normal environmental conditions.
Instrument Instrument ID (Module #) Uncertainty Section Level Transmitters LT-CF3A1, A2, B1, & B2 +/-1.206% span 4.4.7 Current Buffer LT-CF3AIA, A2A, BIA, & B2A (IB4/1B3)
+/-0.284% span 4.5.5 Signal Conditioner
(-K -+E) LY-CF3A1, A2, B1I, & B2 +/-0.183% span 4.6.5 Level Indicator LS-CF3AI, A2, B1, & B2 +/-1.415% span 4.7.6 Voltage Converter LT-CF3A1A, A2A, B1A, & B2A (EB4/EB2)
+/-0.145% span 4.8.5 Level Switch LS-CF3AI, A2, BI, & B2 +/-0.355% span 4.9.5 Multiplexer/A-D Cony L080, L085, L079, L090 +/-0.50% span 4.10.11 Indicator
= SRSS (Xmtr, Current Buffer, Signal Cond., Level Indicator)
= SRSS 0.206, 0.284, 0.183, 3,415)= +/-1.890% span
* 14 fspan 0.265 feet Computer = SRSS (Xmtr, Current Buffer, Signal Cond., Voltage Converter, Mux/A-D)= SRSS (1.206, 0.284, 0.183, 0.145, 0.5)= +/-1.356% span
* 14 fl/span=0.190 feet Alarm= SRSS (Xmtr, Current Buffer, Signal Cond., Level Switch)= SRSS (1.206, 0.284, 0.183, 0.355)= +/-1.302% span
* 14 ftspan=+0.182 feet f Page 20 FuOSC-eMMV CALCULATION COMPUTATION NOP-CC-3002-01 Rev. 00 CALCULATION NO.: C-ICE-051.01-001 TITLE I
 
==SUBJECT:==
Uncertainty Calculation for Core Flood Tank Level Indication 5.2 Surveillance and Alarm Setting Criteria The Analytical Limits for the CFT are 7480-8078 gallons based on 1040 cu. ft. +/- 40 (Ref 36).Evaluating these volumes per C-ME-5 1.01-086 (Ref. 14) results in a level requirement of 12.412 to 13.588 feet. Applying the above calculated uncertainties yields the following surveillance/setting criteria and including the PMA of -0.05 ft for the high values results in: 1 Indicators Computer: 13.588' -0.265' -0.05'12.412' + 0.265'13.588' -0.190' -0.05'12.412' + 0.190'= 13.273'= 12.677'= 13.348'= 12.602'= 13.356'= 12.594'High alarm setpoint:
13.588' -0.182' -0.05'Low alarm setpoint:
12.412' + 0.182'Applying conservative rounding and indicator readability yields the surveillance values/settings:
Indicators:
Computer: Alarm Sitting:< 13.25,;> 12.75 feet< 13.3, > 12.6 feet 13.3, > 12.7 feet
 
==6.0 CONCLUSION==
S Based on the increased accuracy and readability, the surveillance procedures should be revised to include the computer indication as the preferred input. The control board indicator should be the backup. The current field settings for the alarms are supported by this evaluation.
However, a slight increase in operational margin could be gained by adjustment to the calculated values -removing the excess conservatism.
NRC ITS Tracking-Page I of I Return to View.Menu.
a Print Document RAI Screening Required:
Yes This Document will be approved by: Greg Cranston This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC ITS TRACKING NRC Reviewer ID ]200712201259 Conference Call Requested?
No Categoy J BSI -Beyond Scope Issue ITS S~ection0:
TB POC.: JrD WNumbe~r:
Palge.Number(s):;
ITS 3.5 Ross Telson None Information ITS.Number:
OSI:. DOC Number: Bases. JFD...Number:.
3.5.1 5 L.1 None In the CTS, Bases 3.5.1 explains the importance of the minimum / maximum CFT volumes and the minimum / maximum nitrogen cover pressures for the current safety analyses performed.
The staff requests the licensee to explain/demonstrate that the current safety analyses are still applicable to the new Comment values of the specified parameters that are being proposed in the submittal.
Provide a justification for your conclusion with these proposed values. Also, with the proposed changes in the specified parameters, demonstrate that the discharge flow rates remain acceptable with the applicability of the system remaining the same (with reactor coolant pressure > 800 psig).Issue D.ate 112/20/2007 IClose otl 02/08/2008 Logged in User: Anonymous"'Responses Licensee Response by Bryan The response posted to NRC question 200710032123 by Jerry Kays on 01/13/2008 Jones on 12/26/2007 has altered the portions of ITS 3.5.1 for the CFT pressure and volume limits. Davis-Besse believes that this question is now addressed by the new markups and posted_calculations to NRC question 200710032123.
Date Created: 12/20/2007 12:59 PM by Jason Paige Last Modified:
02/08/2008 03:01 PM http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/I fddcealOd3bdbb585256e...
7/17/2008 NRC ITS Tracking Page I of I Return to View Menu Prit:D=1i~n RAI Screening Required:
Yes This Document will be approved by: Greg Cranston This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC Reviewer ID] 200712201301 Conference Call Requested?
No Category]
BSI -Beyond Scope Issue ITS Section: TB.PO.C:.4 JED iNu.m.-ber:
Page ..Number(s):, ITS 3.5 Ross Telson None Information ITS Number: OSI: DO C Numbe-r: Bases JFD Num-bcr:: 3.5.1 5 L.1 None With the increase in volume, provide a justification for the proposed/
existing Comment concentration of boron to ensure a safe shutdown of the reactor.Is sueDate.]
12/20/2007 Close Date [ 02/08/2008 Logged in User: Anonymous'vResponses Licensee Response by Bryan The response posted to NRC question 200710032123 by Jerry Kays on 01/13/2008 Jones on 12/26/2007 has altered the portions of ITS 3.5.1 for the CFT pressure and volume limits. Davis-Besse believes that this question is now addressed by the new markups and posted calculations to NRC question 200710032123, as the volume is no longer increased, but is in fact more restrictive.
Date Created: 12/20/2007 01:01 PM by Jason Paige Last Modified:
02/08/2008 03:01 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal 0d3bdbb585256e...
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3.5.1 5 L.1 None Comment With the decrease of the nitrogen maximum pressure, provide a discussion/
justification regarding how it will affect the LOCA analyses for small breaks.Issue &#xfd;Date 12/20/2007 Clo.se .D~ate ] 02/08/2008 Logged in User: Anonymous'Responses Licensee Response by Bryan The response posted to NRC question 200710032123 by Jerry Kays on 01/13/2008 Jones on 12/26/2007 has altered the portions of ITS 3.5.1 for the CFT pressure and volume limits. Davis-Besse believes that this question is now addressed by the new markups and posted calculations to NRC question 200710032123, as the nitrogen_pressure is no longer decreased, but is in fact more restrictive.
Date Created: 12/20/2007 01:03 PM by Jason Paige Last Modified:
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0.S1:. D.OC N.umber: , Ba s.es JFD Number.: 3.5.1 None None None Attachment 1, Volume 10, rev. 0, Page 16, 17, 18, 23 of 98---Regarding Proposed deviations from NUREG B3.5.1 APPLICABLE SAFETY ANALYSES---Request: Please clarify B-JFDs associated with the following changes: 1. Substitution of phrase "In the LOCA analysis, HPI and LPI are not cedited...
Safety Features Actuation Systems (SFAS) signal." In place of the phrase "As a conservative estimate, no credit is taken... the ESFAS actuation pressure." B-JFD 1 provides insufficient information. (e.g. Do applicable Davis-Besse safety analyses take credit for HPI 40 seconds after SFAS actuation whereas the typical B&W plant modeled in the NUREG takes no Comment credit for HPI for large break LOCAs?).................
: 2. Deletion of phrase "IN addition to LOCA analyses, the CFTs have been assumed...
large steam line break (SLB)." B-JFD 1 provides insufficient information. (e.g. Are the D-B the analyses and evaluation included in the UFSAR different than the typical B&W plant modeled in the NUREG in that CFTs are not assumed to operate under these circumstances?)
: 3. Replacement of the phrase "between 7.0 and 11.0" with the phrase "at a relatively high pH." B-JFD 3 provides insufficient information.
: 4. Replacement of the word "all" with the phrase "50% of the" in referring to those control rod assemblies assumed not to insert following a large break LOCA. B-JFD 1 provides insufficient information.
---Basis for Request: Applicable B-JFDs provide insufficient information for ITSB reviewer to determine:
http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsfl 1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 4 1. Whether deviations impact ITS in a manner warranting further review by organizations external to ITSB that exercise functional responsibility.
: 2. Whether plant-specific deviations affect the completeness of the ISTS---Applicable References NEI 96-06 -Improved TS Conversion Guidance, 2.7 Deviations from the Applicable ISTS: "...a high threshold should be satisfied for deviating from the ISTS..." 58 FR 29132 (pp 39132-39139)
Final Policy on &sect; 50.36 Technical Specifications, IV. The Commission Policy: "...it is the Commission intent that the wording and Bases of the improved STS be used in the Technical Specification related submittal to the extent practicable."---Regulatory Requirements:
-&sect; 50.36 Technical Specifications (a) Each applicant for a license ... shall include ... proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications
... shall also be included in the application, but shall not become part of the technical specifications.(b) ... The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to &sect; 50.34. The Commission may include such additional technical specifications as the Commission finds appropriate.(c)(3) Surveillance requirements...
assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.Issue Date 02/14/2008 Clo~se .D~at~e 06/05/2008 Logged in User: Anonymous' Responses-I Licensee Response by Bill Bentley on 03/08/2008 Response 1 Substitution of the phrase in question in the Applicable Safety Analyses (ASA) section of the Bases (Volume 10, Page 16)was made to correctly reflect the UFSAR licensing basis description.
UFSAR Section 6.3.3.2.3, Acceptable Lag Times states "The current LOCA analysis uses... High and Low Pressure Injection delays of greater than or equal to 40 seconds." Furthermore, NRC Question 200802141645, Item #11 addressed the subject of crediting the High Pressure Injection System in the LOCA analysis.
Response 2The main steam line break analysis is proyided in UFSAR 15.4.4. The High Pressure Injection System is assumed for reactivity control, not the Core Flooding Tanks.Response 3 The change to the ASA section of the Bases (Page 18)was made based on wording in the Davis-Besse system description documents for the Core Flooding Tanks. However, UFSAR 9.3.3.2, Post-LOCA Sump pH-Control, supports the ISTS wording. Therefore this change will not be made and the ISTS Bases words will be maintained.
A draft markup regarding this change is attached.
This change will be reflected in the supplement http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsfY 1 fddcealOd3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 3 of 4 to this section of the ITS Conversion Amendment.
Davis-Besse also notes that the change should have been justified using JFD 1, not JFD 3. JFD 3 is justifying the deletion of the brackets around the 3500 ppm value and will be maintained in the markup.Response 4 This similar question (concerning the amount on control rods assumed to be inserted following a large break LOCA) was asked as part of NRC question 200802121126.
Please see the Davis-Besse response for NRC question 200802121126 (Response 2).NRC Response by Ross Telson on 04/15/2008 To close this question thread, please address the following:
1.Please confirm the reviewer's summarized understanding below or provide correction, as appropriate.
The reviewerunderstands that the Davis-Besse has determined:
(1) The D-B design basis differs from the design basis reflected in the subject portions of B&W STS Rev 3.1, (2) that these differences justify the proposed subject deviations from the STS, and (3) that the proposed ITS and Bases changes are derived from the analyses and evaluation included in the Davis-Besse safety analysis report, and amendments thereto, submitted pursuant to &sect; 50.34. Specifically, please confirm that the analyses and evaluation included in the Davis-Besse UFSAR: a.Credit HPI and LPI beginning 40 seconds after actuation of the associated SFAS signal. b. Do NOT assume the CFTs to provided borated water for reactivity control for severe overcooling events such as a large steam line break (SLB). c. Credit insertion of only 50% of the control rods during a LB LOCA. 2. While reviewing this question for closure, a question arose regarding the licensee-proposed revision to Pg. 18 which is included in the scope of this question.
Pg. 18 was modified in response to Q200710032123 which has since been closed. Rather than opening a new question thread, the reviewer requests the licensee to clarify discussion on the revised Pg. 18 to explicitly address the correlation between the proposed INDICATED CFT level limits (112.6 ft and 13.3 ft) and: (i) the corresponding INDICATED CFT borated water VOLUME limits, (ii) the ACTUAL CFT maximum and minimum water VOLUME assured by the INDICATED CFT level limits, and (iii)the maximum and minimum ANALYTICAL limits assumed in the analyses and evaluation included in the UFSAR. 3. Please provide a properly consolidated markup of Attachment 1, Volume 10, Rev.0, Page 18 of 98. As cautioned by the reviewer in one or more different question threads, the licensee's practice of submitting Attachment 1 markups without revising the associated Rev.number could contribute to loss of revision control. Currently, in addition to the original, there are at least two (2) revised versions of Attachment 1, Volume 10, Rev. 0, Page 18 of 98. The reviewer notes that the markup of Attachment 1, Volume 10, Rev. 0, Page 18 of 98 associated with your 12-26-07 response to Q200710032123 is not reflected in your markup of the same page associated with your 3-8-08 Response 3 to this question thread.Licensee Response by Jerry Jones on 04/25/2008 Question 1 response:
The reviewers summarized understanding is correct. Question 2 response:
A sentence will be added to the http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 4 of 4 Bases that is believed to be sufficient to address the reviewer request. The following sentence will be added to the Bases discussion on Volume 10, Page 18, third paragraph, just prior to the sentence that lists the Technical Specification limits for Core Flood Tank levels: "The analytical limits for CFT volume are 7480 gallons (approximately 12.4 feet) and 8078 gallons (approximately 13.6 feet)." A draft markup regarding this change is attached and supersedes the previous draft markup provided in the 3/08/2008 response.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
Question 3 response: All markups that are submitted on the database correspond precisely to a specific RAI, as shown in the margins. There could be multiple questions affecting the same page. It would cause confusion if we attempted to maintain the markups updated for all RAI responses.
Therefore, each markup shows how Rev 0 is changed for just that specific RAI. The Revision 1 docketed submittal to each volume will consolidate all RAI markups. We do not desire to change our current practice.
However, since this new question thread was developed from question 200710032123, the attached markup includes both the 200710032123 changes for Page 18 from the 12/26/07 response, and all additional markups needed for question 200802140950, as requested by the NRC reviewer Licensee Response by Jerry During a phone conversation between the NRC reviewer and Jones on 05/23/2008 Davis-Besse personnel on 5/2 1, the NRC reviewer brought up a concern over the use of the term "approximately" in the draft markup of the Bases provided in the Davis-Besse response of 4/25/2008.
Davis-Besse has looked at the use of the term, and has decided to delete the term and replace the values with the specific values from the calculation.
A draft markup regarding this change is attached and supersedes the previous draft markup provided in the 4/25/2008 response.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Ross Telson Thank you for your response.
The reviewer has no further.on 06/05/2008 questions for this question thread at this time. Should unanticipated questions arise in the future, another question thread may be opened at that time.Date Created: 02/14/2008 09:50 AM by Ross Telson Last Modified:
06/05/2008 02:04 PM http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1 fddcea lOd3bdbb585256e...
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No Categor In Scope ITS Section: TB POC.:. JFD Number.: Page.Number(s).:.
ITS 3.5 Ross Telson None 34 Information ITS_.Number:.
0.51-: D.OC N.umber: BasesJFD Number, 3.5.2 None M.1 None In Attachment 1, Volume 10, Rev. 0, Page 34of 98, M01 line 5 references Comnment Action B.2. Should this be C.2?Similarly in line 7, should the reference to Action B.1. be C.A?Issue Date 01/15/2008.Close :D~ate l01/21/2008 Logged in User: Anonymous Licensee Response by Bryan The reference to ACTION B. 1 and B.2 in Discussion of Change.Kays on 01/20/2008 (DOC) MO1 (Volume 10, Page 34) should have been ACTION C. 1 and C.2, respectively.
A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Ross Telson Thank you for the correction.
In the future, it might be helpful if on 01/21/2008 the Rev No. on the header and footer of the attached document could be appropriately updated as well. There have been instances in which the document received multiple revisions.
Updating the Rev. No. with each revision will help prevent confusion.
No further questions at this time on this question thread.Date Created: 01/15/2008 03:49 PM by Ross Telson Last Modified:
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TB POC:. JFD.Numnber; Page Number(s):.
ITS 3.5 Ross Telson None 34 Information 0TS Nmb5er: OSI: DO.C Nummbe-r:
Ba s~es.JFD Numb her.: 3.5.2 None LA.1 None Attachment 1, Volume 10, Rev. 0, Page 34 of 98, paragraph 1, line 5 of LA01 states "...but the details of what constitutes an OPERABLE train are moved to the Bases. This changes the CTS by moving the details of what constitutes an OPERABLE train to the Bases." The reviewer notes that the referenced "details" (OPERABLE HPI & LPI pump, decay heat cooler, and flow path from BWST & containment sump)are addressed in a number of key areas beyond the ITS Bases:-ITS 1.1 Definitions (including OPERABLE-OPERABILITY)and
-ITS SR 3.5.2.1-8.
-Analyses and evaluation included in the UFSAR.LAO1 thus appears to OVERSTATE the significance of the TS Bases and UNDERSTATE the importance of other ITS elements, such as ITS 1.1,SR 3.5.2.1-8, and the UFSAR, in assessing what constitutes an OPERABLE train.10 CFR 50.36 describes TS bases as "A sumary statement of the bases or reasons for...specifications...but...not...
part of-the technical specifications." 50.36 further specifies that "technical specifications will be derived from the analyses and evaluation included in the safety analysis report..." Consider revising this and other such LA statements to correctly reflect the significance of the the ITS Bases vs other ITS and CLB elements, and and the UFSAR in the establishing OPERABILITY of SSCs.Maintaining a balanced perspective between these elements may reduce the likelihood of erroneous OPERABILITY conclusions in the future.Issue Date 101/15/2008 Close Date 04/13/2008 http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb5 85256e...
7/17/2008 NRC ITS Tracking Page 2 of 3 Logged in User: Anonymous'Resnonses Licensee Response by Jerry Jones on 03/10/2008 ITS 3.5.2 Discussion of Change (DOC) LAO1 (Volume 10, Page 34) discusses why the details of CTS LCO 3.5.2 are being removed. ISTS LCO 3.5.2 (Page 42) requires two ECCS trains to be OPERABLE.
In ISTS 3.5.2, the explanation of what constitutes an OPERABLE ECCS train is provided in the Bases. Therefore, during the development of the ITS 3.5.2 at Davis-Besse, the details given in CTS 3.5.2 were moved to the Bases to match the ISTS.The intent of Discussion of Change (DOC) LAO1 was to match to the NUREG-1430 Bases as closely as possible.
Furthermore, as stated in DOC LAO 1, these details are related to the system design and this type of information is not required to be included in the Technical Specifications to provide adequate protection of public health and safety. However, Davis-Besse has noted that a statement found in all other of these types of DOCs is inadvertently missing. A new sentence "The ITS still retains the requirement that two ECCS trains shall be OPERABLE," should have been included after the first sentence in the second paragraph.
For example, DOC LA02 includes this type of statement for relocating information concerning the method to perform a Surveillance to the Bases. This added sentence ensures the proper significance is placed on the ITS itself (i.e., the LCO statement and all other Technical Specification requirements that support the LCO statement
-i.e., Surveillances and the definition of OPERABLE).
A draft markup regarding this change is attached.This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
Davis-Besse has also reviewed other ITS submittals, including the latest to be approved by the-NRC (Beaver Valley -a plant also operated by FENOC), and noted that the manner in which these LA type DOCs are described (which include the added sentence) is consistent with the Davis-Besse LA DOCs. Therefore, Davis-Besse does not believe that the LA type DOCs need to be revised.NRC Response by Ross Telson on 03/21/2008 Thank you for your response.
The reviewer has no further questions on this question thread at this time. Should unanticipated questions arise, a new question thread may be initiated at that time.NRC Response by Ross Telson on 04/13/2008 Thank you for your response.
No further information is required at this time. Should an unanticipated question arise, another question thread may be opened at that time. The reviewer notes that a number of details describing what constitutes an OPERABLE SSC may be found in a variety of CLB documents, including the TS Bases and these details contribute to understanding OPERABILITY requirements for the SSC. OPERABILITY, however, by definition, exists only when the TS Limiting Conditions for Operation (LCO) are met. The LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility.
Further, the LCOs can only be met when all applicable SRs are met. Neither TS Bases nor CLB http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/I fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 3 of 3 documents, nor their interpretation should contradict or obfuscate these facts.Date Created: 01/15/2008 04:26 PM by Ross Telson Last Modified:
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NRC ITS TRACKING NRC Reviewer ID *200802051621 Conference Call Requested?
Yes C.a.tegoQy Other Technical Challenge ITS Section: TB POC: JFD Number: Page Numnb-er.(s)1:
ITS 3.5 Ross Telson 3 43 Information ITS Number: OSI: DOC.Number:
Bases JFD Number;3.5.2 None None 3 (Ref: Attachment 1, Volume 10, Rev. 0, Pages 43, 46, 55, and 59 of 98)Regarding iSTS SR 3.5.2.1 and its associated BASES which are not included in the proposed ITS 3.5.2 "ECCS-Operating" surveillance requirements.
iSTS SR 3.5.2.1 is a bracketed SR to verify, every 12 hours, the proper position of those ECCS valves, which if mispositioned, can disable the function of both ECCS trains and invalidate the accident analysis.---Request: Please further clarify JFD 3 with regard to the statement "This is consistent with current licensing basis." Does this statement mean (a) that Davis-Besse has no valves which, if mispositioned, can disable the function of both ECCS trains and invalidate the accident analyses, (b) incorporation of iSTS SR 3.5.2.1 would be considered to constitute an unwarranted backfit to existing Comment licensing requirements, or (c) does it mean something else?..............
---Basis for Request:-JFD 3 provides little useful information for the reviewer to assess whether plant-specific provisions affect the completeness of the ITS.-The CTS 3/4.5 Emergency Core Cooling Systems (ECCS) BASES for 3/4.5.2 and 3/4.5.3 ECCS Subsystems, states, in part: The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.
-According to STS SR 3.5.2.1 BASES, this bracketed SR and its 12-hr frequency are considered reasonable to verify that those valves, which can disable the function of both ECCS trains and invalidate the accident analyses, do not become mispositioned.
-According to &sect; 50.36 Technical Specifications (c)(3), surveillance http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/
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7/17/2008 NRC ITS Tracking Page 2 of 3 requirements...
assure that the necessary quality of systems' and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.-According to NRC Administrative Letter 96-04, the major objective of converting from CTS to iSTS is "to achieve as much consistency in the license requirements as possible, to the extent that the plant-specific design basis can conform with the related typical plant design reflected in the improved STS."---Regulatory Requirements:
&sect; 50.36 Technical Specifications (c)(3) Surveillance requirements...
assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.Issue Date 102/05/2008 Close Date[ 03/10/2008 Logged in User: Anonymous'Responses Licensee Response by Jerry Davis-Besse does not have any power operated valves, which if Jones on 02/13/2008 mispositioned due to an active failure, can disable the function of both ECCS trains and invalidate the accident analyses.
Thus, not including this SR in the Davis-Besse ITS is consistent with current licensing basis, since we do not currently have this requirement, nor do we need this requirement.
NRC Response by Ross Telson Staff understand that Davis-Besse justifies deviation from STS SR on 02/18/2008 3.5.2.1 on the following bases. Please confirm so this question can be closed. The stated STS SR 3.5.2.1 BASES refer to: 1. ...valves that are of type described in Reference 5: IE Information Notice 87-01, "RHR Valve Misalignment Causes Degradation of ECCS in PWRs," January 6, 1987. This reference is deleted from the STS in the proposed ITS. Please confirm this reference is deleted because Davis-Besse has no valves that are of type described in Reference 5? 2. Verification of proper valve position ensures that the flow paths from ECCS pumps to the RCS is maintained to ensure that valves cannot change position as a result of an active failure and invalidate accident analysis.
Please confirm that Davis-Besse has no such valves. (Note -The Bases does not specify that the valves be "power operated" only that they are in the flow path and are subject to an active failure that:can invalidate accident analysis.
It is not clear to the reviewer that only "power operated" can experience an active failure. E.g. Some valves are "flow-activated," others are "spring-activated," etc.Licensee Response by Bill Confirmed.
There are no valves whose active failure can invalidate Bentley on 03/10/2008 Iaccident analysis.NRC Response by Ross Telson Thank you for your response.
The reviewer has no further on 03/10/2008 questions on this question thread at this time and is thus closing the thread. Should unanticipated questions arise at a future date, another question thread will be opened.http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 3 of 3 Date Created: 02/05/2008 04:21 PM by Ross Telson Last Modified:
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TB, P.OC:. JFD.,.N.numb.er:
Page Number(s):z ITS 3.5 Ross Telson 3 43 Information TS Nu.mber:.
OS;: DOC QNu..mber:
Bases.JFD Number:.3.5.2 None None 3 (Ref: Attachment 1, Volume 10, Rev. 0, Pages 43, 56, and 59 of 98) Regarding iSTS SR 3.5.2.7 and its associated BASES which are not included in the proposed ITS 3.5.2 "ECCS-Operating" surveillance requirements but which is explicitly discussed in the CTS Bases. iSTS SR 3.5.2.7 is a bracketed requirement to verify, every eighteen months, the correct settings of stops for HPI stop check valves to ensure the valves are in proper position to prevent the HPI pumps from exceeding runout limits.---Request: Please further clarify JFD 3 with regard to the statement "This is consistent with current licensing basis." Does this statement mean (a) that Davis-Besse has no such stop check valves, (b) that the HPI pumps can not exceed runout limits regardless of the valve position settings, (c) that incorporation of iSTS Comment SR 3.5.2.7 would be considered to constitute an unwarranted backfit to..on t ....... existing licensing requirements, or (d) does it mean something else?---Basis for Request:-JFD 3 provides little useful information for the reviewer to assess whether plant-specific provisions affect the completeness of the ITS.-CTS 3/4.5 Emergency Core Cooling Systems (ECCS) BASES for 3/4.5.2 and 3/4.5.3 ECCS Subsystems, states, in part: The surveillance requirement for throttle valve position stops provides assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking I Page 2 of 6 to all injection points equal to or above that assumed in the ECCS-LOCA analyses.-According to iSTS SR 3.5.2.7 BASES, this bracketed SR ensures that these valves are in the proper position to prevent the HPI pump from exceeding its runout limit and the frequency is based on the need to perform ths SR under the conditions that apply during a plant outage and the potential for an unplanned transient if the SR were performed with the reactor at power.-According to &sect; 50.36 Technical Specifications (c)(3), surveillance requirements...
assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.-According to NRC Administrative Letter 96-04, the major objective of converting from CTS to iSTS is "to achieve as much consistency in the license requirements as possible, to the extent that the plant-specific design basis can conform with the related typical plant design reflected in the improved STS."---Regulatory Requirements:
&sect; 50.36 Technical Specifications (c)(3) Surveillance requirements...
assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.Issue Date 1 02/05/2008 IClose Datel[ 06/19/2008 Logged in User: Anonymous'Responses Licensee Response by Jerry As shown in USAR Figure 6.3-2, the High Pressure Injection Jones on 03/07/2008 (HPI) System does include stop check valves (valves 48, 49, 56, and 57). However, the bracketed ISTS SR 3.5.2.7 (Volume 10, Page 43) has not been included in the Davis-Besse ITS because the stop check valves are not stopped at a mid position (i.e., a position other than full open). The valves are full open when the HPI System is injecting into the cold leg loops. Since the Davis-Besse Current Technical Specifications do not include this requirement, the non-inclusion of this ISTS SR is consistent with current licensing basis. It should be noted that Davis-Besse CTS and ITS does include a similar requirement for the Low Pressure Injection System. Furthermore, the ISTS SR 3.5.2.7 is a bracketed SR, which indicates that plant-specific information should be included since the bracketed information is not common to a typical B&W plant. Of the three B&W plants that have converted it the ISTS format, only the first plant that converted to ISTS (Crystal River 3)adopted this bracketed SR (and they included the exact valve numbers that are in NUREG-143 0). The other two (ANO-1 and Oconee) did not adopt this SR for a reason similar to why Davis-Besse is changing the SR (i.e., they do not need mechanical stops).NRC Response by Ross Telson on 04/15/2008 Please address the following to close this question.
Potentially Impacted SR's: CTS SR 4.5.2 -Requires each ECCS subsystem to be demonstrated OPERABLE;
-CTS SR 4.5.2.a. -Verifies that http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 3 of 6 each valve in the flow path... is in its correct position;
-CTS SR 4.5.2.e. 1. -Verifies that each automatic valve in the flow path actuate to its correct position; and -CTS SR 4.5.2.g. -Verifies the correct position of each mechanical position stop for valves DH-14A and DH-14B. The proposed ITS seeks to omit: STS SR3.5.2.1-Verifies valves are in the listed position with power to the valve operator removed STS SR 3.5.2.7 -Verifies the correct settings of stops for HPI stop check valves. The reviewer requested clarification of JFD-3 for deviating from STS SR's listed above.The licensee justified omission of STS SR 3.5.2.7 primarily on the bases that (a) D-B currently operates with the stop check valves in full open position and (b) the CTS do not include this requirement.
In this context, please address the following questions:
Question 1: Please confirm that the licensee has determined:
(1) The D-B design basis differs from the design basis reflected in the subject portions of B&W STS Rev 3.1, (2) that these differences justify the proposed subject deviations from the STS, and (3) that the proposed ITS and Bases changes are derived from the analyses and evaluation included in the Davis-Besse safety analysis report, and amendments thereto, submitted pursuant to &sect; 50.34. Question 2: Would the ECCS subsystem be demonstrated OPERABLE (CTS SR 4.5.2) if, for the current HPI pumps, the HPI stops should be inadvertently repositioned (by any means -vibration, maintenance error, etc.) to OTHER than full open? Question 3: Should one or more HPI pumps be replaced (e.g. due to unacceptable degradation or failure) with one or more 'improved' pumps (still within the constraints of current design limits), is it conceivable that the stops for the HPI stop check valves might require repositioning to OTHER than full open? Question 4: Regarding omitted, STS SR3.5.2.1, the STS Bases state that misalignment of these valves could render both ECCS trains inoperable and thus invalidate the accident analysis.
Please confirm that the Davis-Besse ECCS Subsystem has such no valves. Question 5: Please reconcile your 3/7/08 statement
"...the non-inclusion of (STS) SR 3.5.2.7 is consistent with current licensing basis" and JFD-3 "This is consistent with current licensing basis." with the CTS 3/4.5 ECCS Bases for 3/4.5.2 which states, in part, that the surveillance requirement for throttle valve position stops provides assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1)prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3)provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.Licensee Response by Jerry Question 1 response:
The HPI pumps for Davis-Besse are different Jones on 04/24/2008 from the pumps for the plant design described in NUREG-1430, http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 4 of 6 Rev 3.1 (B&W STS) because that plant design has a combined Makeup System and HPI System. It is confirmed that the D-B design basis differs from the design basis reflected in the subject portions of the. B&W STS, these differences justify the proposed subject deviations from the B&W STS, and the proposed Davis-Besse ITS and Bases changes are derived from the analyses and evaluation included in the Davis-Besse safety analysis report.Question 2 response:
If a postulated event occurred to misposition the normally locked-open valves such that HPI flow was reduced below that required for the accident analysis, the HPI pump would be inoperable.
The safety analysis assumes a single failure such that only one of the two HPI trains operates to inject coolant into the RCS. Question 3 response:
It is theoretically possible that a modification could be created such that some flow throttling would be necessary.
For example, if Davis-Besse decided to install continuously operating combined makeup/HPI pumps similar to the other B&W plants, Davis-Besse would probably need to limit flow in a similar method to those plants (i.e., control valves set to balance the HPI flow to the RCS). However, the modification process would address the flow requirements, and would require applicable changes to the Technical Specifications.
Question 4 response:
This was confirmed in question 200802051621.
Question 5 response:
The CTS Bases paragraph in question used to contain the phrase "and flow balance testing" in the first sentence.Technical Specification Amendment 256 removed the phrase and relocated to the TRM the Surveillance Requirement to perform a flow balance test, during shutdown, following completion of modifications to the high pressure injection (HPI) or low pressure injection (LPI) subsystems that alter the subsystem flow characteristics.
Technical Specification Amendment 20 provides additional insight into the Surveillance Requirement that was relocated to the TRM by Amendment 256. The License Amendment Request (for Amendment
: 20) was submitted in response to an NRC letter (dated November 9, 1977) that discussed ensuring proper flow resistance and pressure drop in the HPI and LPI lines to the reactor following completion of modification to these systems. The CTS Bases is referring to CTS Surveillance Requirement 4.5.2.g to verify the correct position of each mechanical position stop for valves DH-14A and DH-14B.This requirement is maintained in the ITS as SR 3.5.2.6, and is the only CTS Surveillance Requirement for throttle valves position stops. Not including STS SR 3.5.2.7 in the Davis-Besse ITS is consistent with the current licensing basis.Licensee Response by Jerry During the weekly phone conversation of 5/21, the NRC reviewer Jones on 05/22/2008 raised a concern that, even though the Davis-Besse HPI System stop check valves are not mechanically stopped at a mid-position to protect the HPI pumps from a runout condition, that they might should be checked periodically to make sure they are in the correct (i.e., full open) position.
Thus, even though Davis-Besse does not have a current Surveillance to perform this check, the NRC/K http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 5 of 6 reviewer was considering the idea that it might be appropriate to add an appropriate SR into the ITS. Davis-Besse notes that ISTS SR 3.5.2.1, which requires verification that all ECCS valves in the flow path are in the correct position, specifically exempts valves that are locked open. As stated in the previous response, these HPI stop check valves are locked fully open. Therefore, they are exempt from any ECCS valve position checks of ITS SR 3.5.2.1.NRC Response by Ross Telson on 06/05/2008 Please clarify responses provided on 4/24/08 as follows: Question 2: The licensee's response "If a postulated event occurred to misposition the normally locked-open valves..." reflects a reliance on the licensee's configuration control program (e.g. blocking and tagging program) rather than a TS SR to assure the HPI stop check valves and their mechanical stops remain in the required state to assure ECCS OPERABILITY.
a) Is there a licensee commitment or requirement for these valves and/or mechanical stops to be locked, sealed, or otherwise secured in the correct position?
If so, what is the procedure and how is it controlled?
If not, what assurance is there that the HPI stop check valve mechanical stops will not experience changes to their settings?
b) Can the HPI mechanical stop settings be altered without unlocking, unsealing, or otherwise unsecuring the HPI stop check valves? Question 3: The reviewer understands that an ECCS pump "modification" would require a design change review to assess the need for repositioning the HPI stop check valve mechanical stops. The question intended to assess whether it was conceivable that a "like-for-like" HPI pump replacement (which generally does not require a design change review but would be expected to produce increased flow because the new pump would not be degraded by wear and age). Unless a "like-for-like" HPI pump swap could not credibly require repositioning of the HPI stop check valves and the licensee is confident that it would recognize the need to adopt STS SR 3.5.2.7 should a future plant modification warrant resetting the mechanical stops, it would appear inappropriate to omit STS SR 3.5.2.7. a) Is it not credible that HPI stop check valve mechanical stop setting changes will be required following a like-for-like HPI pump replacement or some other plant maintenance (other than modification) activity?
b) Assuming that STS SR 3.5.2.7 is not adopted at this time, is the licensee committing to adopt STS SR 3.5.2.7. should future plant maintenance or modification activities warrant resetting the mechanical stops? Question 4: The question asked whether "misalignment" of the valves could render both ECCS trains inoperable.
The response referenced question 200802051621 which confirmed only that there were no valves whose "active failure" could invalidate accident analysis.Omission of STS SR 3.5.2.1 is potentially-justified if, in the site-specific D-B design, there are no valves of the type described in STS Ref 5, which can disable the function of both ECCS trains and invalidate the accident analysis.
a) Are there any valves the type described in STS Ref 5, which can disable the function of both ECCS trains and invalidate the accident analysis?
Licensee's http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/lfddcealOd3bdbb585256e'..
7/17/.2008 NRC ITS Tracking Page 6 of 6 5/22/2008 response following a conference call: "Davis-Besse does not have a current Surveillance to..." verify the correct settings of the stops for the HPI stop check valves (e.g. STS SR 3.5.2.7 which D-B seeks to omit from the proposed ITS). The reviewer understands that Davis-Besse does not wish (nor consider it appropriate for the current site-specific design ) to adopt STS SR 3.5.2.7 because the D-B mechanical stop settings are maintained at full open. As such, the mechanical stops are in effect not-used /non-functional.
Staff will review the licensee responses to the previous questions prior to determining the acceptability of the proposed ITS with the deviation to omit STS SR 3.5.2.7.NRC Response by Carl Schulten The staff understands that at least two discrete actions are required on 06/18/2008 to reposition the locked open HPI stop check valve and that HPI flow path is verified by surveillance to be oeprable once per month. The D-B responses to this thread are acceptable with the__deviation to omit STS SR 3.5.2.7 NRC Response by Ross Telson No further questions at this time. (Carl Schulten)on 06/19/2008 Date Created: 02/05/2008 04:45 PM by Ross Telson Last Modified:
06/19/2008 09:16 AM http://www.excelservices.com/exceldbs/itstrack davisbesse.nsfl1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page I of 5 Return to View Menul Print Document RAI Screening Required:
Yes Status: Closed This Document will be approved by: Gerald Regulatory Basis must be included in Comments Waig; Carl Schulten section of this Form This document has been reviewed and Yes information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING NRC Reviewer ID 1200802141351 Conference Call Requested?
No Category In Scope ITS Section: TB.PO.C.:.
JFD. Number.:.
Page Numb.er(s):
ITS 3.5 Ross Telson 3 44 Information ITS.Number:
OS1I. DOC Number: Bases JFD Number: 3.5.2 None None None Attachment 1, Volume 10, rev. 0, Page 44, 46, 57, 59 of 98---Regarding Deviation from NUREG SR 3.5.2.8 (ITS SR 3.5.2.6) & JFD-3---Request: Pleaseclarify the JFD3 & BJFD3 for deviating from the NUREG by substituting, in SR 3.5.2.8 of the NUREG (SR 3.5.2.6 of ITS), the phrase"correct position of each mechanical stop for the following valves" for the phrase "flow controllers for the following LPI throttle valves operate properly," and associated changes to the BASES.---Basis for Request: Applicable B-JFDs provide insufficient information for ITSB reviewer to determine:
Comment 1. Whether deviations impact ITS in a manner warranting further review by organizations exercising functional responsibility
-external to ITSB.2. Whether plant-specific deviations affect the completeness of the ISTS---Applicable References NEI-96-06
-Improved TS Conversion Guidance, 2.7 Deviations from the Applicable ISTS: "...a high threshold should be satisfied for deviating from the ISTS..." 58 FR 29132 (pp 39132-39139)
Final Policy on &sect; 50.36 Technical Specifications, IV. The Commission Policy: ".., it is the Commission intent that the wording and Bases of the improved STS be used in the Technical Specification related submittal to the extent practicable."---Regulatory Requirements:
&sect; 50.36 Technical Specifications (a) Each applicant for a license ... shall include ... proposed technical http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 5 specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications
... shall also be included in the application, but shall not become part of the technical specifications.(b) ... The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to &sect; 50.34. The Commission may include such additional technical specifications as the Commission finds appropriate.(c)(3) Surveillance requirements...
assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.I lssueDate1 02/14/2008 Close 5De 106/04/2008 Logged in User: Anonymous'Responses Licensee Response by Jerry CTS 4.5.2.g (Volume 10, Page 31) requires verification of the Jones on 03/07/2008 correct position of the mechanical stops every refueling interval.This change was approved by License Amendment 20 (Adams Accession Number ML021160305).
In the NRC Safety Evaluation for this change, the NRC stated that maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary, and that at Davis-Besse this was accomplished by mechanical stops on the injection valves, in lieu of electrical stops. Davis-Besse does not use throttle valves to control Low Pressure Injection (LPI) System flow to the core. Valves DH-14A and 14B open upon an SFAS signal until they reach the mechanical stop positions, and flow is not'controlled by any flow controllers on throttle valves. USAR Figure 6.3-2A provides the one-line diagram of the LPI System, and valves DH-14A and 14B are included in the print. Therefore, changing the, bracketed wording (which indicates that plant-specific information should be included since the bracketed information is not common to a typical B&W plant) in SR 3.5.2.6 (Page 44) is consistent with the current Technical Specifications.
Furthermore, of the three B&W plants that have converted to the ISTS format, only the first plant that converted to ISTS (Crystal River 3) adopted this SR (and they included the exact valve numbers that are in NUREG-1430).
The other two (ANO-1 and Oconee) did not adopt this SR for a reason similar to why Davis-Besse is changing the SR.NRC Response by Ross Telson on 04/15/2008 Potentially Impacted SR: STS SR 3.5.2.8 -Verify the flow controllers for the following LPI throttle valves operate properly Proposed ITS SR 3.5.2.6 -Verify the correct position of each mechanical stop for valves DH-14A and DH-14B CTS SR 4.5.2.g-Verify the correct position of each mechanical position stop for valves DH-14A and DH-14B Question 1. Please provide the correct ML for License Amendment 20 (currently listed as ML021160305).
The listed ML brings up Amendment 21 which http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/lfddcealOd3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 3 of 5 does not address the subject Safety Evaluation.
Please confirm the following to close this question, or correct accordingly.
The reviewer understands that the Davis-Besse has determined:
(1) The D-B design basis differs from the design basis reflected in the subject portions of B&W STS Rev 3.1, (2) that these differences justify the proposed subject deviations from the STS, and (3) that the proposed ITS and Bases changes are derived from the analyses and evaluation included in the Davis-Besse safety analysis report, and amendments thereto, submitted pursuant to &sect; 50.34.Specifically, the reviewer understands that the subject valves (DH-14A and DH-14B) do not employ flow controllers.
Rather the valves provide a fixed throttling function by opening until they reach mechanical stops. Also, please indicate whether these valves are verified to stroke from closed to the mechanical stop, positions as part of either ITS SR 3.5.2.1 or ITS SR 3.5.2.4.Licensee Response by Jerry A copy of the NRC Safety Evaluation for Amendment 20 is Jones on 04/26/2008 provided as an attachment to this response.
Davis-Besse confirms that the NRC reviewer's understanding is correct. DH- 14A and DH-14B are tested as part of ITS SR 3.5.2.4, which requires verification that the valves actuate to the correct position on an actual or simulated actuation signal.NRC Response by Ross Telson on 05/16/2008 Applicable
 
==References:==
 
Pg. 31, CTS SR 4.5.2.g -Verification of the correct position of each mechanical position stop for valves DH-14A and DH-14B. 1. Within 4 hours following completion of the opening of the valves to their mechanical position stop or following maintenance on the valve when the LPI system is required to be OPERABLE.
: 2. At least once each REFUELING INTERVAL.
Pg. 32, CTS SR 4.5.2.h- Verification that each ECCS Pump's developed head... is greater than or equal to the required developed head, when tested pursuant to the requirements of Specification 4.0.5. Pg. 39 DOC LO5 -Deletion of 4.5.2.g. 1.requiring a test following repositioning or maintenance on a valve that may alter subsystem flow. NRC SE Amendment 20 referenced by Licensee' on 4/26/2008.
USAR Figure 6.3-2A Functional Drawing DHR / LPI System --------------
Discussion
&Questions:
The reviewer examined Amendment 20, which incorporated SR's for the subject LPI injection throttle valves DH-14A and DH-14B. The reviewer found that CTS SR 4.5.2.h approved in Amendment 20 contained explicit flow balance testing requirements and acceptance criteria which differed from those in the present (Amendment 256) CTS SR 4.5.2.h. CTS Bases at the time of Amendment 20 established that SR's (verifying both (1)throttle valve position stops and (2) flow balance) provided assurance that proper ECCS flows would be maintained in the event of a LOCA. The present CTS Bases (Amendment 241) state that this assurance is achieved by the SR for throttle valve position stops. Reference to flow balance testing was deleted. 1. Please confirm that flow balance testing is NOT required to provide-assurance that proper ECCS flows will be maintained in the event of a LOCA. CTS SR 4.5.2.h. currently requires that each ECCS http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 4 of 5 pump be "tested pursuant to the requirements of Specification 4.0.5." at a frequency of"at least once each REFUELING INTERVAL." ITS SR 3.5.2.2. requires each "ECCS pump's developed head at the test flow point is greater than or equal to the required developed head" at a frequency "In accordance with the Inservice Testing Program." No DOC is provided for the change in mapping CTS SR 4.5.2.h to ITS SR 3.5.2.2. 2. Please provide an appropriate DOC confirming that the wording of ITS SR 3.5.2.2 is an administrative change and constitutes no change in the requirement or controls established by CTS SR 4.5.2.h., or state otherwise, as appropriate.
The Licensee Response on 03/07/2008 stated "CTS 4.5.2.g (Volume 10, Page 31) requires verification of, the correct position of the mechanical stops every refueling interval.
This change was approved by License Amendment 20 (Adams Accession Number ML021160305)." The reviewer examined License Amendment 20 and observed that CTS 4.5.2.g, as approved, required verification "1. Within 4 hours following, completion of opening of the valves to their mechanical position stop or following completion of maintenance on the valve when the LPI system is required to be OPERABLE" AND "2. At least once per 18 months." 3. Given the unique nature of Davis-Besse's design in use of mechanical position stops, the reviewer understands that an STS deviation is justified.
However, the same unique design consideration raises unresolved questions regarding justification for deletion of requirement CTS SR 4.5.2.g.2.
Either retain CTS SR 4.5.2.g.2 or explain, in DOC LO5, what has changed between Amendment 20, when CTS SR 4.5.2.g.2.
was deemed necessary, and present.Licensee Response by Jerry Jones on 05/21/2008 Question 1: Davis-Besse confirms that periodic flow balance testing is not required to provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Question 2:.The Frequency for Specification 4.5.2.h (Volume 10, Page 32) is"when tested pursuant to the requirements of Specification 4.0.5." There is no Frequency of "at least once each REFULEING INTERVAL" associated with this Surveillance.
ITS SR 3.5.2.2 (Page 43), which is the same SR, has a Frequency of "In accordance with the Inservice Testing Program." CTS Specification 4.0.5 is called the Inservice Testing Program, as shown in Specification 5.5.7 (Volume 16, Pages 63 and. 64 for the CTS Markup and Page 85 for the ISTS Markup). DOC AO1 covers the number/title, since the change is purely an administrative change related to renumbering from CTS to ITS. Question 3: Based on a phone conversation with the NRC reviewer, the question was concerning the deletion of CTS 4.5.2.g. 1, not 4.5.2.g.2.
CTS 4.5.2.g. 1 Frequency was deleted to be consistent with the ISTS. Specifically, the previous B&W STS, NUJREG-0103, Rev. 4, included the same Frequency in Specification 4.5.2.g. 1. During development of the B&W ISTS (NUREG-1430), the NRC accepted that the Frequency did not need to be maintained in the Technical Specifications.
The purpose of the http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 5 of 5 Davis-Besse SR is the same as the purpose of the B&W STS SR -the words were changed in the Davis-Besse ISTS Markup for the LPI System only to match up with plant specific design.Furthermore, the CTS 4.5.2.g. 1 Frequency was added as part of Amendment 20 to match up with the original B&W STS. Now that the Frequency is no longer required; as documented by it being not included in the B&W ISTS, Davis-Besse does not believe that we should maintain this Frequency.
The justification is provided in DOC L05 NRC Response by Ross Telson on 06/04/2008 Thank you for your response.
The reviewer has no further questions on this question thread at this time. Should unanticipated questions arise in the future, a new question thread may be opened at that time.* Date Created: 02/14/2008 01:51 PM by Ross Telson Last Modified:
06/04/2008 08:46 AM http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/lfddcea1Od3bdbb585256e...
7/17/2008 0r 3 670 01 6~3?I UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 October 2, 19 Docket No.: 50-346'79 Mr. Lowell E. Roe Vice President, Facilities Development Toledo Edison Company Edison Plaza 300 Madison Avenue Toledo, Ohio 43652&:A,9/
 
==Dear Mr. Roe:==
The Commission has issued the enclosed Amendment No. 20 to Facility Operating License No. NPF-3 for the Davis Besse Nuclear Power Station, Unit No. 1. The amendment consists of changes to the Technical Specifications in response to your application dated January 13, 1978 and staff discussions.
This amendment modifies the Technical Specifications to incorporate surveillance requirements for throttle valves used in the low pressure injection system.Copies of the Safety Evaluation and the Notice of Issuance are also enclosed.Sincerely, Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
 
==Enclosures:==
: 1. Amendment No. 20 2. Safety Evaluation
: 3. Notice cc w/enclosures:
See next page 0 0 7 0 0 1 '6 9 2.I(4A UNITED STATES NUCLEAR REGULATORY COMMISSION 0 WASHINGTON, D. C. 20555 THE TOLEDO EDISON COMPANY AND THE CLEVELANDiELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 IIAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No, 20 License No. NPF-3 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by The Toledo Edison Company and The Cleveland Electric Illuminating Company (the licensees) dated January 13, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I;B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (I) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public;and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements--
have been satisfied.
'. .4.n367 0f 6-2-2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 20. are hereby incorporated in the license. The Toledo Edison Compapy shall operate the facility in accordance with the Technical Specifications.
: 3. This license amendment is effective as of the date of its issuance.FOR THE NUCLEAR REGULATORY COMMISSION Ro er W.Ri, Chief Operating Reactors Branch #4 Division of Operating Reactors
 
==Attachment:==
 
Changes to the Technical Specifications Date of Issuance:
October 2, 1979 0 n.3670 011698 IJ4 0, UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON.
D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 20 TO FACILITY OPERATING LICENSE NO. NPF-3 THE TOLEDO EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346 Introduction By letter dated January 13, 1978, the Toledo Edison Company (TECO or the licensee) requested amendment to Facility Operating License No. NPF-3.The amendment would modify the Technical Specifications for-Davis-Besse Nuclear Power Station, Unit No. 1 (DB-I), to incorporate surveillance requirements for throttle valves used in the low pressure injection system.Discussion and Evaluation The High and Low Pressure Safety Injection system (HPSI and LPSI) designs of many Pressurized Water Reactors (PWR) utilize a commnon low pressure and a common high pressure header to feed the several cold (and in some cases hot) leg injection points. Maintenance of proper flow resistance and pres-sure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration; (2) provide a proper flow split between injection points in accordance with the assumptions used, in the Emergency Core Cooling System -Loss of Coolant Accident (ECCS-LOCA) analyses; and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.On many plants, there are motor operated valve(s) in the lines to each injection point that have stops which are set during preoperational flow testing of the plant to insure that these flow requirements are satisfied.
On other plants, electrical or mechanical stops on the Safety Injection System's isolation valve(s) are used for this purpose. DB-l utilizes mechanical stops to satisfy these ECCS flow requirements.
While preoperational HPSI/LPSI flow testing is utilized to assure that the valves used to throttle flow have been properly set, we have concluded that periodic surveillance requirements are needed to assure that these settings are maintained throughout the life of the plant. Consequently, we requested all PWR licensees to propose changes to their Technical Specifications, as appropriate, to incorporate periodic surveillance require-ments for these valves. Sample surveillance requirements, developed by the NRC staff, were provided to licensees for guidance in developing proposed changes.
.. n 67 0 0 1 6 9 9-2-The sample requirements include periodic verification of throttle valve position stop settings and verification of proper ECCS flow rates when-ever system modifications are made that could alter flow characteristics.
The request for proposed Technical Specification changes was sent to TECO on November 9, 1977.TECO responded to our request with respect to DB-l by submittal dated January 13, 1978. We discussed the submittal with the licensee and they agreed to modifications which would specify the parameter of operation during the flow balance tests. This submittal, as modified, contained proposed changes to the Technical Specifications that are in agreement with our requirements.
Based on our review, we have concluded that TECO's proposed increased surveillance requirements would provide sufficient additional assurance that proper valve settings for ECCS flow and flow distributions will be maintained throughout plant life; and thus, the proposed changes are acceptable.
Environmental Consideration we have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact, and pursuant to 10 CFR &sect;51.5(d)(4), that an environmental impact'statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of the amendment.
Conclusion We have concluded, based on the considerations discussed above, that: (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the commnon defense and security or to the health and safety of the public.Date: October 2, 1979 3 7 0 1 7 0 0 7590-01 UNITED STATES NUCLEAR REGULATORY-COMMISSION DOCKET NO. 50-346 THE TOLEDO EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE The U. S. Nuclear Regulatory Commission (the Commission) has issued.Amendment No. 20 to Facility Operating License No. NPF-3, issued to The Toledo Edison Company and The Cleveland Electric Illuminating Company (the licensees), which revised Technical Specifications for operation of the Davis-Besse Nuclear Power Station, Unit No. 1 (the facility) located in Ottawa County, Ohio. The amendment is effective as of its date of issuance.The amendment modifies the Technical Specifications to incorporate surveillance requirements for throttle valves used in the low pressure injection system.The application for the amendment complies with the standards and requirements'of the Atomic Energy Act-of 1954, as amended (the Act), and the Commission's rules and regulations.
The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 1.0 CFR Chapter I, which are set forth in the license amendment.
Prior public notice of this amendment was not required since the amendment does not involve a significant hazards consideration.
7590-01-2-The Commission has determined that the issuance of this amendment will not result in any significant environmental impact and-that pur-suant to 10 CFR &sect;51.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with iAsuance of this amendment.
For further details with respect to this action, see (1) the appli-cation for amendment dated January 13, 1978, (2) Amendment.No.
20 to License No. NPF-3, and (3) the Commission's related Safety Evaluation.
All of these items are available for public inspection at the Commission's Public Document Room, 1717 H Street, N.W., Washington, D.C., and at the Ida Rupp Public Library, 310 Madison Street, Port Clinton, Ohio.A copy of items (2) and (3) may be obtained upon request addressed to the U. S. Nuclear Regulatory.Commission, Washington, D.C. 20555, Attention:
Director, Division of Operating Reactors.Dated at Bethesda, Maryland, this 2nd day of October 1979.FOR THE NUCLEAR REGULATORY COMMISSION Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
--4 0 36 -7 0 1 ,)9 -1%9" ATTACHMENT TO LICENSE AMENDMENT NO. 20 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
Pages* 3/4 5-5 3/4 5-5a (added)B 3/4 5-1 B 3/4 5-2 i EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: 5. Verify that a representative sample of TSP from a TSP storage basket has a density of > 53 lbs/cu ft.6. Verifying that when a representative sample of 0.35 + 0.05 lbs of TSP from a TSP storage basket is submerged, without agitation, in 50 + 5 gallons of 180 + 10F borated water from the BWST, the pH of the mixed solution is raised to> 6 within 4 hours.e. At least once per 18 months, during shutdown, by 1. Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.2. Verifying that each HPI and LPI pump starts auto-matically upon receipt of a SFAS test signal.f. By performing a vacuum leakage rate test of the watertight enclosure for valves DH-11 and DH-12 that assures the motor operators on valves DH-Il and DH-12 will not be flooded for at least 7 days following a LOCA: 1. At least once per 18 months.2. After each opening of the watertight enclosure.
: 3. After any maintenance on or modification to the watertight enclosure which could affect its integrity.
: g. By verifying the correct position of each mechanical position stop for valves DH-14A and DH-14B.1. Within 4 hours following completion of the opening of the valves to their mechanical position stop or following completion of maintenance on the valve when the'LPI system is required to be OPERABLE 2. At least once per 18 months.DAVIS-BESSE, UNIT 1 3/4 5-5 Amendment No. 20 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: h. By performing a flow balance test, during shutdown, following completion of modifications-to the HPI or LPI subsystems that alter the subsystem flow characteristics and verifying the following flow rates: HPI System -Single Pump Injection Leg 1-1 Injection Leg 1-2 Injection Leg 2-1 Injection Leg 2-2> 375> 375 gpm at 400 psig*gpm at 400 psig** 375 gpm at 400 psig*375 gpm at 400 psig*LPI System -Single Pump Injection Leg 1 Injection Leg 2 2650 gpm at 100 psig**; 2650 gpm at 100 psig*** Reactor coolant pump discharge.
** Reactor coolant vessel.pressure at the HPI nozzle in the reactor coolant pressure at the core flood nozzle on the reactor DAVIS-BESSE,.UNIT 13 3/4 5-5a Amendment No. 2 0
.36 7 0 fl 6 9 6 EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS -T < 280&deg;F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:.a. One OPERABLE high pressure injection (HPI) pump, b. One OPERABLE low pressure injection (LPI) pump, c. One OPERABLE decay heat cooler, and d. An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) and transferring suction to the containment emergency sump.APPLICABILITY:
MODE 4.ACTION: a. With no ECCS subsystem OPERABLE because of the inoperability of either the HPI pump or the flow path from the borated water storage tank, restore at least one ECCS subsystem to OPERABLE status within one hour or be in COLD SHUTDOWN within the next 20 hours.b. With no ECCS subsystem OPERABLE because of the inoperability of either the decay heat cooler or LPI pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System Ta less than 280&deg;F by use of alternate heat removal methods.avg
: c. In the event the ECCS is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.SURVEILLANCE REQUIREMENTS 4.5.3 The ECCS subsystems shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.11 DAVIS-BESSE, UNIT 1 3/4 5-6
-*,O36 6 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)BASES 3/4.5.1 CORE FLOODING TANKS The OPERABILITY of each core flooding tank ensures that a sufficient volume of borated water will be immediately forced into the reactor vessel in the event the RCS pressure falls below the pressure of the tanks.This initial surge of water into the vessel provides the initial cooling mechanism during large RCS pipe ruptures.The limits on volume, boron concentration and pressure ensure that the assumptions used for core flooding tank injection in the safety analysis are met.The tank power operated isolation valves are considered to be"operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these tank isolation valves fail to meet single failure criteria, removal of power to the valves is required.The limits for operation with a core flooding tank inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional tank which may result in unacceptable peak cladding tempera-tures. If a closed isolation valve cannot be immediately opened, the full capability of one tank is not available and prompt action is required to place the reactor in a mode where this capability is not required.3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems with RCS average temperature
> 280&deg;F ensures that sufficient emergency core cooling capability wTll be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.
Either subsystem operating in conjunction with the core flooding tanks is capable of supplying-sufficient core cooling to maintain the peak cladding tempera-tures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward.In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period.DAVIS-BESSE, UNIT 1 B 3/4 5-1 Amendment No. 2 0 r- ) " tp EMERGENCY CORE COOLING SYSTEMS BASES With the RCS temperature below 280 0 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.
The Surveillance Requirements provided to ensure OPERABILITY of each component ensures, that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.
The decay heat removal system leak rate surveillance requirements assure that the leakage rates assumed for the system during the recirculation phase of the low pressure injection will not be exceeded.Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper 'flow resistance and pressure drop in the piping system to each injection point'is necessary to: (1)prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the.ECCS-LOCA analyses.3/4.5.4 BORATED WATER STORAGE TANK The OPERABILITY of the borated water storage tank (BWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on BWST minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in the cold condi-tion following mixing of the BWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly.These assumptions are consistent with the LOCA analyses.The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
The limits on contained water volume, and boron concentration ensure a pH value of between 7.0 and 11.0 of the solu-tion sprayed within containment after a design basis accident.
The pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion cracking on mechanical systems and components.
DAVIS-BESSE, UNIT 1 B 3/4 5-2 Amendment No. -;
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3.5.2 None None None Attachment 1, Volume 10, Rev. 0, Page 44, 46, 57, 59 of 98---Regarding Verification of current plant configuration vs NUREG SR 3.5.2.9 (ITS SR 3.5.2.7) and INSERT 1---Request: Please confirm that the Davis-Besse plant configuration continues to rely on inlet trash racks and screens (e.g as opposed to strainers).
Also please confirm the location/content of INSERT 1 and the applicability of JFD 4 (pg 44)---Basis for Request: 1. Many PWRs have reconfigured ECCS sumps. Most of the configurations rely on reinforced strainers in place of trash racks and screens. Further, in ITS BASES SR 3.5.2.7 includes reference to vertical strainers.
: 2. Reviewer could not locate INSERT 1 nor correlate JCO 4 to it.comment Commen. ---Regulatory Requirements:
&sect; 50.36 Technical Specifications (a) Each applicant for a license ... shall include ... proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications
... shall also be included in the application, but shall not become part of the technical specifications.(b) ... The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to &sect; 50.34. The Commission may include such additional technical specifications as the Commission finds appropriate.(c)(3) Surveillance requirements...
assure that the necessary quality of systems and components is maintained, that facility operation will be within safety http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 2 limits, and that the limiting conditions for operation will be met.Issue Date 1 02/14/2008 Close Da!te1 03/04/2008 Logged in User: Anonymous"'Responses Licensee Response by Bill 1. UFSAR Section 6.2.2.6, Containment Vessel Emergency Sump Bentley on 02/19/2008 provides a detailed description of the sump configuration.
In summary, the sump includes (1) Vertical strainer assemblies (2)System of trash racks in key passageways in CTMT to intercept large debris (3) Trash racks are shown in UFSAR Figure 6.2-33a, surrounding the upper sump vertical strainer assemblies.
: 2. Insert 1 is on page 45 of Volume 10.NRC Response by Ross Telson The reviewer understands that ITS SR 3.5.2.7, as amended, on 02/25/2008 visually inspects each ECCS train containment sump system suction inlet, including but not necessarily limited to: (1) Vertical strainer assemblies, (2) System of trash racks that intercept large debris in key passageways in CTMT, and (3) Trash racks shown in UFSAR Figure 6.2-33a, surrounding the upper sump vertical strainer assemblies.
With this inspection, the licensee will verify that there is no evidence of structural, distress, abnormal corrosion, or other abnormal degradation that would impact the ability of each ECCS train contanment sump to perform its specified safety function.
Please confirm for question closure, or correct.Licensee Response by Bill The reviewer understanding stated in the 2/25/08 response is Bentley on 02/28/2008 correct.NRC Response by Ross Telson Thank you for your timely response.
The reviewer has no further on 03/04/2008 questions on this question thread at this time. This thread is being closed. Should unanticipated questions arise, another question thread will be opened.Date Created: 02/14/2008 02:08 PM by Ross Telson Last Modified:
03/04/2008 10:58 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsfl1 fddceal Od3bdbb585256e...
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3.5.2 None None 2 Attachment 1, Volume 10, Rev. 0, 48, 49, 51, 52, 53, 55,59 of 98---Regarding B3.5.2 ECCS -Operating BASES -BACKGROUND, APPLICABLE SAFETY ANALYSES, ACTIONS---Request: Please clarify the applicable BJFD (e.g. BJFD2 or 3) with regard to the following changes: 1. Insertion of the word "Main" prior to the word "Steam" in sub-paragraph
: d. description of functions of the ECCS. (e.g. Does ECCS not provide core cooling following steam line breaks other than on the main steam lines?; Is this an error in the NUREG or is the design basis for ECCS different for Davis-Besse than for the typical B&W plant modeled in the NUREG?2. Deletion of the phrase "Control valves are set to balance the HPI flow to the RCS...3. Substitution of the phrase "of approximately 1600 psig" for the phrase"above the opening setpoint of the pressurizer safety valves." 4. Substitution of "psig" for "psia" 5. Substitution of INSERT 1 for the phrase "...is from the hot leg through..." 6. Substitution of the phrase "HPI pump 2 in piggy-back..." for the phrase"one LPI train into the pressurizer..." 7. Deletion of "Engineered" and "E" from "Engineered Safety Feature" and"ESFAS" 8. Substitution of "ESF" for "safeguards" 9. Substitution of "essential" for Safety Feature (ESF).10. The addition of the word "associated" before "ESFAS" and the deletion of the "E" in "ESFAS" http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 4 11. Deletion of "For a large break LOCA, HPI is not credited at all." 12.... (the justification for most NUREG BACKGROUND BASES deviations are neither well understood by the reviewer nor clearly justified in BJFDs 2 and 3.) Please schedule a conference call when ready discuss/clarify NUREG deviations.
---Basis for Request: Applicable B-JFDs provide insufficient information for ITSB reviewer to determine:
: 1. Whether deviations impact ITS in a manner warranting further review by organizations external to ITSB that exercise functional responsibility.
: 2. Whether plant-specific deviations affect the completeness of the ISTS---Applicable References NEI 96-06 -Improved TS Conversion Guidance, 2.7 Deviations from the Applicable ISTS: "...a high threshold should be satisfied for deviating from the ISTS..." 58 FR 29132 (pp 39132-39139)
Final Policy on &sect; 50.36 Technical Comment Specifications, IV. The Commission Policy: "...it is the Commission intent that the wording and Bases of the improved STS be used in the Technical Specification related submittal to the extent practicable."---Regulatory Requirements:
&sect; 50.36 Technical Specifications (a) Each applicant for a license ... shall include ... proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications
... shall also be included in the application, but shall not become part of the technical specifications.(b) ... The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to &sect; 50.34. The Commission may include such additional technical specifications as the Commission finds appropriate.(c)(3) Surveillance requirements...
assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.Issue Date] 02/14/2008 Close Date [03/20/2008 Logged in User: Anonymous'Responses Licensee Response by Bill 1. UFSAR 15.4.4 describes the results of the analysis for steam Bentley on 02/25/2008 line breaks. Only the main steam line break result discusses operation of the ECCS. 2. The HPI discharge valves open fully upon SFAS actuation.
These valves are not preset to any throttled position as described in the deleted phrase. UFSAR 6.3.3.1.3 describes the Small Break LOCA analysis.
The following paragraph can be found at the top of page 6.3-20 "A second special type of SBLOCA is the HPI line break. For this accident; a break is postulated in one of the injection lines with the active HPI pump, between the last check valve in the line and the RCS cold http://www.excelservices.coni/exceldbs/itstrack-davisbesse.nsf/1 fddcealOd3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 3 of 4, leg pipe. The HPI flow rate assumptions for this SBLOCA are different than for the classical RCS cold leg SBLOCA and CFT line break since the back pressure in the broken and intact HPI lines vary dramatically.
At Davis Besse, one HPI pump.feeds two cold legs and due to the difference in back pressure, no HPI flow is initially available for core cooling. Operator action is credited at 10 minutes after loss of subcooling margin to balance the flow between the legs." 3. Shutoff head for the HPI pumps is approximately 1600 psig. Therefore, the HPI pumps are not capable of discharging at a pressure above the opening setpoint of the pressurizer safety valves (nominal 2500 psig setting).
: 4. 200 psig is the approximate shutoff head for the LPI pumps. Therefore, that is the point at which they are capable of discharging to the RCS. 5 and 6. UFSAR 6.3.3.1.2.1 describes the 2 methods of Boron Precipitation Control. Substitution of the questioned phrases makes the description read correctly.
: 7. The term "Engineered" and "E" have no meaning at Davis-Besse.
We refer to Safety Features Actuation System or SFAS. This change was made throughout the ITS Bases in any ITS section that described this systems. 8. At Davis-Besse, "safeguards" is synonymous with"security." The term "safeguards" was replaced with "Engineered Safety Features (ESF)" in any ITS Bases section that used this terminology.
: 9. At Davis-Besse, we do not refer to ESF buses. We refer to essential buses. This change was made inany ITS Bases section that used this terminology.
: 10. The associated SFAS signals are described in the ITS 3.3.5 Bases for SFAS. In particular, page 201 and 202 of Volume 8 describe that different incident levels of actuation in SFAS cause certain components to be actuated.
See #7 for deletion of "E". 11. For the current type of fuel loaded, the Large Break LOCA analysis credits HPI flow. 12.If the above explanations are not adequate to close this question, they can be discussed on a call.NRC Response by Ross Telson on 03/19/2008 For closure of this question thread, please confirm or correct, as appropriate, the following staff understanding from your responses:
: 1. (Questions 1, 2, 3, 5 and 6) The analyses and evaluation included in the updated final safety analysis report: a.Do not rely on ECCS operation for steam line breaks other than the main steam line. b. Do not rely on control valves being set to balance the HPI flow to the core to meet analysis assumptions following a small break LOCA in one of the RCS cold legs near an HPI nozzle [for the current- and for any future ECCS pump that could be installed in accordance with current design requirements].
: c. Do not rely on HPI'pumps being capable of discharging to the RCS at an RCS pressure greater than the opening set point of the pressurizer safety valves. d. Rely, during long term cooling, on establishing the following LPI flow paths to preclude the possibility of boric acid in the core region reaching an unacceptably high concentration:
i) One flow path uses the discharge of LPI pump 1 through a line that bypasses the RCS to Decay Heat Removal (DHR) system suction line and allows http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/lfddcealOd3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 4 of 4 reverse flow into the DHR system loop ii) One flow path is through the pressurizer auxiliary spray line from HPI pump 2 in piggy-back with LPI pump 2. 2. (Previous Questions 7 and 8) The terms "Engineered" and "E" HAVE meaning at Davis-Besse when used in the phrases "Engineered Safety Features" or "ESF" in place of the STS term "Safeguards." However, Davis-Besse refers to electrical buses powering ESF equipment as "Essential" buses rather than the STS terms "Engineered Safety Feature" bus or"ESF" bus. Likewise, Davis-Besse uses the terms "Safety Features Actuation System" and "SFAS" in place of the STS terms"Engineered Safety Features Actuation System" and "ESFAS." The licensee uses this alternate terminology consistently throughout the FSAR, ITS, and other licensing and licensee-controlled documents and does not wish to conform to the STS with regard to this terminology.
Licensee Response by Bill The following corrections are needed for the staff understanding Bentley on 03/20/2008 posted on 3/19/2008:. (L.b) Substitute "pre-set" for "set". (1.d.i)Substitute "drop line" for "loop". Everything else is confirmed to be correct.NRC Response by Ross Telson Thank you for the confirmation
/ correction.
The reviewer has no on 03/20/2008 further questions regarding this question thread at this time.Should unanticipated questions arise, a new question thread may be opened.Date Created: 02/14/2008 04:45 PM by Ross Telson Last Modified:
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Bases JFD Number: 3.5.3 None None 2 Attachment 1, Volume 10, Rev. 0, 69, 71, 73, 75, 78 of 98---Regarding 3.5.3 ECCS-Shutdown and B3.5.3 ECCS -Shutdown BASES -BACKGROUND, APPLICABLE SAFETY ANALYSES, ACTIONS vs JFD 1, 3 and BJFD 2, 3---Request: Please clarify the applicable JFDs with regard to the following changes: 1. Alternating use of the term "LPI subsystem" in place (a) the word "train" in the context of the LCO and (b) the term "DHR loops" in the context of the NOTE in the ACTIONS. (e.g. Are ECCS DHR loops equivalent to One ECCS train?)2. Addition of phrase "(i.e., decay heat cooler)" following the phrase "heat exchanger" 3. Deletion of the phrase "Included in these reductions is that..." 4. Deletion of the word "-shutdown" in the phrase "The ECCS train-shutdown satisfies..." 5. Insertion of the word "manually" between the words "and" and"transferring suction to the containment sump." 6. Insertion of the word "emergency" between the words "containment" and"sump." 7. Insertion of the INSERT 2 following the word "sump." 8. Substitution of the phrase "core flood" for the phras "four cold leg injection." 9. Deletion of the phrase "and to supply its flow to the RCS hot and cold legs."---Basis for Request: Applicable B-JFDs provide insufficient information for ITSB reviewer to http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/l fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 4 determine:
: 1. Whether deviations impact ITS in a manner warranting further review by organizations external to ITSB that exercise functional responsibility.
: 2. Whether plant-specific deviations affect the completeness of the ISTS---Applicable References NEI 96-06 -Improved TS Conversion Guidance, 2.7 Deviations from the Applicable ISTS: "...a high threshold should be satisfied for deviating from the ISTS... This high threshold is used to preserve the standardization of the use and meaning of the requirements for the industry and the NRC." 58 FR 29132 (pp 39132-39139)
Final Policy on &sect; 50.36 Technical Specifications, IV. The Commission Policy: "...it is the Commission intent that the wording and Bases of the improved STS be used in the Technical Specification related submittal to the extent practicable." August 3, 2007 Letter from Mark B. Bezilla, Vice'President
-Nuclear, FirstEnergy Nuclear Operating Company: "FENOC proposes to revise the current Technical Specifications (CTS) to the Improved Technical Specifications (ITS) consistent with Improved Standard Technical Specifications (ISTS) as described in NUREG-1430, 'Standard Technical Comment Specifications Babcock and Wilcox Plants,' Revision 3.1, and certain generic changes to the NUREG. The guidance of NEI 96-06, "Improved Technical Specifications Conversion Guidance," dated August 1996, and Nuclear Regulatory Commission (NRC) Administrative Letter 96-04, "Efficient Adoption of Improved Standard Technical Specifications," dated October 9, 1996, were used in preparing this submittal."---Regulatory, Requirements:
&sect; 50.36 Technical Specifications (a) Each applicant for a license ... shall include ... proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications
... shall also be included in the application, but shall not become part of the technical specifications.(b) ... The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to &sect; 50.34. The Commission may include such additional technical specifications as the Commission finds appropriate.(c)(3) Surveillance requirements...
assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.Issue 02/15/2008 Close Date 03/17/2008
-. -Logged in User: Anonymous' Responses I,.. .1 Licensee Response by Bill L .a. The term "LPI subsystem" is used in a manner equivalent to Bentley on 02/23/2008 that used in ITS 3.5.2; Condition A. Since in Mode 4, Davis-Besse does not require HPI'to be operable, use of the term ECCS train did not appear to be appropriate.
l.b. In the context of ECCS requirements, it is believed that LPI is the more appropriate term.http://www.excelservices;com/exceldbs/itstrack-davisbesse.nsf/lfddcealOd3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 3 of 4 When the subsystem is being used for its normal decay heat removal mode, then DHR is appropriate (as for ITS 3.4.6, 3.4.7, 3.4.8, 3.9.4 and 3.9.5). 2. Decay Heat Cooler is the appropriate Davis-Besse terminology.
This change was made in any ITS Bases that discussed DHR heat exchangers.
: 3. The subject phrase is not applicable.
HPI is not required to be operable, so hence there are no requirements for manual initiation of HPI. The LPI subsystem that is required to be operable must remain in its normal LPI lineup, ready for automatic actuation.
: 4. "ECCS train-shutdown" is the title of the specification.
Typically, the title is not used in the Criterion statement of the Bases. This was changed to stipulate the LCO requirement.
: 5. The ITS 3.5.2 Bases made it clear that this transfer was a manual transfer.
Adding the word "manually" makes it clear in the ITS 3.5.3 Bases also. 6. The suction of the ECCS pumps is manually transferred from the BWST to the containment emergency sump. Since that is the name of the sump, the word "emergency" was inserted.
: 7. Insert 2 is simply a recognition that once the BWST has been emptied and suction transferred to the emergency sump, the recirculation phase of ECCS operation begins (just as was described in the ITS 3.5.2 Bases). 8. The HPI pumps discharge into the cold leg injection nozzles. Since the HPI pumps are not required to be operable, the phrase was changed to reflect the flowpath for the LPI pump. 9.The flow path supply to the RCS was already described in the preceding sentence, so there was no need to repeat it. The phrase was also deleted because it does not apply. HPI discharges to the cold legs, and is not required to be operable.
LPI does not discharge to the hot legs.NRC Response by Ross Telson on 03/13/2008 Staff discussion of this question thread with the licensee during a recent conference call clarified the licensee's justification for deviating from the iSTS with regard to questions 1, 2, 3, 4, 6, 8, and 9. However, staff understood that it might be more appropriate to return to (or more closely approximate) iSTS BASIS wording with regard to those proposed changes corresponding to questions 5, 7. Specifically:
Items 5 and 7: The licensee proposed to deviate from the following iSTS LCO BASIS phrase as follows: "In MODE 4, an ECCS train consists of an HPI subsystem and an LPI subsystem.
Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the BWST and transferring suction to the containment sump." Licensee proposed the following in its place: "In MODE 4, an ECCS subsystem consists of an LPI subsystem.
An LPI subsystem consists of an LPI pump, a decay heat cooler, and the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction form the BWST and manually transferring suction to the containment emergency sumpduring the recirculation phase, of operation." Staff understands: (a) that, contrary to the iSTS, the CTS do not require an OPERABLE HPI train in the applicable MODE and that the licensee does not wish nor consider it necessary to conform to the iSTS in this regard, (b)http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 4 of 4 that switch-over from LPI injection to recirculation requires BOTH the satisfaction of an AUTOMATIC permissive interlock (which is based on BWST level and ensures an adequate volume of injection prior to switch-over) and control room operator switch manipulations to effect MANUAL transfer, and (c) that the ECCS subsystem must be CAPABLE of transferring suction to the containment sump at all times in the applicable MODE -not just during the recirculation phase of operation.
Thus, staff questioned (b) the proposed addition of the word "manually" in that it might lead to confusion in understanding the basis or reasons for the TS LCO and the mistaken belief that the "automatic" switch-over permissive interlock is NOT required to be OPERABLE, and (c)the addition of the phrase "during the recirculation phase of operation" in that it might lead to confusion as to when the ECCS subsystem must be capable of transferring suction to the containment sump. Q I: Please confirm or correct staff understanding stated above. Q2: Justify, in the context of the above discussion, why the addition of the word "manually" and the phrase "during the recirculation phase of operation" are necessary and will not contribute to confusion or incorrect interpretation of the ITS... or modify/eliminate these iSTS Bases deviations, as appropriate.
Licensee Response by Bryan Kays on 03/16/2008 Q I response:
In the body of the staff response on 3/13/2008, the following is stated "(b) the proposed addition of the word"manually" in that it might lead to confusion in understanding the basis or reasons for the TS LCO and the mistaken belief that the"automatic" switch-over permissive interlock is NOT required to be OPERABLE." The insertion of the word "manually" is exactly equivalent to the use of the word "manually" in the ISTS 3.5.2 Bases, Volume 10, page 59, LCO Bases section, second paragraph, last sentence.
Since manually transferring suction inherently includes receipt of the permissive interlock, the chance of this confusion or mistaken belief seems unlikely.
Also, the permissive is required to be operable as part of SFAS Instrumentation (Volume 8, ITS 3.3.5, Function 5). Therefore, Item 5 should be maintained as is. There are no issues with the remainder of the staff understanding, Q2 response:
See above justification for addition of the term "manually." The term "during recirculation phase of operation" will be deleted. A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Ross Telson on 03/17/2008 Thank you for the clarification and revision.
The reviewer has no further questions regarding this question thread at this time.Should unanticipated questions arise in the future, another question thread will be initiated.
Date Created: 02/15/2008 10:05 AM by Ross Telson Last Modified:
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NRC ITS TRACKING NRC Reviewer ID 200802111514 Conference Call Requested?
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ITS 3.5 Ross Telson 1 83 Information ITS Number: 0S.:. !DO.C. Number:. Basesl JFD Number.:.3.5.4 None M.1 None Attachment 1, Volume 10, Rev. 0, Pages 83, 84, 87, 94 of 98---Regarding:
addition of a proposed maximum outside air temperature to the minimum outside air temperature in CTS SR 4.5.4.b, (iSTS SR 3.5.4.1).---Request: 1. Please correct the disagreement between MO1 (pg 84) and ITS SR 3.5.4.1.MO1 states: "...the ITS SR 3.5.4.1 Note only requires the BWST water temperature to be verified within the maximum limit if the ambient air temperature is greater than new maximum BWST water temperature limit. This changes the CTS by adding a new maximum BWST water temperature limit and requires it checked every 24 hours unless the ambient air temperature is less than or Comment equal to the maximum water temperature limit..." In contrast, the ITS SR 3.5.4.1 Note, as proposed, (pg 87) actually states "Only required to be performed when ambient air temperature is < 35 F or > 90 F." This SR and note are further discussed in the ITS SR 3.5.4.1 BASIS (pg 94).BWST water temperature must be verified every.24 hours to be within the allowed temperature band any time ambient temperature is outside (above or below) the specified band -not just when it is above the specified band.2. Please confirm whether the TS SR-proposed temperature band includes appropriate margin to accommodate instrument uncertainty, drift, etc. What is the magnitude of that margin? If not accommodated in the TS SR acceptance criteria, please identify the licensee process under which it is controlled and indicate whether the process is subject to 10 CFR 50.59 change control regulation.
---Basis for Request: http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/IfddceaIOd3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 2 MO1 and ITS SR 3.5.4.1 appear to differ unnecessarily and a new SR numerical limit is being added. It is necessary to understand how that value will be applied.---Regulatory Requirements:
&sect; 50.36 Technical Specifications (c)(3) Surveillance requirements...
assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.F Issue Date 1]02/11/2008 Close Date [03/17/2008 Logged in User: Anonymous'Responses Licensee Response by Jerry Response 1 The statement in Discussion of Change (DOC) M01 Jones on 03/10/2008 (Volume 10, Page 84) that the Note only requires the BWST water temperature to be verified within the maximum limit if the ambient air temperature is greater than the new maximum BWST water temperature limit is essentially correct since there is really no conceivable way the BWST temperature can be greater than the maximum limit when the air temperature is less than the minimum limit of 35 degrees F. While SR 3.5.4.1 (Page 87) requires verification that the BWST water temperature is within both the minimum and maximum limits when air temperature is less than 35 degrees F, in reality when less than 35 degrees F outside air temperature the BWST SR is really ensuring that the BWST water temperature is greater than or equal to 35 degrees F, the minimum limit. The wording in the DOC is presented in this manner since the DOC is justifying the addition of a new maximum temperature limit. However, for clarity, the DOC will be modified as shown in the attached markup. This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
Response 2 BWST Temperature was addressed in a previous question thread (NRC Question 200801021633, Jerry Jones Response on 02/11/08 for Action 2d). Any changes to the surveillance procedure and calculations are subject to 10 CFR 50.59.NRC Response by Ross Telson Thank you for the clarifications and corrections.
The reviewer has on 03/17/2008 no further questions regarding this question thread at this time.Should unanticipatedquestions arise, a new question thread will be opened.Date Created: 02/11/2008 03:14 PM by Ross Telson Last Modified:
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No Categoy J In Scope ITS Section: TB POC: JFD Number: Page Number(s),:&#xfd; ITS 3.5 Ross Telson None Information ITS Nummbr: OS DOCNumber:
Bas.es JFD.. N.mbe.r.: 3.5.4 None None 3 Attachment 1, Volume 10, Rev. 0, Pages 93 & 96 of 98,---Regarding:
ITS BASES B3.5.4 Borated Water Storage Tank (BWST) ACTION A.1. (pg 93) B-JFD 3 (pg 96)---Request: 1. Please clarify substitution of the word "Containment" for the words"Reactor Building" in the phrase "In this condition, neither the ECCS nor the Reactor Building Spray System can perform its design functions." B-JFD 3 states that this change is made to reflect the plant specific nomenclature.
Does Comment Davis-Besse consider the term "Containment" to mean the same at Davis-Besse as the term "Reactor Building" in the typical B&W plant modeled in the STS basis? If not, is the NUREG BASIS in error?---Basis for Request: The reviewer is unfamiliar with use of the term "Containment" to mean the same as the term "Reactor Building."---Regulatory Requirements:
&sect; 50.36 Technical Specifications (c)(3) Surveillance requirements...
assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.Issue Date 02/11/2008 Close Date 1 03/04/2008 Logged in User: Anonymous Responses http://www.excelservices.com/exceldbs/itstrackdavisbesse.insf/l fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 2 Licensee Response by Bill Bentley on 02/20/2008 The term "reactor building" is generally equivalent to the term"containment." The term "reactor building" is not used at Davis-Besse. The difference between the use of the term "reactor building" in ISTS and the use of the term "containment" for Davis-Besse is best illustrated in Volume 11, ITS 3.6.1 Bases, Background section. In the particular instance noted by the reviewer, it appears that using the term "containment" instead of the term "reactor building" in ISTS 3.5.4 Action A. 1 provides better consistency with the rest of the ISTS 3.5.4 Bases. The term"containment spray" is used throughout the ISTS 3.5.4 Bases, as listed below: (1) Background, 1 st paragraph (2) Background, 2nd paragraph (3) Applicable Safety Analysis, 1 st paragraph (note also that a reference is made to LCO 3.6.6, with the term "Containment Spray" in the title). (4) Applicable Safety Analysis, 3rd paragraph (5) LCO section (6) Applicability, 1 st paragraph (7) Action B. I NRC Response by Ross Telson on 03/04/2008 Thank youfor providing clarification.
The reviewer has no further questions on this question thread at this time and is closing the thread. Should unanticipated questions arise in the future, another question thread will be opened.Date Created: 02/11/2008 03:39 PM by Ross Telson Last Modified:
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ITS 3.5 Ross Telson None 0 Information ITS Number.:.
OS.1:' DOC Number: Bases.JFD Number: None None None 1 This Question stems from discussion in Question ID: 200710032123.
Although this question is posted under ITS Section 3.5, it's scope extends to all ITS sections in which numerical acceptance values are stipulated.
---- ACTIONS NEEDED: 1. Clarify licensee handling of uncertainty associated with ITS conversion.
In doing so: a. Explicitly identify all instances in which uncertainty is NOT incorporated in the ITS specified values and NOT explicitly identified as such in ITS or ITS Bases. For each instance, amend the proposed ITS or ITS Bases to explicitly identify this fact.b. Explicitly identify all instances in which the incorporation of uncertainty in specified values CHANGES from "INCLUDED" in the CTS to "EXCLUDED" in the proposed ITS.c. For each instance identified above, amend DOCs and JFDs, as necessary, to explicitly identify and justify the acceptability of deviation to exclude uncertainty in terms of your current licensing bases, providing adequate protection, and compliance with applicable NRC regulations.
Include, as applicable, discussion of &sect; 50.46 and Part 50 Appendix K requirements regarding uncertainty and &sect; 50.36(c)(3) requirements regarding assurance that the necessary quality of systems and components is maintained and that facility operation will be within safety limits. Include applicable analyses and evaluation included in the safety analysis report. Describe and attach plant procedures/programs that support your justification that uncertainty, which is excluded from ITS, is and will continue to be managed and applied in an appropriately controlled manner acceptable to the NRC to provide the required assurance of safety and regulatory compliance..http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 7 d. For the reviewer to better understand the scope, actions to resolve, and licensee bases for the mixed incorporation of uncertainty in technical specification values, attach the corrective action documentation, referenced in 11/28/07 TELECON.---- BASIS FOR REQUEST: 1. 10/31/2007 response to 200710032123 Action 2 by Jerry Jones stated that instrument inaccuracies of parameters in Surveillance Requirements are not normally included.
This is contrary to guidance in some portions of the ITS.Other portions of the ITS may be silent with regard to incorporation of uncertainty.
To assure that operators and inspectors will interpret and apply the ITS in a safe manner, it is essential that there be no ambiguity as to the meaning of ITS-specified values -especially when an incorrect assumption could lead to a non-conservative application of the ITS or to unrecognized plant operation exceeding limiting conditions for operation.
: 2. TELECON between licensee and reviewer in which the licensee indicated that incorporation of uncertainty inCTS was somewhat of a 'mixed-bag' in that some surveillance requirements included uncertainty While others did not and that this issue had been addressed in a licensee corrective action document.3. Staff precedent has generally been that uncertainty is (or should be)incorporated into TS numerical limits. This is consistent with a number of regulations, including, but not necessarily limited to, &sect;50.36, &sect;50.46, and Part 50, Appendix K and is evidenced by NRC INSPECTION MANUAL PART 9900: TECHNICAL GUIDANCE -STS Section 3.0 "Acceptable Comment Measurement, Tolerances For TS Limits,"which provides guidance to NRC Comment. Inspectors.
It states, in part, "The TS limits are established with allowance for measurement tolerances already incorporated.
The limits take into consideration measurement uncertainties as necessary to assure safe plant operation.
The stated limit presupposes that the licensees have tolerances consistent with normal industry standards (e.g., ASTM, IEEE, ACI, etc.)." 4. The use of limiting values in ITS that exclude uncertainty margins (but do not explicitly identify this fact in the ITS or ITS Bases) does not provide reasonable assurance that associated ITS limiting values will be correctly interpreted and applied. This practice could contribute to an incorrect interpretation or application of the ITS limiting value which, in turn, could lead to unrecognized plant operation exceeding limiting conditions for operation.
---- REGULATORY REQUIREMENT:
&sect; 50.36 Technical Specifications Title 10 Code of Federal Regulations 50.36, "Technical Specifications" (a) Each applicant for a license ... shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications
... shall also be included in the application, but shall not become part of the technical specifications.(b) ... The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to &sect; 50.34. The Commission may include such additional technical specifications asthe Commission finds appropriate.(c) Technical specifications will include items in the following categories:
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7/17/2008 NRC ITS Tracking Page 3 of 7 (2) Limiting conditions for operation.(i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria: (C) Criterion
: 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.(3) Surveillance requirements.
Surveillance requirements are requirements
...to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation Will be met.&sect; 50.46 Acceptance Criteria For ECCS Requires, in part, that uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated.
This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of this section, there is a high level of probability that the criteria would not be exceeded.Appendix K to Part 50--ECCS Evaluation Models Requires in part......
An assumed power level lower than the level specified in this paragraph (but not less than the licensed power level) may be used provided the proposed alternative value has been demonstrated to account for uncertainties
...... Shutdown reactivities resulting from temperatures and voids shall be given their minimum plausible values, including allowance for uncertainties...
The comparisons shall quantify the relation of the correlations to the statistical uncertainty of the applicable data...Correlations of heat transfer from the fuel cladding to the surrounding fluid in the post-CHF regimes of transition and film boiling shall be compared to applicable steady-state and transient-state data using statistical correlation and uncertainty analyses...
IssueDate][
11/16/2007 Close Date [03/03/2008 Logged in User: Anonymous'Responses 1 .1 Licensee Response by Bill The NRC requested that Davis-Besse clarify the handling of Bentley on 12/18/2007 uncertainty associated with values in the ITS. The following are our specific responses to the individual questions:
L.a This request represents a beyond scope change. If the CTS or CTS Bases is not explicit with respect to the handling of uncertainty, and if the associated ISTS or ISTS Bases is not explicit with the handling of uncertainty, then no change is necessary.
The explicit handling of uncertainty is beyond the scope of the ITS Conversion for these sections of the ITS. Davis-Besse will examine question (la) after http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/lfddcea1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 4 of 7 implementation of the ITS Conversion has been completed in accordance with the Davis-Besse corrective action program. 1.b The following response to question 1.b does not include ITS.Specification 3.5.1. The explicit handling of uncertainty for ITS Specification 3.5.1 is being addressed under question 200710032123, and all changes for ITS 3.5.1 (DOCs, JFDs, etc)will be addressed as part of the response to question 200710032123.
The' following response to question 1.b does not include ITS Specification 3.5.4. Changes to ITS 3.5.4 regarding the explicit handling of uncertainty are needed as described in the response for question 1.c. In none of the other ITS sections was uncertainty that is "INCLUDED" in the CTS changed to be"EXCLUDED" in the ITS. The above conclusion can be reached from a review of the CTS, CTS Bases, ISTS and ISTS Bases. It can be seen that during the development of the ITS Conversion submittal, the explicit handling of uncertainties had to be addressed for the following ITS Specifications:
2.1.1, 3.1.2, 3.1.4, 3.1.6, 3.1.7, 3.2.1, 3.2.2, 3.2.3, 3.2.4, 3.2.5, 3.3.1, 3.3.5, 3.3.8, 3.3.11, 3.3.15, 3.3.16, 3.4.1, 3.4.4, 3.4.9, 3.4.11, 3.7.9, 3.9.1, 3.9.2, 4.0. The explicit handling of uncertainties in these and any other ITS Specification is addressed as part of the Davis-Besse ITS conversion in the Discussion of Changes and Justification for Deviations.
Any specific question with regard to the explicit handling of uncertainty should be addressed as part of the review of any particular ITS Specification.
1.c Davis-Besse has identified that two changes are needed for ITS Specification 3.5.4. (1) The ISTS 3.5A4 Bases (Volume 10, Page 92) states that "The numerical values of the parameters stated in the SR are actual values and do not include allowance for instrument errors." This statement is true for all values, except the BWST minimum volume amount. The minimum volume specified in CTS LCO 3.5.4.a (Volume 10, Page 83) for the BWST is 500,100 gallons. The explicit handling of instrument uncertainty for this value is not addressed in the CTS Bases. The explicit handling of instrument uncertainty is also not directly addressed in License Amendment Request 241 (the license amendment that changed the value to 500,100 gallons).
However, instrument uncertainty is included as part of the value. The 500,100 gallon value is based on several values starting with a minimum level to ensure adequate protection from vortexing (75 inches) and Operator response time (87 inches -i.e., 75 inches plus an additional 12 inches for Operator response time) to initiate opening of the containment sump valves and closing the BWST suction valves. This value is then adjusted for instrument uncertainty (for a total of 108.5 inches for the current Tech Specs)to establish a permissive setpoint to provide protection against vortexing if the transfer is started too late due to instrument error.The Allowable Values are on either side of the trip setpoint are 101.6 inches and 115.4 inches. These Allowable Values are stipulated in the Allowable Value for CTS Table 3.3-4, Functional Unit f (Volume 8, Page 178) and maintained in ITS 3.3.5, Table http://www.excelservices.com/exceldbs/itstrack_davisbesse'.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 5 of 7 3.3.5-1, Function 5 (Volume 8, Page 195). Additional instrument uncertainty is added to the setpoint, then the minimum volume assumed in the analyses of 360,000 gallons injected into the core/containment is added to establish a minimum level for surveillance.
This additional instrument uncertainty is included to ensure the transfer is not started too early due to instrument error and thus not meeting the 360,000 gallon minimum injection requirement.
This minimum value is 500,051 gallons, and is then increased for the Technical Specifications to the current 500,100 gallon value. (2) While not directly related to the explicit handling of uncertainty, Davis-Besse has identified the need for a change to ITS Specification 3.5.4. CTS LCO 3.5.4 (Volume 10, Page 83)states "The Borated Water Storage Tank (BWST) shall be OPERABLE with: a. An available borated water volume of between..." ISTS SR 3.5.4.2 (Page 87) verifies "BWST borated water volume is..." The term "available" in CTS LCO 3.5.4.a (Page 83) was not adequately addressed as part of the conversion.
The CTS BWST water volume limits are based on the water volume that is available above the top of the discharge line penetration into the bottom of the tank. The top of the discharge line is 4 inches above the bottom of the BWST. The ISTS water volume limits are based on total volume (as described in the ISTS Bases). Davis-Besse requests that an RAI question be posted against ITS Specification 3.5.4, so that we may provide appropriate markups to the ITS Conversion to address these two changes. 1 .d A portion of the corrective action document is being provided for information and is included as an attachment to this response.
While the attached information is believed to be accurate as of the year 2003, it has not been verified for any changes that might have occurred since then.NRC Response by Ross Telson on 12/19/2007 Licensee Response by Bill Bentley on 12/18/2007 to RAI-screened question l.a. "This request represents a beyond scope change," does not resolve the request. If staff considered the 'request to represent a beyond scope item (BSI), a TAC would be opened to facilitate technical branch review of that issue. Rather, staff consider the request to represent a potential emergent staff issue (ESI) -A staff concern or a documented staff position that opposes a proposed ITS that IS congruent with either CTS or STS (or is justified by site-specific CLB). The request specified (1) Actions Needed, (2) the. Basis for the Request, and (3) Applicable Regulatory Requirements and was approved by the ITSB Branch Chief. If these elements are not clearly understood or under dispute, this topic may be placed on the agenda for further discussion during next week's conference call. Additional applicable guidance for consideration:
The nonvoluntary addition of new requirements (during an ITS conversion) may only be imposed through application of &sect; 50.109 (The Backfit Rule).However, if staff-suggested additional changes are needed to make the licensee-requested changes acceptable from the standpoint of (1) adequate protection or (2) compliance with NRC regulations, http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 6 of 7 then &sect; 50.109 does not apply and the request may be denied without the additional items. -Requests for additional information (RAIs) serve thepurpose of enabling the staff to obtain all relevant information needed to make a decision on a licensing action request that is fully informed, technically correct, and legally defensible.
-RAls are necessary when the information was not included in the initial submittal, is not contained in any other docketed correspondence, or cannot reasonably be inferred from the information available to the staff. -RAIs should be directly related to the applicable requirements related to the amendment application, and consistent with the applicable codes, standards, regulatory guides, and/or the applicable Standard Review Plan sections.
-RAIs should not be used as general information requests or as a means to encourage commitments from licensees.
&sect; 2.108 Denial of application for failure to supply information (a) The Director of Nuclear, Reactor Regulation
... may deny an application if an applicant fails to respond to a request for additional information...
Licensee Response by Bill Bentley on 02/19/2008 During the 2/6/08 phone call, the NRC reviewer made a request with respect to this question.
The request was to provide examples of instances where instrument uncertainty is not included in the Tech Spec limits. For these examples, show the barriers in place that prevent exceeding the applicable limits. See attached file for 5 examples.NRC Response by Ross Telson on 02/25/2008 Staffs recollection of the 2/6/08 phone call differs slightly:
The NRC reviewer asked the following questions and recalls the following licensee responses:
: 1. Are there instances, in the requested ITS, in which uncertainty is not factored into ITS-specified values and not explicitly identified as such in the ITS or ITS Bases -Licensee Response:
Yes, approximately 3 instances.
: 2. Do barriers exist to prevent operators and inspectors from incorrectly interpreting the ITS? -Licensee:
Yes 3. Please describe the barriers and their controls, give one or more examples, and document your responses to this discussion.
-Licensee:
Surveillance procedures, will.comply.
: 4. The reviewer is prepared to close this question upon receipt of the licensee's written confirmation of the following reviewer understanding: "In all instances, in ITS Sec 3.5, in which uncertainty was not factored into ITS-specified values and not explicitly identified as such in the ITS or ITS Bases, the licensee has established and will maintain effective barriers to prevent operators and inspectors from incorrectly interpreting and applying those ITS-specified values." 5. If the reviewer's understanding is not correct, please provide the correct interpretation of the licensee's 2/6 verbal and 2/19 written responses.
Licensee Response by Bill Bentley on 02/29/2008 The reviewer's understanding for item #4 posted on 2/25/08 is correct. Effective barriers have been established and will be maintained for instances where uncertainty is not factored into the ITS limit and not explicitly identified as such in the ITS or ITS http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/
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7/17/2008 NRC ITS Tracking Page 7 of 7, Bases. For clarification
-the only ITS Section 3.5 limits that are applicable are BWST Temperature and BWST Maximum Volume.As described in question 200801021633, site specific calculations provide adjusted surveillance limits, which have been incorporated into the surveillance procedures.
Ongoing maintenance of the barrier is established by the fact that the 10 CFR 50.59 process must be applied for any future procedure changes to these limits.NRC Response by Ross Telson Thankyou for confirming the reviewer's understanding.
The on 03/03/2008 reviewer has no further questions on this question thread at this time. Should there be any further questions regarding question 200801021633, they will be addressed in that thread. Should new, unanticipated, questions arise involving this question thread, a new thread will be opened.Date Created: 11/16/2007 05:16 PM by Ross Telson Last Modified:
03/03/2008 10:14 AM http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 FirstEnergy, DAVIS-BESSE NUCLEAR POWER STATION'I---I Root Cause Analysis Report Instrument Uncertainty for Non-LSSS Applications CR 02-06407, Dated 09/16/02 REPORT DATE: 02/21/03 Prepared by: Reviewed by: Reviewed by: Approved by: DEEC Problem Statement Description of reason for investigation Original Scope During a review of the Service Water and High Pressure Injection Systems in September 2002, it was discovered that instrument drift was not being properly factored into instrument uncertainty." Drift was not factored into the instrument uncertainty for some control room instrumentation.
* Drift calculations were not entered into the system as formal calculations.
Expanded Scope The original scope of this condition report focusing on the questionable treatment of drift values used as part of determining instrument calibration tolerance, but drift is merely a component of the overall treatment of instrument uncertainty.
The current Davis-Besse safety-related Instrument Uncertainty/Setpoint methodology is limited, in its application, for the most part, to the population of instrument strings that provide automatic protective action (i.e., Limiting Safety System Settings -LSSSs).Numerous condition reports were initiated during the current restart effort (2002 -2003 Reactor Head Degradation Issue) on several facets of instrument uncertainty and the lack of a comprehensive instrument uncertainty program.Thus, the scope of this investigation has been expanded to also include all of the program/procedure level concerns associated with instrument uncertainty identified during the SHRR and LIR reviews.Consequences of event/condition investigated The issues imposed by the absence of a comprehensive Instrument Uncertainty/Setpoint application philosophy are manifest in that: 1. There is a possibility that the values recorded in Surveillance Test Procedures would not comply with the Technical Specification limits if instrument uncertainty were factored into the acceptance criteria.2. Guidance provided for instrument uncertainty, as it applies to indications, setpoints, alarms, and procedural decision points that are NOT automatic protective functions, is non-conservative with respect to current industry standards.
: 3. Operators use instruments to ensure that the plant is operating within the design and licensing basis. The guidance they use may not have instrument uncertainty factored in.4. Operators use instruments to make decisions about required actions during accident conditions in the EOPs.. The procedures they use may not have instrument uncertainty factored in.Compensatory and Remedial Actions taken None Root Cause Analysis Report for CR 02-06407 Problem Statement
* 2 Sitn: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-06407 CR Category:
Action Type: Schedule Type: CA Number: SR ( ) ( A ) Owner Assigned/Controlled 4 Corrective Action Type: Cause Code: Resp Org: ( CA ) Corrective Action (B06 ) Prog/process weak DES R
 
== Description:==
 
I Develop and publish a comprehensive instrument uncertainty/setpoint Design Criteria Manual G change that extends to all plant instrumentation.
N This policy should provide sufficient guidance to determine the level of rigor with which instrument A uncertainty consideration must be applied to an instrument, based on its safety significance.
The T classification of safety-significant instruments, and the application of uncertainty can be a "graded" O approach (TRd-ISA-67.04.09 can be used for guidance in this respect).
At a minimum, the policy R should address how instrument uncertainty is to be applied to:-Setpoints for RPS, ESFAS, SFRCS-Instruments required by Technical Specifications and the Technical Requirements Manual-Indications used for compliance with Technical Specifications and the Technical Requirements Manual-Instrumentation used for compliance with licensing basis documents, e.g., ODCM, COLR, RG 1.97, etc.-Instrumentation used for compliance with design basis documents (but NOT reflected in the Technical Specifications), e.g., ASME XI, calculations, etc.-Instrumentation used for critical decision points in emergency and abnormal operating procedures
-Instrumentation used to alert operators to degraded plant conditions, e.g., annunciators, computer alarms, etc.-Non safety-related instrumentation used to protect major plant equipment (i.e., economic considerations)
Short-term action required for safe restart and operation.
Recommend Restart Completed By: Organization:
Date: Phone: Attachments:
DES 3/8/2003 7497 ED Yes W] No If a Refueling Outage is required, Other Tracking # Corrective Action Due Date: ACC- Enter the Refueling Outage number: .A N/A 5/30/2003 EPT Approval: (Enter Name and Sign) Section: Date: DES 3/812003 QUAL Quality Organization Approval:
Date: ITY Page 6 of 38 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-06407 I Response:
Completed as witmn 0 Revised/Alternate Solution 0 Not Performed M DCM change 075 has been completed to address this item. (Approved 5/30/03)P L The policy document includes the Mission Statement, program description and scope of application E for the comprehensive Instrument Uncertainty Evaluation Program.M E Attachment 1, Graded Approach Classification, details the methodology for determining the level of N rigor required to evaluate various plant instrumentation and endorses the ISA methodology for T applying the safety significance approach.I N Attachment 2, Setpoint Methodology, details the Davis-Besse specific policy on the development of G instrument uncertainty evaluations and provides the current level of commitment to the governing industry standards.
It also provides clarifications, exceptions, and operating experience associated O with the development of of these evaluations.
R Alternate Corrective Action or Justification if Corrective Action not performed:
G Corrective Action Implementation Date: 5/30/2003 j Signature indicates Corrective Action complete: Completed By: Date: 5/23/2003 jj Signature indicates verification for SCAQ CRs: Verified By: Date: 5/30/2003 j Enter Name and Sign: Implementing Organization Approval:
Date: 5/30/2003 Q V Comments: U E A R L I I F T I Y E R Approval:
Date: Page 7 of 38 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-06407 CR Category:
Action Type: Schedule Type: CA Number: SR ( ) (A) Owner Assigned/Controlled 5 Corrective Action Type: Cause Code: Resp Org: ( CA ) Corrective Action ( '06) Prog/process weak DES R
 
== Description:==
 
I Identify Critical Restart Parameters.
G 1) Develop and publish screening guidance for. determination of instruments necessary for safe I plant startup and operation.
N 2) These are plant indications and measurements most critical to compliance with design and A licensing bases.T Develop population of Critical Restart Parameters per the approved screening guidance.0 R Short-term action required for safe restart and operation.
Recommend Restart Completed By: Organization:
Date: Phone: Attachments:
DBE 1/11/2003 7497 W Yes El No If a Refueling Outage is required, Other Tracking # Corrective Action Due Date: ACC- Enter the Refueling Outage number: N/A N/A 4/22/2003 EPT Approval: (Enter Name and Sign) Section: Date: DES 3/6/2003 QUAL Quality Organization Approval:
Date: ITY I Response:
Completed as wrtten 0 Revised/Alternate Solution 0 Not Performed M Action P L Identify Critical Restart Parameters.
E 1) Develop and publish screening guidance for determination of instruments necessary for safe M plant startup and operation.
E 2) These are plant indications and measurements most critical to compliance with design and N licensing bases.T Develop population of Critical Restart Parameters per the approved screening guidance.I N Short-term action required for safe restart and operation.
G Recommend Restart 0 R Action Taken G A comprehensive Action Plan was developed by the DEEC section to address the above requirements.
Its overall goal was to develop screening criteria to identify those plant indications and measurements most critical to compliance with design and licensing bases. The following provides a summary of the Action Plan: The population of parameters must be sufficient to provide reasonable assurance for safe plant startup and operation; i.e., the instruments chosen should be those most important to safety. This list should include indications and process measurements used for compliance with Technical Specifications and Technical Requirments Manual and test indications and process measurements used to demonstrate OPERABILITY.
For each identified critical restart parameter, perform an evaluation to: Page 8 of 38 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-06407 1.) Evaluate an "order of magnitude" total uncertainty.
3.) Determine "margin" available to trade for uncertainty 4.) Compare margin available to total uncertainty estimate 5.) Impose corrective actions as necessary Evaluation of instrument uncertainty and corrective action imposed should be sufficient to provide reasonable assurance for safe plant startup and continued operation.
In some instances, it may be necessary to estimate instrument errors such that they are conservatively bounding for worst-case plant conditions.
The intent of this evaluation is to show that the probability of encroaching on licensing and design basis limits, due to instrument uncertainty consideration, is acceptably small, and/or that the consequences of such an encroachment would be minimal. This evaluation will be performed in a manner consistent with the "graded" approach.Critical parameters, i.e. instrument strings determined to be significant enough to merit some consideration of instrument uncertainty, will be evaluated for things like: 1.) Has the instrument uncertainty/error associated with the parameter been appropriately quantified?
2.) Has analytical margin associated with the parameter been quantified?
3.) Is analytical margin appropriate and sufficient to account for instrument errors?4.) What actions are necessary to correct any deficiencies?
The ties between critical parameters, uncertainty calculations, design and licensing basis limits, and procedural use must be clearly documented and recoverable.
This will enable future calculation and procedure revisions to be performed with a clear understanding of a revision's impact on instrument uncertainty and the range of impacted documents.
This guidance was used in the evaluation of a population of restart critical parameters.
A task team was commissioned to complete the activity and address any required additional corrective actions.After application of the above guidance, a scope review resulted in a population of.110 Technical Specification and Technical Requirements to be evaluated.
For each evaluated parameter, the affected instruments used to perform the surveillance were identified and the degree of uncertainty associated with the measurement was qualitatively or quantitatively assessed.A database was developed to capture the results and to specify which parameters required detailed investigation and/or additional calculational support to demonstrate conservatism between the surveillance values and the Analytical Limits. The database was prepared by I&C/Electrical Engineering and reviewed by Nuclear Safety Analysis.The results and necessary Corrective actions have been'captured under Action #6.Alternate Corrective Action or Justification if Corrective Action not performed:
Corrective Action Implementation Date: 4/15/2003 j Signature indicates Corrective Action complete: Completed By: 40 , Date: 4/15/2003 j Signature indicates verification for SCAQ CRs: Page 9 of 38 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-06407 Verified By: Date: 4/17/2003 Enter Name and Sign: Implementing Organization Approval:
Date: 4/21/2003 Q V Comments: U E A R L I I F T I Y E R Approval : Date: Page 10 of 38 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-06407 CR Category:
Action Type: Schedule Type: CA Number: SR ( ) (A) Owner Assigned/Controlled 6 Corrective Action Type: Cause Code: Resp Org: ( CA ) Corrective Action (B06 ) Prog/process weak DES R
 
== Description:==
 
I (Review Critical Restart Parameters.
G I -Review for an "order of magnitude" total uncertainty N -Determine "margin" available to trade for uncertainty A -Compare margin available to total uncertainty estimate T -Impose corrective actions as necessary 0 R Review of instrument uncertainty and corrective action imposed should be sufficient to provide reasonable assurance for safe plant startup and continued operation.
In some instances, it may be necessary to estimate instrument errors such that they are conservatively bounding for worst-case plant conditions.
The intent of this evaluation is to show that the probability of encroaching on licensing and design basis limits, due to instrument uncertainty consideration, is acceptably small, and/or that the consequences of such an encroachment would be minimal. This evaluation will be performed in a manner consistent with the "graded" approach in the comprehensive instrument uncertainty/setpoint policy established by Corrective Action 3.Short-term action required for safe restart and operation.
Recommend Restart Completed By: Organization:
Date: Phone: Attachments:
DES 3/8/2003 7497 Yes Dl No If a Refueling Outage is required, Other Tracking # Corrective Action Due Date: ACC- Enter the Refueling Outage number: A N/A 5/8/2003 EPT Approval: (Enter Name and Sign) Section: Date: DES 3/8/2003 QUAL Quality Organization Approval:
Date: ITY I Response: Completed as wr/aten 0 Revised/Alternate Solution 0 Not Performed M Action P L Review Critical Restart Parameters.
E M -Review for an "order of magnitude" total uncertainty E -Determine "margin" available to trade for uncertainty N -Compare margin available to total uncertainty estimate T -Impose corrective actions as necessary I N Review of instrument uncertainty and corrective action imposed should be sufficient to provide G reasonable assurance for safe plant startup and continued operation.
In some instances, it may be necessary to estimate instrument errors such that they are conservatively bounding for worst-case 0 plant conditions.
The intent of this evaluation is to show that the probability of encroaching on R licensing and design basis limits, due to instrument uncertainty consideration, is acceptably small, G and/or that the consequences of such an encroachment would be minimal. This evaluation will be performed in a manner consistent with the "graded" approach in the comprehensive instrument uncertainty/setpoint policy established by Corrective Action 3.Page 11 of 38 Site" 2N91 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-06407 Action Taken A comprehensive review of 110 Technical Specifications and Technical Requirements was conducted per the above-described guidelines.
For each evaluated parameter, the affected instruments used to perform the surveillance were identified and the degree of uncertainty associated with the measurement was qualitatively or quantitatively assessed.A database was developed to capture the results and to specify which parameters required detailed investigation and/or additional calculational support to demonstrate conservatism between the surveillance values and the Analytical Limits. The database was prepared by I&C/Electrical Engineering and reviewed by Nuclear Safety Analysis.In all but 10 of the cases, it was found that conservative margin (in excess of the contribution of instrument uncertainty) existed between the surveillance and the Tech Spec value, within the safety analysis that formed the basis for the spec, or intrinsic to the operational characteristics of the parameter.
In the remaining 10 cases, this margin could not be readily demonstrated.
As a result, detailed calculations or engineering evaluations will need to be developed to quantify the applicable instrument uncertainty and establish new surveillance guidelines for these parameters.
This effort is currently underway.These parameters include: RCS Flowrate (CR 02-06885) t/Rated Thermal Power Allowance( CR 03-00970)
V HPI Flow (TS 4.0.5) (CR 02-04514)V*, LPI Flow (TS 4.0.5) (CR 02-04514)-
'd" BWST Volume (CR 02-05157)V Cc, 1 J -,4I C Containment Temperature (CR-03-03592)V/,UHS level (CR 02-05356) v &#xfd;, (/-')UHS Temperature (CR 02-05356)
-'BWST Boron Concentration (CR 03-02644) v/Core Flood Tank Volume (CR 03-02547)
V Detailed evaluation of each parameter is provided in the Notes entry in the database. (A printout from the database is an attachment)
Alternate Corrective Action or Justification if Corrective Action not performed:
Corrective Action Implementation Date: 5/8/2003 Signature indicates Corrective Action complete: Completed By: Date: 5/8/2003 Signature indicates verification for SCAQ CRs: Verified By: ' Date: 5/8/2003 Enter Name and Sign: Implementing Organization Approval:
Date: 518/2003 Page 12 of 38/,
Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-06407 Q V IComments:
U E A R L I I F TI Y E Approval:
Date:-a Page 13 of 38 Review Complete Fr- -- +,, o f Parameter Source Value Applicable Mode Rated Thermal Power TS 1.3 2772 MWt MODE 1 Est Inst Error Error Source Margin Margin Source unknown*32-1240384-01 2%IOCFR50 App. K +safety analyses Notes Surveillance Test Instrument(s)
Surveillance Limit CR 03-0970, There is no calculation, or error analysis, to ensure that the error associated with the determination of reactor thermal power, in procedure DB-NE-03230 is within the 2% instrumentation error assumed in safety analyses.As described in the Bases for TS 2.2.1, the high flux trip allowable value accounts for "transient overshoot, heat balance and instrument errors"; the instrumentation errors are assumed to be 2%.The following analyses use 102% RTP as an initial condition:
SG Blowdown Line Break (3.6.2.7.1.15)
Control Rod effects analysis (6.3.3.2.1)
Loss of Flow accident Locked Rotor accident Excessive Heat Removal accidents Steam line Break As required by I0CFR50, Appendix K, LOCA analyses also use 102% power.Also, the source term determination for extended cycles (USAR 15.A.7.0) uses 102% power as an assumption.
For the most part, although not called out specifically in every analysis, the 102% results from an "allowance of +2% for heat balance error".IOCFR50, Appendix B, Criterion Ill, requires that "measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions".
Contrary to this, there is no analysis that ensures that the secondary side heat balance determination of reactor thermal power has an error -within the 2%assumed within the design basis.DB-NE-03230 performs the heat balance, but does not contain any analysis of methodology error.Calculation 32-1240384-00 (revised in 1997by 32-1240384-01) provides an error equation for the secondary side heat balance, but does not calculate a total error specific to Davis-Besse.
This calculation states in its purpose that it is "only for measurement string errors"; the calculation also states that it is "for example purposes only". This calculation would need to be revised significantly in order to provide an adequate basis for ensuring that the DB-NE-03230 heat balance methodology is within the required 2%error.Calculation 32-5012428 was prepared by Framatome, in order to support the App. K Measurement Uncertainty Recapture power uprate. This calculation ensures that the secondary side heat balance determination of reactor thermal power is within the assumptions of the power uprate License Amendment Request (and supporting analyses).
The applicability of this calculation, however, depends on the use of the Caldon LEFM for feedwater flow measurement.
A calculation must be prepared to ensure that the error associated with the current methodology, for DB-NE-03230 various Thursday, May 08, 2003 Page 1 of 48 Parameter Source Value Applicable Mode Heat Trace Temperature TS 4.1.2.1 Est Inst Error Error Source Margin Margin Source 10 deg. F DB-ME-09521 15 deg. F License Amendment 67, Log No. 1519 N/A Notes Surveillance Test Instrument(s)
Surveillance Limit>/= 105 Deg. F 5&6 RCS Heatup/Cooldown Rate deterimining reactor thermal power, is within the bounds of the design basis. This calculation will be needed prior to Mode 1.The surveillance limit is to provide verification that the pipe temperature of the heat traced portion of the Concentrated Boric Acid Storage System is >/= to 105 deg F. The temperature limit is established to preclude the precipitation of the boric acid solution.
The crystallization temperature for a solution of 7 weight percent (14000 ppm) boric acid is about 95 deg F. This is 10 deg F lower than the surveillance limit and provides an allowance for any measurement errors/margin associated with the process. It is concluded that the surveillance limit is acceptable for this parameter.
Note that the improved Standard Tech Specs (NUREG-1430) do not require surveillance ofheat trace.The pressure-temperature limits of the reactor coolant pressure boundary are established in accordance with the requirements of Appendix G to 10 CFR 50. The maximum P/T during normal heatup and cooldown assures that the total stresses on the Reactor Vessel will not cause it to fail in a brittle fracture mode. The BAW methodology for compliance is documented in B&W Topical Report BAW- 10046A.The heatup is plotted using the wide range temperatures, and is observed through programmable points (driven from the plant computer) assigned to a Control Console recorder.Though it is possible for uncertainty to be introduced by the measuring element and the PPC/recorder, the relative nature of the measurement tends to cancel out any drift or accuracy components.
Likewise, the number of points collected (typically 30 minute intervals) provides a large data sample and subsequently improves accuracy.This, combined with the significant mechanical margins present in the Appendix G brittle fracture limits, ensures the conservatism of the 100 deg limit.DB-SC-03059 Doric Recorder or if inoperable, use local temperature indicators listed on pg 6 of procedure>/= 105 Deg. F DB-OP-6900 DB-OP-6903 DB-PF-3065 Not specified in the procedure(s)
Figure 3.4-2 & 3.4-3 & 3.4-4; 50 Deg in one hour heatup/ 100 deg. F in one hour cooldown when >/= 270 deg. F, 50 deg F cooldown in one hour when <TS 3.4.9.1 N/A Figure 3.4-2 & 3.4-3 & 3.4-4; 50 Deg in one hour heatup/ 100 deg. F in one hour cooldown when >/= 270 deg. F, 50 deg F cooldown in one hour when <ALL ALL Significant Brittle Fracture stresses SD-039A BAW-2011 32-1170722 32-1170331 32-1170726 Calc C-NSA-64.02-023 TE Ltr, Chlapowski (TE) to Walters (B&W), NEN-87-10306, File 064-02, 888-01 Thursday, May 08, 2003 Page 2 of 48 Parameter Source Value Applicable Mode Steam Generator Secondary Coolant Temperature Est Inst Error Error Source Margin Margin Source N/A Notes TS 4.7.2.1 N/A> 110 Deg. F 70 deg ALL Structural Analysis Steam Generator Secondary Pressure N/A N/A The SG brittle fracture limits are typically surveilled during the initial stages of plant heatup from Mode 5 to Mode 4. The verification of the TS limit is performed in the above NOP, and is included in the normal operations procedures for several related systems such as AFW.The assumptions used in the development of this limit typically contain significant mechanical conservatism with respect to RTNDT, existing flaws, service history, and stress concentration limits. The TS Bases alone contains a margin of 70 deg to the calculated RTNDT (which is itself conservative for a pressure vessel of this type.)Based on this and the fact that significant operational margin typically exists between the actual plant condition and the limits, no additional penalties for the measuring instruments need be assessed at this time.This conclusion is applicable to both the temperature and pressure limits.The SG brittle fracture limits are typically surveilled during the initial stages of plant hearup from Mode 5 to Mode 4. The verification of the TS limit is performed in the above NOP and is included in the normal operations procedures for several related systems such as AFW.The assumptions used in the development of this limit typically contain significant mechanical conservatism with respect to RTNDT, existing flaws, service history, and stress concentration limits.Based on this and the fact that significant operational margin typically exists between the actual plant condition and the limits, no additional penalties for the measuring instruaments need be assessed at this time.This conclusion is applicable to both the temperature and pressure limits.Nominal Value As noted for the SG pressure and Temperature limits: The assumptions used in the development of this limit typically contain significant mechanical conservatism with respect to RTNDT, existing flaws, service history, and stress concentration limits.Surveillance Test Instrument(s)
Surveillance Limit DB-OP-6900 T887-891 T903-907>110 deg with pressure above 237 psig.DB-OP-6900
<237 psig with temp <110 deg TS 4.7.2.1> 237 psig Significant Structural ALL T-Average TS 4.7.2.1< 200 Deg. F ALL N/A N/A N/A N/A N/A N/A N/A N/A DB-OP-6900 Sealed Source Contaminatiodi&#xfd; TS 3.7.8.1>/= 0.005 microcuries ALL Not related to plant startup or operation.
Thursday, May 08, 2003 Page 3 of 48 Parameter Source Value" Applicable Mode EDG Day Tank Volume TS 3.8.1.2.b.1
>/= 4000 gal.Est Inst Error Error Source Margin Margin Source N/A N/A 175 gal.(LI-1 12))/300 gal (LI-2787A or B)Surveillance N/A N/A 500 gal uncertainty 3000 gal operational Surveillance 5&6 EDG Fuel Storage Tank Volume Notes Surveillance Test Instrument(s)
Surveillance Limit TS 3.8.1.2.b.2
>/= 32,000 gal.5&6 Day Tank Volume -Value appears to be based on the ability to mitigate the initial stages of a LOCA without requiring makeup from the outside tank. The USAR capability statement of 20 hours of operation easily bounds this time period and appears to be arbitrarily based on the 6000 gallon tank delivered and the known vendor measured fuel consumption for the 4000 minimum level. There is no other analytical/accident basis for the 20 hour value.The Operations and Surveillance procedures require initiation of refill at 4750 gal., and margin is included in surveillance procedures to account for instrument inaccuracies that tnight be introduced by the level monitor. Margin is also included in the calculation (C-ME-024.01-005, Rev. 1), in the form of conservatisms.
Engineering judgment says that the combination of "margins" is sufficient to conservatively cover the existing instrument error.Outside Tank Volume -generally sized to allow a seven day supply though there are no specific requirements in the existing Reg. Guide for load carried or the duration to be carried. Since fuel economy is variable, but certainly less than the assumed 110% load factor, the sizing is a conservative design.Based on existing surveillance performance, the nominal 32K gal capacity is deemed adequate for LOCA/LOOP performance and when combined with the 4000 day tank, provides 7-day full load capability that is significantly greater than a design basis accident scenario requirements.
Margin is included in surveillance procedures to accomodate instrument inaccuracies and level is administratively controlled at greater than 35000 gal. Margin is also included in the calculation (C-ME-024.01-005, Rev. 1), in the form of conservatisms.
Engineering judgment says that the combination of "margins" is sufficient to conservatively cover any instrument error introduced by the level indication loop.CR 02-03362.Nominal Value -The load limit is set as practical at or below the TS load limit within the capabilities of the Dillon Cell. The load limit is a practical design value, not specifically related to any safety analyses or accident analyses.DB-SC-3070, 3071, 3076, 3077 LI-4891, 4892 35000 refill requirement DB-SC-3070, 3071, 3076, 3077 LI-I 121 LI-2787A1B 4000 gal.Control Rod Hoist Overload TS 4.9.6.1<= 2650 pounds During movement of control Fuel Assembly Hoist Overload TS 4.9.6.2</= 2700 pounds During movement of control N/A N/A N/A NIA N/A N/A N/A N/A Nominal Value -The load limit is set as practical at or below the TS load limit within the capabilities of the Dillon Cell. The load limit is a practical design value, not specifically related to any safety analyses or accident analyses.Thursday, May 08, 2003 Page 4 of 48 Parameter Source Value Applicable Mode DHR Flow Est Inst Error Error Source Margin Margin Source N/A N/A Notes Surveillance Test Instrument(s)
Surveillance Limit TS 4.9.8.L.a>/= 2800 GPM Mode 6 when >/= 23 feet of DHR Core Outlet Temperature TS 4.9.8.l.b</= 140 Deg. F Mode 6 when >/= 23 feet of Operations procedure N/A N/A Significant Analysis Refueling Water Level TS 3/4.9.10>/= 23 Feet N/A N/A N/A N/A The purposes of the DHR System in MODE 6 are to remove decay heat and sensible heat from the Reactor Coolant System (RCS), as required by GDC 34, to provide mixing of borated coolant, to provide sufficient coolant circulation to minimize the effects of a boron dilution accident, and to prevent boron stratification.
The 2800 gpm value is equivalent to the normal low end range of DHR flows used in the higher modes and is based on maintaining the above requirements as well as sweeping the suction line of air accumulations.
It is not necessary to correct this nominal value for additional uncertainty, since it contains a conservative estimate of minimum recirculation requirements (a typical value is approx. 1500 gpm to provide necessary boron mixing).Maintenance of the core outlet at a temperature below 140 F provides the nominal starting point for the time-to-boil analysis.
Ifthe reactor coolant temperature is not maintained below 200*F, local boiling of the reactor coolant could result and lead to inadequate cooling of the reactor fuel and a possible reduction in boron concentration.
The evaluation contains numerous conservatisms with respect to decay heat fraction and heat release (no credit for surface steaming) to offset any small amount of instrument error.The fuel handling accident analyses assume that 100% of the fuel assembly gap activity (99% of iodine is retained in the water) escapes from the fuel assembly and from the cavity. Thus, the accident analyses are relatively insensitive to the fuel pool water level. This is not a restart critical parameter.
it is clear that significant conservatisms exist with respect to the iodine scrubbing analysis, gap activity released, bumup history, and enrichment.
In the case of the cavity, the actual level over the fuel is on the order of 45 feet when the cavity level is flooded to the nominal 23 ft. value. The SFP level does not contain this margin but is only a challenge to the analysis during the movement of irradiated fuel.Since this activity has been completed, no additional evaluation is required for this parameter prior to restart.The fuel handling accident analyses assume that 100% of the fuel assembly gap activity (99% of iodine is retained in the water) escapes from the fuel assembly and from the cavity. Thus, the accident analyses are relatively insensitive to the fuel pool water level. This is not a restart critical parameter.
The surveillance test takes a physical measurement from the SFP water surface to the pool lip and verifies that the measurment, conservatively rounded, is less than or equal to 32 in. Instrument error is estimated on the assumption that a tape measure, or rule, with legible graduations of+/- 1/8 in. is used. Margin assumes that the 2 inches assumed for fuel pin growth is conservatively bounding and that the measured value is rounded up at least 1/8 in. This issue should be revisited post restart.DB-NE-03292 tape measure Logsheet FYI DH2A 0-5000 gpm.50 gal. incr.>2800 gpm.</= 33 in. between water level and transfer canal lip Mode 6 during movement of SFP Water level TS 3/4.9.11 N/A N/A N/A N/A>/= 23 feet Whenever irradiated fuel Thursday, May 08, 2003 Page 5 of 48 Parameter Source Value Applicable Mode Liquid holdup Tank Radioactivity Content TS 3/4.11.1<1= 10 curies At all times Est Inst Error Error Source Margin Margin Source N/A N/A N/A N/A Notes Surveillance Test Instrument(s)
Surveillance Limit Not Mode Related -While important to safe plant operation, this parameter is not specifically critical to plant restart. This issue will be revisited post-restart.
Explosive Gas mixture TS 3.11.2 Oxygen </= 2% by volume when Hydrogen > 4% by volume At all Times N/A N/A N/A N/A N/A N/A Not Mode Related -While important to safe plant operation, this parameter is not specifically critical to plant restart. This issue will be revisited post-restart.
Boric Acid Addition Tank Volume TRM 3.1.2.8.a.1
>/= 900 gal.5&6 12 gal C-NRE-040.01-003 The Boric Acid Addition System is credited for a tornado event. The limit on volume provides sufficient boric acid to provide shutdown from 200F to 75 F.900 gallons is at the threshold of the lower range for tank level instrumentation.
It is unlikely that the tank would be maintained at a level that is equivalent to "off-scale low". The tank volume capacity is 7369 gallons. Due to the low significance of the tank level function, and the low probability that tank level will be below the limit, there is reasonable asssurance that this parameter is acceptable fro restart.The estimate of boron sampling accuracy is based on discussions with Chemistry personnel and considers the concentration range in question.DB-CH-03027 LI-MU-49-2, LI-MU-65-2
>/= 900 gal Boric Acid Addition Tank Concentration TRM 3.1.2.8.a.2
>/= 7875 ppm and </= 13,125 ppm 5&6+/- 20 ppm DB-CH-03027 Sample analyzer The TRM Bases provides the minimum indicated values in the Table, and it varies according to the contained volume. The analytical basis assumes very conservative conditions of fuel reactivity, xenon worth (none), approx. 300 ppm and low temperature.
Also, as stated in the TRM bases, the values have been conservatively increased to account for instrument and chemical analysis tolerances.
TRM Although not quantified here, the combination of the above conservatism provides a reasonable assurance that the BAT tank limits will ensure sufficient SDM capability.
>/= 7875 ppm and </= 13,125 ppm Thursday, May 08, 2003 Page 6 of 48 Parameter Source Value Applicable Mode Boric Acid Addition Tank Solution temperature TRM 3.1.2.8.a.3
>/= 105 deg. F.5&6 Est Inst Error Error Source Margin Margin Source 5 deg.Notes Surveillance Test Instrument(s)
Surveillance Limit see notes 10 deg.see notes The Boric Acid Addition System is credited for a tornado event. The limit on volume provides sufficient boric acid to provide shutdown from 200F to 75 F.Keeping the temperature greater than 105 degrees ensures that the tank volume remains soluable.The temperature limit is established to preclude the precipitation of the boric acid solution.
The crystallization temperature for a solution of 7 weight percent boric acid is about 95 deg F. This is 10 deg F lower than the surveillance limit and provides an allowance for any measurement errors/margin associated with the process. It is concluded that the surveillance limit is acceptable for this parameter.
TI-MU-48-1, TI-MU-64-1
>/= 105 deg. F.DB-CH-03027 Estimated error: Engineering judgement BWST Volume TRM 3.1.2.8.b.1
>/= 3000 gal.5 inches Indicator Significant
(-30,000 gal)Surveillance Log 5&6 BWST Concentration TRM 3.1.2.8.b.2
>/= 2600 ppm 5&6 26 ppm 38-1290170-00
>200 ppm 51-5016651-00 Based on review of the Calc. C-ICE-48.01-004, a minimum of approx. 54 inches (4.5 fi) is required to meet the minimum requirement for 3000 gal. volume and allow NPSH margin.The current tank volume of 33 ft. easily bounds this requirement.
At the time the refueling cavity is filled, there is 6.9 fl. remaining.
Even this low levels equates to approx. 30,000 gal of inventory to the lower limit even when considering instrument uncertainty.
Estimated error: Boron titration error is estimated to be a maximum of 26 ppm in the memo documented in 38-1290170-00.
Margin: Mode 5 & 6 BWST boron concentration is required in order to provide an available source of borated water to provide 1% shutdown (TRM Bases). Calculation 51-5016651-00 (supplemented by rev. 1, 51-5016651-01), "DB CY14 Boron calculations", determines that therequired BWST concentration to maintain a 1% dK/K during post-LOCA conditions is 2312 ppm. This calculatio his bounding for the shutdown requirements of Modes 5&6. This gives a margin of 2600-2312
= 288 ppm.DB-OP-03004 Chemistry M&TE>/= 2600 ppm LI-1525 A-D Thursday, May 08, 2003 Page 7 of 48 Parameter Source Value Applicable Mode BWST solution temperature Est Inst Error Error Source Margin Margin Source+/- 4 deg.Notes Surveillance Test Instrument(s)
Surveillance Limit TRM 3.1.2.8.b.3
>/=35 deg. F.5 &6-15 deg.Operational RCS Chemistry limits TRM 3.4.7, Table 3.4-2 DO </= 0.10 ppm/Chloride
</=0.15 ppm/ Fluoride </= 0.15 ppm At all times Pressurizer Heatup/Cooldown Rate TRM 3.4.9.2.a 100 Deg. F in one hour At all times NA N/A N/A N/A This value is used in the Inadvertant Containment Spray analysis -which assumes 35 deg. injection temperature.
The existing surveillance requirement does not include loop uncertainty.
During Modes 5&6, there is not sufficient heating within the Containment to create the conditions for a damaging inadvertant spray. Also, the CS system and its actuation circuitry is not required to be operable in these modes (though they may be functional).
As a result, the only basis in the lower modes is the prevention of freezing.
The current operating procedure for the BWST requires steam heating to be placed in service below 50F as well as heat tracing on external piping. These actions, combined with the fact that even under severely cold external temperatures the tank requires weeks to cool down to the 35F limit, provide adequate protection for the operability limits on the BWST.Crystallization is also not a concern at this temperature since the solubility limit of a boric acid solution at 35 Deg. F is approximately 4900 ppm: From Langes Handbook of Chemistry, 13th Edition (McGraw-Hill 1985), the solubility of Boric Acid(H3BO3) is: 2.67g/100g solution@
0 Deg. C (32 Deg. F) = 4668 ppm No applicable LIR CRs found DB-CH-01002 EPRI Primary Water Chemistry Guidelines Reactor Coolant System chemistry is maintained with the limits referenced in the TRM 3.4.7, Thl 3.4-1.Chemistry limits are based on EPRI pwr water chemistry guidelines which established consistent standards, NSSS vendor recommendations, and good practice.
There are no analytical limits associated with these surveillance values. The limits help to ensure the maintenance ofthe proper RCS chemistry and provide an environment that is compatible with reactor coolant materials.
It is concluded that the inclusion of additional uncertainties for the surveillance limits is not necessary since the chemistry limits have been conservatively established based on consistent industry standards.
The pressurizer heatup is plotted using the wide range liquid and/or vapor space temperatures.
The data is observed through programmable points (driven from the plant computer) assigned to a Control Console recorder.Though it is possible for uncertainty to be introduced by the measuring element and the PPC/recorder, the relative nature of the measurement tends to cancel out any drift or accuracy components.
Likewise, the number of points collected (typically 30 minute intervals) provides a large data sample and subsequently improves accuracy.This, combined with the significant mechanical margins present in the Appendix G brittle fracture limits, DB-CH-03001 Sampling lAW DB-CH-06002 Within steady state limits shown and transient limits of DO </= 1.00 ppm/Chloride
</=1.50 ppm/ Fluoride <1= 1.50 ppm>35 deg N/A N/A Significant Brittle Fracture stresses DB-OP-6900 Recorder/PPC Thursday, May 08, 2003 P Page 8 of 48 Parameter Source Value Applicable Mode Pressurizer Spray Differential Temperature TRM 3.4.9.2.b</= 410 Deg. F Est Inst Error* Error Source Margin Margin Source+/- 5 deg WR Temp Approx. 160 deg Notes Surveillance Test -Instrument(s)'
Surveillance Limit At all times Operational Pressurizer Minimum Temperature TRM 3.4.9.2.c>/= 120 Deg. F when Pressurizer pressure >/= 625 psig At all times Reactor Core Safety limits TS 2.1,1, Figure 2.1-1+/- 5 deg WR Temp Indic.Significant Operational N/A N/A Since DB does not have an auxiliary spray system, it is unlikely that this limit could ever be approached during normal heatup and cooldown conditions.
The operating procedure specifies a limit of 250 deg. when spray is available from a running RCP, and the combination of RCS pressure and temperature required to violate the limit would almost certainly result in a violation of the P/T limits. Even when using DHR spray at SD condition, the temperature difference even assuming l0OF drop across the DHR HX would not be sufficient to violate the limit.As a result, it is concluded that the current operational guidance and plant configuration provide sufficent conservatism to preclude consideration of additional instrument uncertainty.
Current operational methodology precludes consideration of this value. The pressurizer is currently brought to operability with a bubble in Mode 5. Upon energization of the pressurizer heaters, shell temperature is raised to approximately 220 deg., where bubble formation occurs at a pressure of around 20-50 psig. RCP starting is then commenced with a full bubble.As such, the cold overpressure limit is not approached.
Since there is no provision for solid plant RCP starting, the possibility of a cold pressure spike is eliminated.
The safety limit of temperature/pressure depicted in Figure 2.2-1 is not verified using installed Control Room indicators but rather is ensured by proper performance of the RPS system activating at its designated LSSS setpoints.
The selection of low/hi reactor coolant pressure, reactor p/t, and hi reactor coolant temperature reactor trip setpoints ensures that sufficient margin exists between the maximum operating envelope of the RCS and the safety limit. The LSSS settings have been evaluated in detailed setpoint calculations, which are not the subject of this review.The safety limit of RCS pressure is not verified using installed Control Room indicators but rather is ensured by proper performance of the RPS and RCS overpressure protection systems activating at their designated setpoints.
The selection of the HI RCS pressure reactor trip, PORV, and Pressurizer Code Safety Valve settings (which are all less than the above limit) ensure that sufficient margin exists between the maximum operating envelope of the RCS and the safety limit. The RPS settings have been evaluated in detailed setpoint calculations, which are not the subject of this review. The overpressure protection devices have been tested and installed to tolerances which provide conservative margin to the limit.Should a situation arise which would challenge the safetylimit (such as a loss of feedwater ATWS) the fast nature of the transient is such that only a recorder/computer log could capture the necessary data to do the comparison.
The accuracy of the Wide Range Pressure loop is approx. I%, which is adequate to evaluate such a transient during the post-trip review.DB-OP-6900 DB-OP-6900 Per the Table I &2 RCS Pressure 30# loop WR pressure transmitter loop and PPC TS 2.1.3</= 2750 psig 1,2,3,4 & 5 Thursday, May 08, 2003 Page 9 of 48 Parameter Source Value Applicable Mode DHR Flow Est Inst Error Error Source Margin Margin Source N/A Notes Surveillance Test Instrument(s)
Surveillance Limit TS 4.1.1.2.b>/= 2800 GPM N/A ALL RCS Minimum Temperature for Criticality TS 4.1.1.4>/= 525 Deg. F l&2 N/A N/A N/A N/A The purposes of the DHR System are to remove decay heat and sensible heat from the Reactor Coolant System (RCS), as required by GDC 34, to provide mixing of borated coolant, to provide sufficient coolant circulation to minimize the effects of a boron dilution accident, and to prevent boron stratification.
The 2800 gpm value is equivalent to the normal low end range of DHR flows used in the higher modes and is based on maintaining the above requirements as well as sweeping the suction line of air accumulations.
It is not necessary to correct this nominal value for additional uncertainty, since it contains a conservative estimate of minimum recirculation requirements (a typical value is approx. 1500 gpm is necessary for boron mixing).Nominal Value -As described in NUREG-1430 Bases: Establishing the value for the minimum temperature for reactor criticality is based upon considerations for: a. Operation within the existing instrumentation ranges and accuracies;
: b. Operation with reactor vessel above its minimum nil ductility reference temperature when the reactor is critical.The reactor coolant moderator temperature coefficient used in core operating and accident analysis is typically defined for the normal operating temperature range (532oF to 579oF). The Reactor Protection System (RPS) receives inputs from the narrow range hot leg temperature detectors, which have a range of 520oF to 620oF. The integrated control system controls average temperature (Tavg) using inputs ofthe same range. Nominal Tavg for making the reactor critical is 532oF. Safety and operating analyses for lower temperatures have not been made.APPLICABLE There are no accident analyses that dictate the minimum SAFETY ANALYSES temperature for criticality, but all low power safety analyses assume initial temperatures near the-5!25oF limit (Ref. 1).LCO The purpose of the LCO is to prevent criticality outside the normaloperating regime (532oF to 579oF)and to prevent operation in an unmanalyzed condition.
The LCO limit of 525oF has been selected to be within the instrument indicating range (520oF to 620oF).TS 4.1.2.2.a SD-037A, SD-048 The surveillance limit is to provide verification that the pipe temperature of the heat traced portion of the Concentration Boric Acid Storage System is >/= to 105 deg F. The temperature limit is established to preclude the precipitation of the boric acid solution.
The crystallization temperature for a solution of 7 weight percent boric acid is about 95 deg F. This is 10 deg F lower than the surveillance limit and provides an allowance for any measurement errors/margin associated with the process. It is concluded that the surveillance limit is acceptable for this parameter.
DB-OP-06900 DB-OP-06001 FYI DH2A 0-5000 gpm.50 gal. incr.>/= 2800 GPM DB-OP-6912
>525 Deg. F Heat Trace Temperature TS 4.1.2.2.a>/= 105 Deg. F 1,2,3, & 4 N/A N/A Adequate margin (not quantified)
Eng Judgement DB-SC-03059 Doric Recorder, or if inoperable, use local temperature indicators listed on pg 6 of procedure>/= 105 Deg. F Thursday, May 08, 2003 Page 10 of 48 Parameter Source Value Applicable Mode Individual Control Rod Position TS 4.1.3.1.1+/- 6.5% of group average height 1&2 Est Inst Error Error Source Margin Margin Source 1.5%Bases 3/4.1.3 see notes BAW-10179P, Rev.6 Notes Surveillance Test Instrument(s)
Surveillance Limit Bases 3/4.1.3, SD-0049, DB-OP-03006 From BAW-10179P, Rev. 6: "Regulating rod position (rod index), APSR position, axial power imbalance, and quadrant power tilt are process variables that together characterize and control the three-dimensional power distribution in the reactor core... During the reload safety evaluation, a three-dimensional power distribution analysis is performed to set operating limits that will preserve the accident initial condition assumptions for the ECCS, loss of flow, ejected rod worth, and SDM analyses...
The power distribution analysis is based on calculating the three-dimensional peaking distribution over the range of potential power operation.
Parameterization
.on power level, cycle bumup, control rod and APSR positioning, and xenon distribution is included in the analysis.
The simulation of xenon transients.. .is further augmented by APSR motion in order to generate conservative (higher) peaking factors...
Correlations between peaking margins and axial offset, regulating rod insertion, and APSR position are generated from this database, and the cycle-specific LCO limits are determined...
The limits on regulating rod insertion, axial power imbalance, and quadrant power tilt given in the COLR typically represent the measurement system-dependent limits beyond which the core power distribution and reactivity criteria could be violated during an accident Measurement system-independent limits are obtained directly from the reload safety evaluation analysis, without adjustment for measurement system error and uncertainty.
Adjustments'are applied to account for uncertainties in the indicated regulating rod position, APSR position, axial power imbalance, and quadrant power tilt due to their measurement systems...
Regulating rod insertion and APSR insertion measurement system-dependent limits are derived by adjusting the measurement system-independent limits to allow for thermal power level uncertainty and rod position errors...The axial imbalance measurement system-dependent limits are determined from a statistical combination of incore system measurement and observability errors. The imbalance error adjustment is formulated such that when the measured imbalance is within the error adjusted limits, the true axial imbalance is within the unadjusted limit at a 95/95 probability/confidence level for the limiting core conditions.
The quadrant tilt measurement system-dependent limits are determined from a statistical combination of incore system measurement and observability errors. The tilt error adjustment is determined such that when the measured tilt is equal to or less than the error adjusted limit, the peaking increase due to tilt is equal to or less than the peaking increase used in the power distribution analysis at a 95/95 probability/confidence level." Summary: The effect of rod position, APSR position, axial power imbalance, and quadrant power tilt on core peaking factors is analyzed on a cycle specific basis. The resultant LCO limits, as presented in the COLR and TS, are adjusted for measurement errors. Even though the quantitative value of these errors is not explicit here, the reload analysis methodology described here provides reasonable assurance that instrument Control rod absolute position indicators Thursday, May 08, 2003 Page 11 of 48 Parameter Source Value Applicable Mode APSR Position Est Inst Error Error Source Margin Margin Source Notes Surveillance Test Instrument(s)
Surveillance Limit Bases 3/4.1.3, SD-0049, From BAW-10179P, Rev. 6: DB-OP-03006 TS 4.1.3.2.1+/- 6.5% of group average height I&2"Regulating rod position (rod index), APSR position, axial power imbalance, and quadrant power tilt are process variables that together characterize and control the three-dimensional power distribution in the reactor core... During the reload safety evaluation, a three-dimensional power distribution analysis is performed to set operating limits that will preserve the accident initial condition assumptions for the ECCS, loss of flow, ejected rod worth, and SDM analyses...
The power distribution analysis is based on calculating the three-dimensional peaking distribution over the range of potential power operation, Parameterization on power level, cycle bumup, control rod and APSR positioning, and xenon distribution is included in the analysis.
The simulation of xenon transients..
is further augmented by APSR motion in order to generate conservative (higher) peaking factors...
Correlations between peaking margins and axial offset, regulating rod insertion, and APSR position are generated from this database, and the cycle-specific LCO limits are determined...
The limits on regulating rod insertion, axial power imbalance, and quadrant power tilt given in the COLR typically represent the measurement system-dependent limits beyond which the core power distribution and reactivity criteria could be violated during an accident Measurement system-independent limits are obtained directly from the reload safety evaluation analysis, without adjustment for measurement system error and uncertainty.
Adjustments are applied to account for uncertainties in the indicated regulating rod position, APSR position, axial power imbalance, and quadrant power tilt due to their measurement systems...
Regulating rod insertion and APSR insertion measurement system-dependent limits are derived by adjusting the measurement system-independent limits to allow for thermal power level uncertainty and rod position errors...The axial imbalance measurement system-dependent limits are determined from a statistical combination of incore system measurement and observability errors. The imbalance error adjustment is formulated such that when the measured imbalance is within the error adjusted limits, the true axial imbalance is within the unadjusted limit at a 95/95 probability/confidence level for the limiting core conditions.
The quadrant tilt measurement system-dependent limits are determined from a statistical combination of incore system measurement and observability errors. The tilt error adjustment is determined such that when the measured tilt is equal to or less than the error adjusted limit, the peaking increase due to tilt is equal to or less than the pealing increase used in the power distribution analysis at a 95/95 probability/confidence level." Summary: The effect of rod position, APSR position, axial power imbalance, and quadrant power tilt on core peaking factors is analyzed on a cycle specific basis. The resultant LCO limits, as presented in the COLR and TS, are adjusted for measurement errors. Even though the quantitative value of these errors is'not explicit here, the reload analysis methodology described here provides reasonable assurance that instrument Thursday, May 08, 2003 Page 12 of 48 Parameter Source Value Applicable Mode Absolute vs. Relative Rod Position (API)TS 4.1.3.3+/- 3.46% between RPI and API channels l&2 Est Inst Error Error Source Margin Margin Source See Caic Calc 32-1176260-01 N/A Calc 32-1176260-01 Notes Surveillance Test Instrument(s)
Surveillance Limit Calculation 32-1176260-01 establishes the surveillance criterion fdr Tech Spec 4.1.3.3. The criterion for the comparison of the API and RPI systems for a given control rod was.determined to be +/- 3.46%. This accounts for systems uncertainties and rod index uncertainty of 1.5%. The surveillance limit calculated included uncertainties for both the RPI and API, computer uncertainty, the assumption of one failed switch, and has a probability level of two standard deviations, (2. sigma). It is concluded that all appropriate uncertainties and conservatism have been accounted for relative to the comparison of API and RPI positions for this surveillance parameter.
DB-OP-03006 API, RPI+/- 3.46% between RPI and API channels C Thursday, May 08, 2003 Page 13 of 48 Parameter Est Inst Error Notes Surveillance Test Source Error Source Instrument(s)
Value Margin Surveillance Limit Applicable Mode Margin Source Rod Insertion Limits see notes DB-NE-03220, SD-049, COLR, BAW-1 0179P DB-OP-03006 TS 3.1.3.6 requires that the regulating rod position be maintained above the limits given in the COLR.TS 4.1.3.6 see notes These limits include errors for both the heat balance and control rod position.
This is noted in the COLR, page C-5 through C-8, Figures L.a through I .d. The analytical methods used to determine the core operating Within limits provided in COLR see notes limits are found in BAW-10179P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Within limits provided in COLR Analyses".
I & 2 BAW-10179P, Rev.6 SD-049 Section 10 or SD-049, Position Indications, describes the absolute position indicator and the relative position indicator including an accuracy for the analog signals and an approximate accuracy for the readouts.There are no references in this section of the System Description for these values. API accuracy of the analog signal is +/- 1.1% which produces a readout of approximately
+/-2.1 % accuracy.
RPI analog signal accuracy is +/- 0.7% producing a position readout of +/- 1.7% accuracy.
The error terms are given as %accuracy and not described in terms such as % full scale or % reading. Vendor manual M-515-0061-12 provided a specification for meter accuracy of 2% full scale with meter scale graduations in percent (withdrawn)
Rod Insertion Limits (RILs) are imposed in order to preserve the Shutdown Margin (SDM) and ejected rod worth assumptions of safety analyses.From BAW-10179P, Rev. 6: "Limits on the rod index versus power level are determined for each reload to ensure that the SDM obtained from the worth of the scrammable control rods is equal to or greater than that assumed in the safety analyses.The SDM obtained by the control rod system is the difference between (I) the total available rod worth and (2) the total required rod worth... The total available rod worth is determined by reducing the calculated total rod worth by the worth of the maximum stuck CRA and the applicable uncertainties.
Examples of the applicable uncertainties include the uncertainty associated with the accuracy of the total rod worth calculations and the effects of control rod poison depletion...
Testing during the startup of each reload cycle confirms the validity of this uncertainty.
The total required rod worth is comprised of the the power deficit, the worth of the rods inserted prior to trip, and the applicable uncertainties.., an example of the an applicable uncertainty would be an allowance to bound deviations in the actual core conditions versus those assumed in the calculations...
The difference between the total available rod worth and the total required rod worth is the SDM... the power level and the control rod insertion prior to the reactor nrip are the two components of the SDM that can be readily controlled by reactor operators.
Therefore, limits are placed on rod index versus power level to ensure that the SDM requirements of the safety analyses are always met....core conditions used and the uncertainties applied in the calculations ensure that the rod insertion limits are conservative for ensuring that the ejected rod worth is less than or equal to the values demonstrated to be acceptable in the safety analyses.. .The ejected rod worth is calculated by determining the difference in reactivity between the core configuration prior to the rod ejection and the core configuration after the rod is ejected. When appropriate, uncertainties are applied to the worth of the ejected rod..." Summary: The effect of rod insertion limits is analyzed on a cycle specific basis. The resultant LCO limits, as presented in the COLR and TS, are adjusted for measurement errors. Even though the quantitative value of Thursday, May 08, 2003 Page 14 of 48 Parameter Source Value Applicable Mode APSR Insertion Limits Est Inst Error Error Source Margin Margin Source Notes Surveillance Test Instrument(s)
Surveillance Limit N/A N/A these errors is not explicit here, the reload analysis methodology described here provides reasonable assurance that instrument uncertainty has been adequately considered.
Thus, RIL is acceptable for restart, with no further action.DB-NE-03220, SD-049, COLR, BAW-1 0179P TS 3.1.3.9 requires that the APSR position be maintained within the physical insertion limits given in the COLR. There are no limits imposed on APSR insertion by the COLR(Figure 3). Thus, safety analyses are not sensitive to APSR'insertion limits.TS 4.1.3.9 DB-OP-03006 Within limits provided in COLR Within limits provided in COLR N/A I &2 N/A Thursday, May 08, 2003 Page 15 of 48 Parameter Source Value Applicable Mode Axial Power Imbalance Est Inst Error Error Source Margin Margin Source Notes Surveillance Test Instrument(s)
Surveillance Limit From BAW-10179P, Rev. 6: TS 4.2.1 Within limits provided in COLR Mode I Above 40% RTP C-ICE-58.01-008, Rev. 2 see notes BAW-10179P, Rev.6"Regulating rod position (rod index), APSR position, axial power imbalance, and quadrant power tilt are process variables that together characterize and control the three-dimensional power distribution in the reactor core... During the reload safety evaluation, a three-dimensional power distribution analysis is performed to set operating limits that will preserve the accident initial condition assumptions for the ECCS, loss of flow, ejected rod worth, and SDM analyses...
The power distribution analysis is based on calculating the three-dimensional peaking distribution over the range of potential power operation.
Parameterization on power level, cycle bumup, control rod and APSR positioning, and xenon distribution is included in the analysis.
The simulation of xenon transients...
is further augmented by APSR motion in order to generate conservative (higher) peaking factors...
Correlations between peaking margins and axial offset, regulating rod insertion, and APSR position are generated from this database, and the cycle-specific LCO limits are determined...
The limits on regulating rod insertion, axial power imbalance, and quadrant power tilt given in the COLR typically represent the measurement system-dependent limits beyond which the core power distribution and reactivity criteria could be violated during an accident.
Measurement system-independent limit ts are obtained directly from the reload safety evaluation analysis, without adjustment for measurement system error and uncertainty.
Adjustments are applied to account for uncertainties in the indicated regulating rod position, APSR position, axial power imbalance, and quadrant power tilt due to their measurement systems...
Regulating rod insertion and APSR insertion measurement system-dependent limits are derived by adjusting the measurement system-independent limits to allow for thermal power level uncertainty and rod position errors...The axial imbalance measurement system-dependent limits are determined from a statistical combination of incore system measurement and observability errors. The imbalance error adjustment is formulated such that when the measured imbalance is within the error adjusted limits, the true axial imbalance is within the unadjusted limit at a 95/95 probability/confidence level for the limiting core conditions.
The quadrant tilt measurement system-dependent limits are determined from a statistical combination of incore system measurement and observability errors. The tilt error adjustment is determined such that when the measured tilt is equal to or less than the error adjusted limit, the peaking increase due to tilt is equal to or less than the peaking increase used in the power distribution analysis at a 95/95 probability/confidence level." Summary: The effect of rod position, APSR position, axial power imbalance, and quadrant power tilt on core peaking factors is analyzed on a cycle specific basis. The resultant LCO limits, as presented in the COLR and TS, are adjusted for measurement errors. Even though the quantitative value of these errors is not explicit here, the reload analysis methodology described here provides reasonable assurance that instrument DB-OP-03006 DB-NE-03220 (manual calc of Axial Pwr Imbal with inop alarm Computer, or alternately manually using Within limits provided in COLR Thursday, May 08, 2003 Page 16 of 48 Parameter Source Value Applicable Mode Hot Channel Factors TS 4.2.2.1/4.2.3.1 Within limits provided in COLR MODE I Above 15% RTP Est Inst Error Error Source Margin Margin Source Notes Surveillance Test Instrument(s)
Surveillance Limit see notes Measurement Error is accounted for per TS 4.2.2.2/4.2.3.2 As described in the TS Bases, and in BAW-10179P, Rev. 6: 32-9130-00 DB-NE-03222 Incore Detectors
+ computer program Within limits provided in COLR 7.5% for FQ, 5% for FNdH TS 4.2 + Bases Hot Channel (power peaking) Factors are not a directly observable parameter.
The reload safety analyses determine that peaking factors will be maintained within their limits when the observable parameters of Rod position, rod insertion, quadrant power tilt, and axial power imbalance are maintained with their limits.Sufficient consideration of uncertainty is built into the limits on observable parameters to give reasonable assurance that there is not a restart safety concern, with respect to instrument uncertainty, for these parameters (see the parameters:
Axial Power Imbalance, Quadrant Power Tilt, APSR position, Individual Control Rod Position, Rod Insertion Limits, and APSR Insertion Limits).When power peaking factors are measured via the incore detectors, then uncertainty is applied to account for measurement error. Measurement uncertainty for the peaking factors was determined in B&W document 32-9130-00, "Measurement Uncertainty for FQ and FNdI-', 6/13/78. This document cannot be located.Although the uncertainty associated with the incore determination of peaking factors is not specifically known, the current methodology accounts for measurement uncertainty.
There is reasonable assurance that there is not a restart safety concern, with respect to instrument uncertainty, for the determination of hot channel factors.Thursday, May 08, 2003 Page 17 of 48 Parameter Source Value Applicable Mode Quadrant Power Tilt Est Inst Error Error Source Margin Margin Source Notes Surveillance Test Instrument(s)
Surveillance Limit From BAW-10179P, Rev. 6: DB-NE-03233 TS 4.2.4 Within limits provided in COLR MODE I Above 15% RTP see notes BAW-10179P, Rev.6"Regulating rod position (rod index), APSR position, axial power imbalance, and quadrant power tilt are process variables that together characterize and control the three-dimensional power distribution in the reactor core... During the reload safety evaluation, a three-dimensional power distribution analysis is performed to set operating limits that will preserve the accident initial condition assumptions for the ECCS, loss of flow, ejected rod worth, and SDM analyses...
The power distribution analysis is based on calculating the three-dimensional peaking distribution over the range of potential power operation.
Parameterization on power level, cycle bumup, control rod and APSR positioning, and xenon distribution is included in the analysis.
The simulation of xenon transients...is further augmented by APSR motion in order to generate conservative (higher) peaking factors...
Correlations between peaking margins and axial offset, regulating rod insertion, and APSR position are generated from this database, and the cycle-specific LCO limits are determined...
The limits on regulating rod insertion, axial power imbalance, and quadrant power tilt given in the COLR typically represent the measurement system-dependent limits beyond which the core power distribution and reactivity criteria could be violated during an accident.
Measurement system-independent limits are obtained directly from the reload safety evaluation analysis, without adjustment for measurement system error and uncertainty.
Adjustments are applied to account for uncertainties in the indicated regulating rod position, APSR position, axial power imbalance, and quadrant power tilt due to their measurement systems...
Regulating rod insertion and APSR insertion measurement system-dependent limits are derived by adjusting the measurement system-independent limits to allow for thermal power level uncertainty and rod position errors...The axial imbalance measurement system-dependent limits are determined from a statistical combination of incore system measurement and observability errors. The imbalance error adjustment is formulated such that when the measured imbalance is within the error adjusted limits, the true axial imbalance is within the unadjusted limit at a 95/95 probability/confidence level for the limiting core conditions.
The quadrant tilt measurement system-dependent limits are determined from a statistical combination of incore system measurement and observability errors. The tilt error adjustment is determined such that when the measured tilt is equal to or less than the error adjusted limit, the peaking increase due to tilt is equal to or less than the peaking increase used in the power distribution analysis at a 95/95 probability/confidence level." Summary: The effect of rod position, APSR position, axial power imbalance, and quadrant power tilt on core peaking factors is analyzed on a cycle specific basis. The resultant LCO limits, as presented in the COLR and TS, are adjusted for measurement errors. Even though the quantitative value ofthese errors is not explicit here, the reload analysis methodology described here provides reasonable assurance that instrument Incore Detectors
+ computer program Within limits provided in COLR Thursday, May 08, 2003 Page 18 of 48 Parameter Source Value Applicable Mode RCS Temperature TS 4.2.5.1, Table 3.2-2</= 610 Deg. F MODE I Est Inst Error Error Source Margin Margin Source+/-3 deg. F NR RTD's Accuracy+rcadabilit y+ drift Approx. 25 deg.Operational Notes BAW-10179P explains that the LCO on RCS hot leg temperature is established such that during normal plant operation the steady-state RCS hot leg temperature will not be greater than that corresponding to the initial conditions assumed for the DNBR analysis.The RCS outlet temperature loops have RTDs that are used to provide a narrow range temperature signal (520 to 620F). The RCS narrow range temperature indications are Type 180 single Edgewise with a standard accuracy of 1.5% of full scale span (M-324-145-2).
The temperature elements are 100 ohm platinum type RTDs with a specified accuracy of 0.5% (Instr. Index). Other instrument effects are readability of the indicator (1/2 minor scale division), drift, and any unidentified errors. Engineering judgement is that temperature readings are within 3 deg. F.Normal operation results in a Tavg of approx. 582 deg. Under expected transient conditions such as load changes, the limits are not strictly applicable.
This is reflected in the bases and in the accident analyses.
The only time the limit would be in effect would be under steady-state conditions, wherein the program Tavg of 582 F provides ample margin to the limit. Consideration of the additional 3 deg possible uncertainty is irrelevant given this amount of operational margin.The RCS narrow range hot leg pressure indication is utilized to verify RCS operation within'this Technical'Specification limit. The LCO on RCS pressure is established so that during normal plant operation the steady-state core exit pressure will be maintained at a level greater than or equal to the value assumed for DNB analysis.Surveillance Test Instrument(s)
Surveillance Limit DB-OP-03006 TI RC3B4, TI RC3A2 MAX 610 Deg. F RCS Pressure TS 4.2.5.1, Table 3.2-2>/= 2062.7 psig(>/= 2058.7 psig for 3-loop operation)
+/- 30 psi NR pressure loops and indicators DB-OP-03006 PRS RC2A2, PC2B or as an alternative:
computer points P721, P722, P729, P730>/= 2062.7 psig (>/= 2058.7 psig for 3-loop, Difference Tolerance of 30 PSID per Attachment I of procedure The surveillance limit is verified by a channel check comparison of the RCS narrow range pressure indicators Approx. 60 psig. (PRS RC2B, PRS RC2A2) or alternately with the computer points. The lowest reading is subtracted from the highest reading and that calculated maximum difference is compared to the difference tolerance of 30 psid.Normal operation results in a pressure of approx. 2155 psig. Under expected transient conditions such as Operational and load changes, the limits are not strictly applicable.
This is reflected in the bases and in the accident analyses.surveillance The only time the limit would be in effect would be under steady-state conditions or during slow ramped load changes, wherein the program pressure provides ample margin to the limit. Consideration of the additional 30 psig possible uncertainty is irrelevant given this amount of operational margin.MODE I Thursday, May 08, 2003 Page 19 of 48 Parameter Source Value Applicable Mode RCS Flow Rate Est Inst Error Error Source Margin Margin Source+1-5%Notes Surveillance Test Instrument(s)
Surveillance Limit DB-OP-03006, DB-SP-03358 LIR CRs 02-06885(Mode 5), 02-06215(Mode 4), 02-07395(mode 1), 02-06593 (Mode 4), 03-01450 TS 4.2.5.1/4.2.5.2, Table 3.2-2>/= 389,500 gpm (>/= 290,957 gpm for 3-loop operation)
MODE I Draft sensitivity Margin: Technical Specification Table 3.2-2 states that the DNB Margin limit on minimum RCS flow rate analysis includes an uncertainty of 2.5%.various 2.5%Reload T-H analyses BAW-2417, Unit I Cycle 14 Reload The 2.5% RCS flow uncertainty is an assumption used in core design and thermal-hydraulic analyses.
This uncertainty assumption is documented in B&W Calculation BAW-10187P-A, "Statistical Core Design for B&W Designed 177FA Plants".Estimated error: A 1978 B&W calculation
("Determination of Total RC Flowrate and Its Accuracy for Davis-Besse V", by Robert W. Winks, dated 5/25/78) finds the error associated with the determination of total RCS flow, via a secondary side heat balance, to be +/-2.09%.
This calculation also determines the error associated with calibrating the RCS flow meters against the secondary side heat balance to be +/- 2.2%. This calculation has been used in the past to demonstrate that the RCS flow determination heat balance methodology error, of DB-SP-03358, is within the design basis 2.5% uncertainty assumption.
There are several problems with this calculation:
-Not enough information is presented in the calculation to determine if string errors are bounding, e.g., there is no information regarding transmitter head correction, temperature effects, instrument drift or process measurement errors assumed.--The calculation uses data specific to Cycle I that may no longer be bounding.
Flow coefficients, instrument spans and instrument calibration accuracies may have changed over time.-The calculation string errors do not include the process computer points. The current test methodology uses only the computer points.-The calculation assumes a constant RCP heat addition.
The current test methodology calculates RCP heat based on pump power indication and an assumed efficiency.
-The calculation does not consider heat additions and losses due to makeup and let down. This is inconsistent with the test methodology.
-The calculation uses an assumed value for radiative heat losses that is not consistent with the test procedure assumptions.
Based on the above, the B&W calculation does not support an assumption that error inherent in the RCS flow measurement via secondary side heat balance is less than the 2.5% assumed in the design basis .analyses.
The DB-SP-03358 test~methodology and acceptance criteria need to be consistent with the test error determination.
For example, it may be necessary to take more accurate data (e.g., take transmitter voltages or RTD resistances with M&TE) or take a greater number of data points, in order to reduce the test error. As an alternative, it may be necessary to revise the test acceptance criteria, in order to ensure that measured RCS flow is greater than the design limit.Calculation C-ME-064.02-027 was prepared in 1990 in an effort to improve the accuracy of the secondary side heat balance RCS flow determination (DB-SP-03358).
This calculation cannot be used to support the 3%Thursday, May 08, 2003 Page 20 of 48 Parameter Source Value Applicable Mode Est Inst Error Error Source Margin Margin Source Notes Surveillance Test Instrument(s)
Surveillance Limit current test methodology due to the following:
Report-The calculation uses cycle specific data that may no longer be bounding.-The calculation requires that a minimum of 180 data points be used, in order to reduce error. The current test methodology takes only about 18 data points.-The calculation is inconclusive in that it calculates an error of+/- 1.33% using a square root sum of the squares, but also calculates an error of +/- 4.74% by summing errors. Application of the 1.33% error would require demonstration.that all measured parameters were random, normal and independent.
-The calculation assumes zero error associated with flow primary elements.A sensitivity model was drafted, modeling the current DB-SP-03358 methodology, and using calculated loop uncertainties, or assumptions (when a loop uncertainty was not available).
This resulted in an approximate error of+/-5% of design flow.DHR Relief Valve Setpoint TS 3.4.2</= 330 psig Modes 4 & 5 Pressurizer Safety Valves Setpoint TS 3.4.3</= 2525 psig 1,2 & 3<=+/-I% of specified SD-042 test pressure TS 3.4.2 has surveillance requirements (4.4.2) to determine operability ofDHR relief valve DH-4849 per the surveillance requirements of Specification 4.0.5. DB-PF-3002.00 is used to satisfy TS 4.0.5 requirements for Procedural the Inservice Test Program (IST) safety and relief valve. TS note that the lift setting pressure shall correspond requirment of to ambient conditions of the valve at nominal operating temperature and pressure.
The procedure requires DB-PF-3002.00 that the overall combined accuracy of the test equipment is less than or equal to +/- 1% at the specified test pressure.
The procedure also specifies accuracy and pressure range requirements for the M&TE.IST of valves to be dispositioned by IST Group.1% (+/- 25 psi) From License Amendment 250: "The pressurizer code safety valves must be set such that the peak Reactor Coolant System pressure does not DB-MM-03000 exceed 110% of design system pressure (2500 psig) or, 2750 psig. The control rod group withdrawal accident will result in the most limiting high pressure in the RCS. The analysis assumes RPS high pressure 1% (+/- 25 psi) trip at 2355 psig and the code safety valves open at 2500 psig. The tolerance on the RPS instrument accuracy is 30 psi and, it is + 1% for the code safety valve settings.
The pressurizer pilot operated relief valve was TS Bases assumed not to open for this transient.
The resulting systim peak pressure was calculated to be 2700 psig.Therefore, the code safety valve setpoint is conservatively set at -< 2525 psig which is the maximum pressure of 2500 psig + I% for tolerance." DB-PF-3002.00 M&TE</= 330 psig DB-MM-03000 M&TE 2500 psig +/- 25 psi There is adequate margin to provide reasonable assurance that additional consideration or application of uncertainty is not required prior to restart.Thursday, May 08, 2003 Page 21 of 48 Parameter Source Value Applicable Mode Pressurizer Pilot Operated Valve Setpoint TS 3.4.3 Est Inst Error Error Source Margin Margin Source 13.4 psi C-ICE-080.0!
-001 15 psi C-ICE-080.01
-001 Notes>/= 2435 psig 1,2&3 From TS Bases 3.4.3: "The pressurizer pilot operated relief valve should be set such that it will open before the code safety valves are opened. However, it should not open on any anticipated transients.
BAW-l 890, September 1985 identified that the turbine trip from full power would cause the largest overpressure transient.
This report demonstrated that with a RPS high pressure trip setpoint of 2355 psig the resulting overshoot in RCS pressure would be limited to 50 psi. Consequently, the minimum PORV setpoint needs to accommodate both the RCS pressure overshoot and the RPS instrument string error of 30 psi." The PORV is set at a field setpoint of 2450 psig increasing, or 15 psi higher than the TS required setpoint and allowable value.Calculation C-ICE-080.01-00, Rev. 0, calculates the Channel uncertainty as 13.4 psi, exclusive of the transmitter that is common to the PORV and RPS.There is adequate margin to provide reasonable assurance that additional consideration or application of uncertainty is not required prior to restart.Surveillance Test Instrument(s)
Surveillance Limit DB-MI-03051/03052 DB-MI-03075/03076 M&TE>/= 2435 psig Thursday, May 08, 2003 Page 22 of 48 Parameter Source Value Applicable Mode Pressurizer Level TS 4.4.4 45<Level<305 inches l&2 Est Inst Error Error Source Margin Margin Source Notes Surveillance Test Instrument(s)
Surveillance Limit-85" 220" assumed in safety analysis (32-1171148-00)
-305" allowable by TS& DB-OP-03006 LIR CR 02-08093 -TS 3/4.4.4 requires that the pressurizer be OPERABLE with.. .a water level between 45 and 305 inches. B&W calculation 32-1171148 uses'an initial pressurizer level of 220 inches (with no uncertainty) for the Loss of Normal Feedwater (LOFW) transient analysis.
Thus, there is -75 inches of margin for the high pressurizer level limit. Note that the pressurizer level is normally controlled at 220 inches during normal operation.
The TS Bases state that "the high level limit is based on providing enough steam volume to prevent a pressurizer high level as a result of any transient".
Contrary to this, if pressurizer level were at 305 inches during a LOFW event, then the pressurizer would assuredly go solid. As noted in calculation 32-1171148, the LOFW transient is the bounding event for high pressurizer level. From the USAR: The acceptance criteria, however, for the loss-of-feedwater transient are that l)Fuel damage shall not occur, and 2)Reactor Coolant System pressure shall not exceed code pressure limits. There is no acceptance criterion that requires the pressurizer maintain a steam bubble. This is supported by NUREG-1430, Standard Technical Specifications for B&W Plants, Bases which state that "prevention of water relief[through the pressurizer safety valves] is a goal for abnormal transient operation, rather than a SL (Safety Limit), the value for pressurizer level is nominal and is not adjusted for instrument error". From this it can be concluded that there is no regulatory requirement to maintain pressurizer level conservatively with respect to the LOFW analysis assumption of 220 inches, or to apply instrument uncertainty.
That said, the potential of the pressurizer going solid during a LOFW transient, and the subsequent probability of damage to the PORVs or code safeties, and the pressure transient to which the reactor vessel will be subjected, are not something that should be cavalierly accepted.
It would be prudent to limit the pressurizer level to a value that will ensure that a bubble remains in the pressurizer during the LOFW transient.
License Amendment Request LAR-01 -0012 has been submitted to revise the TS 3/4.4.4 limit to a maximum of 228 inches.In October 1999, the NRC approved Framatome Topical Report BAW-10193P-A, which describes the use of RELAP5IMOD2-B&W for analyzing the non-LOCA transients.
In this topical, it states that "...for those accidents that cause an increase in pressurizer liquid level and/or reactor coolant system pressure, the initial pressurizer level will be set to a value greater than or equal to the nominal value plus measurement uncertainty.
The reason for adding measurement uncertainty for these types of calculations is that the pressurizer level has an effect on pressure that is not addressed by adding uncertainties to reactor protection system trip setpoints or pressurizer safety valve setpoints." Thus, any future LOFW transient analysis will contain a consideration of pressurizer level instrument uncertainty.
The maximum operating steam generator level is based primarily on preserving the initial condition assumptions for steam generator inventory used in the FSAR steam line break (SLB) analysis.
An inventory of 62,500 lb was used in this analysis and must not be exceeded due to the concerns of a possible return to criticality because of primary side cooling following an SLB and the maximum pressure in the reactor building.Figure 3.4-5:, in the accompanying LCO, is based upon maintaining inventory
< 57,000 lb, which is approximately 10% less than the inventory used in the FSAR accident analysis, and therefore is conservative.
Also, the normal operational performance of the OTSG level provides significant margin to the high level limit. A typical 100% value is approx. 60% Operating Range and ramps down to the no-load value of approx.20%. This represents a curve that is well below that shown in Fig. 3.4-5.DB-OP-03006 LRS-RC-14, L768 Steam Generator Level TS 4.4.5.6 N/A N/A DB-OP-03006 LRSSP9A(B), LISP9A 1 (2), LISP9B1(2), U881(82,83,84), L891(92,93,94)
TS FIG. 3.4-5 or Calculate the max limit: Max SG Operate Range Level=(l.23 x Main Steam Superheat)
+ 43, but not greater than 96% level See TS Significant 1,2,3&4 1,2,3&4 Analysis Thursday, May 08, 2003 Page 23 of 48 Parameter Source Value Applicable Mode RCS leakage TS 4.4.6.2.1 Est Inst Error Error Source Margin Margin Source N/A Notes Surveillance Test Instrument(s)
Surveillance Limit N/A< I GPM Unidentified Leakage/<1 0 GPM Identified Leakage 1, 2,3 & 4 Surveillance method is comparative SD-039A, DB-OP-01200, DB-SP-03357, NUREG 0800 (Sect. 5.4.5), NRC Bulletin 88-0008, Reg. Guide 1.45 10CFR50 Appendix A Criterion 14 and 30 provide the requirements for RCS leakage. Regulation Guide 1.45 "Reactor Coolant Pressure Boundary Leakage Detection System" describes methods acceptable to the NRC of implementing General Design Criteria 30 with regard to the selection of leakage detection systems for the reactor coolant pressure boundary.
This includes the identification of unidentified sources of leakage with the flow rate monitored with an accuracy of one gallon per minute or better. The leakage detections systems should have a sensitivity and response time of one gpm in less than one hour for measuring unidentifed leakage.DB-OP-01200 provides the philosophy and guidance to be utilized when evaluating Reactor Coolant system (RCS) Leakage. Procedure DB-SP-03357 (RCS Water Inventory Balance) verifies the RCS leak rate is within the required limits by calculating the rate of change of RCS liquid volume.The procedure explains that the accuracy of the test is dependent on steady state plant conditions and deviations will affect the accuracy of the results. The RCS water inventory balance is performed by recording measurements for the RCP Seal leakage indicators, Quench Tank Level, sump level, MU Tank level and other parameters and then comparing results taken over a given period of time. Though quantification of the individual instrument errors would be difficult, the comparative nature of the measurement tends to cancel out any accuracy/drift components.
The remaining components (those associated with resolution) must be evaluated for compliance with the I gpm in 1 hour sensitivity limits of the Reg. Guide. The ability to detect a 60 gallon change in a one hour period is a requirement for the overall system, and it is well within the capability of the above mentioned tank level indicators.
The containment sump monitor has a resolution of 30 gal/inch over a 48" span and is considered acceptable.
D-B does not employ separate condensate collectors, but instead routes CAC drainage to the containment sump. The radiation and humidity detectors do not have a specific sensitivity, though industry practice has shown that they are able to detect even small (-.I gpm) steam leaks.In summary, the leakage detection systems used at D-B and the methods employed to calculate leakrates are conservative and consistent with industry standards.
Based on evaluation of the ranges of the instruments and the methods employed to perform the calculations, it is the judgement of this reviewer that compliance with the Tech Spec limits may be demonstrated without consideration of additional instrument uncertainties for the leak detection system components.
This conclusion is applicable for all RCS leakage specifications (except primary-to-secondary).
This value is calculated from radiation monitor readings and/or isotopic analysis using industry standard techniques.
The 150 gpd limit is a nominal value chosen to limit secondary contamination below the assumed activity cohcentration for SGTR analysis and to signal the onset of tube degradation.
Current SGTR analyses assume preexisting coolant activities that are well above the levels that would result from a 150 gpd leakrate. (USAR Table 15.A-4)It would not be practical to attempt to quantify the uncertainty associated with this measurement.
DB-SP-03357 Inventory Balance by RCS Leakage Rate Program, Automatic data retrieval calc or Attach. 8, manual data retrieval, numerous instruments
< I GPM Unidentified Leakage/< 10 GPM RCS Primary to Secondary Leakage TS 4.4.6.21 .e< 150 GPD 1,2,3 &4 Significant Accident analysis Detailed investigation is provided in CR 02-05948.Thursday, May 08, 2003 Page 24 of 48 Parameter Source Value Applicable Mode RCS Specific Activity Est Inst Error Error Source Margin Margin Source No estimate Notes Surveillance Test Instrument(s)
Surveillance Limit TS 4.4.8 See TS see notes see notes 1, 2, 3 & 4 (see TS)USAR LIR CR 02-05948 (Mode 2),.02-05951 (mode 2)As noted in the TS Bases for TS 3/4A.8:"The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour doses at the site boundary will not exceed an appropriately small fraction of the Part 100 limit following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM." The doses reported in the USAR for a steam generator tube rupture accident are based on original Bechtel Calculations 60.16 and 61.01. These original calculations used the source terms given USAR Tables 15A-4 and 15A-5 for RCS and secondary side Iodine activities respectively.
From USAR Section 15.A.7.0: "Among the accidents analyzed in Chapter 15, only the control rod ejection accident (15.4.3), the Loss of Coolant Accident (15.4.6), and the fuel handling accident (15.4.7) utilize the activity in the fuel and/or in the fuel rod gap in the calculation of off site doses. For the other accidents, the dose consequences were evaluated based on the fission product activity in the reactor coolant. Since the coolant activity used in the accident analysis is significantly higher than the Technical Specifications limits, the USAR dose analysis is bounding for accidents that do not result in a release from the fuel rods." Although not specifically quantified here, there is conservative margin included in the design basis calculations and source term that give reasonable assurance that a further consideration of uncertainty is not required prior to restart..
-DB-CH-01815 DB-CH-03000 Chemistry Counting Equipment-gamma spectroscopy Thursday, May 08, 2003 Page 25 of 48 Parameter Source Value Applicable Mode CFT Volume Est Inst Error Error Source Margin Margin Source>0.21 ft > 107 gal Notes Surveillance Test Instrument(s)
Surveillance Limit TS 4.5.1.a.I see notes EMPAC 7555<Volume<8004 1,2 & 3 0.28 ft. upper limit 0.18 ft lower limit B&W 32-1175316-00, C-NSA-051.01-003 CFT Volume has both upper and lower limits that are critical to safety analyses.
The upper limit bounds the volume of gas available as a motive force to move the CFT volume into the RCS; the rate at which the CFT volume is injected is modeled dependent on this gas volume. The lower limit bounds the fluid volume available to be injected into the RCS. CFT volume must be maintained within the upper and lower limits such that safety analysis assumptions are maintained.
Current safety analyses use 1000 ft^3 and 1080 ft^3 as the lower and upper limits respectively; there is approximately 1% (10 ft^3) built into these analysis limits w.r.t. the TS limits.'Margin: B&W document 32-1175316-00 used as input to the accident study a volume of 7,480 gallons (120F), 75 gallons less than the minimum TS requirement for CFT inventory.
This 75 gallons is to account for instrumentation uncertainties.
The 75 gallons is derived in calculation C-NSA-051.01-003, which assumed an indication uncetainty of 0.14 ft^3 at 508.1 gal./f, taken from the instrument string data sheet.Calculation C-ME-51.01-086 added attachment B which determined that an indicated CEFT level of 13.44 ft correspondes to 8004 gal. And 12.56 ft correspondes to 7555 gal. Procedure DB-OP-03006 verifies CFT level between 12.6 and 13.3 feet. This results in an upper level margin of 0.14 + 0.14(13.44-13.3)
= 0.28 ft and a lower level margin of-0.14- 0.04(12.56-12.6)
= 0.18 ft.Estimated Instrument error: The 0.14 ft assumed as margin is calibration tolerance of the indicator ONLY. It does not take into account the calibration tolerance of the transmitter, process measurement effects (boric acid solution density), M&TE, temperature effects, drift, etc. Taking ONLY the calibration tolerances of the transmitter and indicator (transmitter
= 0.25% of 0-780 iwc span- 0.16 ft): SQRT(.1 6^2 + .14A2) = 0.21 ft.Using the conversion from C-NSA-051.0l-002:
508.1 gallft * .21 ft.= 107 gal.Even considering ONLY the calibration tolerance, the margin is not sufficient to account for instrument uncertainty.
No instrument uncertainty calculation exists for the level instrumentation.
Estimated error: 26 ppm is considered to be the maximum credible titration error, based on 38-1290179-00, which documents a memo from Framatome to DB (10/02).Margin:Lower limit -the post LOCA boron calculations (51-5016651-01) assume 2500 ppm in the CFT, giving a 100 ppm margin for the lower limit.The upper limit preserves assumptions used in the sump pH calculation(86-5024418) and the boron precipitation analyses (86-5006059-00).
The post-LOCA sump pH determination contains sufficient conservatisms, and a broad enough acceptance criterion, to accommodate any error associated with -measurement of boron concentration.
The boron precipitation analysis (86-5006059-00) uses 4000 ppm CFT boron concentration.
DB-OP-03006 LI CF3AI (A2), LI CF3B 1 (B2)Alternative measurements made with computer points L079(80,89,90) 7555<Volume<8004 CFT Boron Concentration TS 4.5.L.b 2600</= Concentration</=
3500 ppm .1,2&3 26 ppm 38-1290170-00 100 ppm (lower limit)51-5016651-01 DB-CH-03028, DB-CH-03028 Grab Sampling, Mettler Model DL55 Titrator or inT Surveillance value is to within the maximum and minimum TS values, 2600<<=Concentration<=
3500 Thursday, May 08, 2003 Page 26 of 48 Parameter Source Value Applicable Mode CFT Cover Pressure Est Inst Error Error Source Margin Margin Source 7 psi Notes Surveillance Test Instrument(s)
Surveillance Limit B&W 32-1175316-00, SD-040 DB-CH-03003 TS 4.5.1.a.1 575<Press.<625 1, 2.& 3 B&W The Core Flooding Tank cover gas is controlled to maintain 370 ft3 of non-condensible gas between 575 32-1175316-00 psig and 625 psig (Ref. SD-040, B&W 32-1175316-00).
The proper Core Flooding Tank pressure is 600 +/-25 psig to allow reasonable pressure variations.
The core flooding system description states that the LOCA safety analysis used the values of 370 ft3 and 568 psig (pg 1-3 of SD-040). This allows for 7 psig relative to the TS minimum value. The accident analysis, B&W 32-1175316, explains that 7 psi was included for See notes. instrumentation error. The pressure indicators arc Bailey Controls' model RY21 IX and have an accuracy of+/- 1% over a range of 700 psig (+/- 7 psig). The alternate use of the computer points for verification of the TS limits should allow for more accurate readings.
No calculation was located that determines the instrument uncertainty for these indicators.
The surveillance value accounts for the accuracy of the indicator.
There are other errors that contribute to the total instrument uncertainty and those should be considered with the preparation of a calculation.
The Core Flooding Tank also has two alarms (HILO) on each tank that provide operations with indication of high and low pressure and are set within the technical Specification minimum and maximum surveillance values (615/585 psig per M-720I). These alarms provide added assurance that adequate pressure is being maintained in the CFT. It is concluded, based on the instrument error and conservatisms in the accident analysis, that instruments for this parameter provide reasonable assurance that they are acceptable for verification of CST pressure limits.This issue will need to be revisited post-restart.
PI CF4AI(A2), PI CF4Bl (B2)Alternative measurements made with computer points P079(80,89,90)
Surveillance value is to within the maximum and minimum TS values, 575<Press.<625 TSP Storage Basket Volume TS 4.5.2.d.4> 290 Cubic Feet 1,2 & 3 N/A N/A 40 113 of TSP TS Bases 3/4.5.2 HPI Flow TS 4.5.2.h>/= 375 gpm at 400 psig 1,2&3 N/A N/A N/A N/A Tech Spec Bases 4.5.2 provides an explanation for the surveillance value of> 290 Cubic Feet of TSP in the TSP storage baskets. The amount of TSP required is based on the mass of TSP needed to achieve the required pH. The minimum required volume of TSP to meet all analytical requirements is 25063. The surveillance requirement of 290ft3 includes 40ft3 of spare TSP as margin. A carpenter rule or equivalent is used to make the measurements.
Typically, the accuracy associated with reading a scale is +/- one half of a division.
The error is not significant for this surveillance.
Surveillance 4.5.2.h is to be performed every refueling interval by a flow balance test, during shutdown, following completion of modifications to the HPI or LPI subsystems that alter the subsystem flow.Baseline testing was done by procedures DB-PF-04207(04208) which established pump curves for the HPI pumps as the basis for the Inservice Testing Program Reference Values. The testing was performed to verify pump flow between 425 and 450 GPM as indicated on the flow indications.
Current IST requirements specify an accuracy of +/- 2% of full scale for analog instruments, +/-2% of total loop accuracy for a combination of instruments, or +/- 2% of reading over the calibrated range for digital.instruments.
Though these allowances would tend to confirm that consideration of instrument uncertainty was included, recent re-allocations of available margin with respect to pump degradation have compromised this assumption.
Therefore, it is concluded that without further calculational review, sufficient operability margin to account DB-CH-03003 Carpenter rule or equivalent
> 290 Cubic Feet DB-PF-04207(04208)
Baseline Tests see notes FYI HP3C, FYIHP3D HPI P I FYI HP3A, FYIHP3B HPI P 2 413 gpm Thursday, May 08, 2003 Page 27 of 48 Parameter Source Value Applicable Mode LPI Flow Est Inst Error Error Source Margin Margin Source N/A Notes Surveillance Test Instrument(s)
Surveillance Limit DB-PF-03236(03237)
NUREG-1482, NG-EN-00314, LTR BWT-1619 DB-SC-03109, Pump Performance Curves TS 4.5.2.h>/= 2650 gpm at 100 psig 1,2&3 N/A N/A N/A BWST Volume 22 inches Surveillance 4.5.2.h is to be performed every refueling interval by a flow balance test, during shutdown, following completion of modifications to the HPI or LPI subsystems that alter the subsystem flow.Baseline testing was done by procedures DB-PF-04207(04208) which established pump curves for the LPI pumps as the basis for the Inservice Testing Program Reference Values. No tolerances were associated with the tested flow rates. The measurements were performed with instruments that were verified to have +/-2% or better accuracy (+/-60 gpm) at flow reference point. (ASME OM Code -1995 Ed. With 96 Addenda) Though these allowances would tend to confirm that consideration of instrument uncertainty was included, recent re-allocations of available margin with respect to pump degradation have compromised this assumption.
Therefore, it is concluded that without flrther calculational review, sufficient operability margin to account for instrument uncertainty cannot be verified.The upper limit on BWST level is related to minimum post-LOCA pH, flooding of equipment in the containment, and the allowance of margin to overfill the tank resulting in a possible environmental issue.Sufficient conservatisms exist in the inventory of TSP to offset a minor increase in BWST inventory delivered to containment, and in practice the suction of the ECCS pumps is always shifted to recirculation well before the tank is emptied. Neither of these aspects are critical to nuclear safety and do not warrant the application of additional uncertainties to the surveillance limits.The lower limit will need to be adjusted for instrument uncertanties and Instrument string and indicator uncertainty will need to be included to ensure that the Tech Spec limit of 500,100 gal. supports the minimum injection volume of 360,000 gal. injecting prior to the Level 5 permissive being reached.In summary, it is concluded that the existing specifications for maximum contained volume in the BWST is conservative and supports operability of the system. However, the lower limit could not be verified as conservative at this time.TS 4.5.4.a.1 500,100<Volume<550,000 ICE Cale 48.01-004 Min level -Approx I ft.(after change)Calc 48.01-04 and.TS surveillance DB-OP-03004 LT-1525A-D 524.000 1,2,3 &4 Thursday, May 08, 2003 Page 28 of 48 Parameter Source Value Applicable Mode BWST Concentration Est Inst Error Error Source Margin Margin Source 1% of reading 38-1290170-00 lower limit ->200 ppm Upper limit -0 ppm 51-5016651-00 86-5006059-00 Notes Surveillance Test Instrument(s)
Surveillance Limit TS 4.5.4.a.2 2600 ppm</=concentration</=2800 ppm 1,2,3 & 4 BWST solution temperature TS 4.5.4.b 15 deg. F EMPAC see notes>/= 35 deg. F.1,2,3 & 4 Approx. 6.5 deg. F Estimated error: Boron titration error is estimated to be a maximum of 26 ppm for the lower limit (1% of reading) in the memo documented in 38-1290170-00.
Margin: For the Lower limit of BWST boron concentration (2600 ppm) -Calculation 51-5016651-00 (supplemented by rev. 1, 51-5016651-01), "DB CYI4 Boron calculations", determines that the required BWST concentration to maintain a 1% dK/K during post-LOCA conditions is 2312 ppm. This gives a margin for the lower limit of 2600-2312
= 288 ppm. This provides reasonable assurance that the lower limit for this parameter does not require further consideration of measurement uncertainty prior to restart.For the BWST boron concentration upper limit (2800 ppm): 2800 ppm is used as a key input parameterin the boron precipitation analysis (86-5006059-00), with no measurement uncertainty assumed. Thus, there is no margin available for the upper limit. There are conservatisms built into the boron precipitation analysis, e.g., boron solubility limit is conservatively increased by 4 w/o: These conservatisms, however, may not be sufficient to give reasonable assurance that the current limits do not require attention prior to restart. It would be prudent to adjust the surveillance limits to account for measurement error. CR 03-02644 was initated to document this.This value is used in the Inadvertant Containment Spray analysis -which assumes 35 deg. injection temperature.
The existing surveillance requirement does not include loop uncertainty.
The BWST operating procedure, DB-OP-06015, limits BWST operation to the range 50 -70 Deg. F. This current operating procedure for the BWST requires steam heating to be placed in service below 50F as well as heat tracing on external piping. These actions, combined with the fact that even under severely cold external temperatures the tank requires weeks to cool down to the 35F limit, provide adequate protection for the operability limits on the BWST.The temperature indicator is a GE-180 with a range of 0-250 deg. F, reference accuracy of 1.5% and an as-left tolerance of 3%. This results in a measurement error of approximately 8.5 deg. F.Solubility of the boric acid solution is another related criteria.
In a 1995 telecon between D.R. Woukko and the NRC, related to LAR 95-0010, we (FENOC) stated that "as long as the boron concentration at DBNPS remains less than 4000 ppm, then there will be no precipitation problem at 35 Deg. F".Crystallization is also not a concern at this temperature since the solubility limit of a boric acid solution at 35 Deg. F is approximately 4900 ppm: From Langes Handbook of Chemistry, 13th Edition (McGraw-Hill 1985), the solubility of Boric Acid(H3BO3) is: 2.67g/lOOg solution@
0 Deg. C (32 Deg. F) = 4668 ppm 3.73g/100g solution @ 10 Deg. C (50 deg. F) = 6522 ppm The above shows that (using conservative assumptions) temperatures indicated at 35 Deg. F will not render the boric acid solution, at the maximum allowed concentration, insoluble.
Thus, there is adequate analytical margin available to show reasonable assurance that this parameter is acceptable for restart.Given the large potential indication error, it would be prudent to revisit this parameter post restart and adjust surveillance limits accordingly.
DB-OP-03007 T11534>/= 35 deg. F. </= 90 Deg. F DB-CH-03004 M&TE (Chemistry Autotitrator) 2600 ppm</=concentration</=2800 see notes Thursday, May 08, 2003 Page 29 of 48 Parameter Source Value Applicable Mode Containment Leak Rate Purge and Exhaust Valve Special Test Pressure Est Inst Error Error Source Margin Margin Source+/-2%Notes TS 4.6.1.2.2>/= 20 psig 1,2,3 &4 M&TE Slight Surveillance Containment Pressure+/- 3 iwg DP loop TS 3/4.6.1.4-14 iwg < Press. < +25 iwg 1,2,3 & 4 Significant Safety Analysis This special test is performed in accordance with App. J standards and includes allowances for M&TE accuracy per ANSL/ANS 56.8-1994.
This, combined with conservative acceptance criteria, is sufficient to demonstrate operability and successful leakrate performance without additional consideration of uncertainties.
See Leakrate testing program description for additional detail.Bases explains that the limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psi and 2) the containment peak pressure does not exceed the design pressure of 40 psig during LOCA conditions.
Max peak pressure from LOCA is 37 psig, which is less than the maximum internal pressure of 40 psig.Review of the containment analyses indicated that an initial pressure condition of 17.04 psia is assumed in the development of the SFAS setpoints.
This value includes I psia (28 iwg) of generic margin and assumes conservative environmental conditions (maximum barometric pressure).
These quantities are applied to the maximum operating pressure of 25 iwg (0.8 psig), along with instrument errors to determine the trip setpoints.
The vacuum condition is prevented through the use of the 10 Vacuum Relief devices which overlap the normal operating range of negative pressures.
The vacuum breakers are staggered opening check valves which begin opening at -4.2 iwg (0.15 psid) and are credited to prevent design negative pressure from being exceeded.
Therefore, the low pressure TS limit of-14 iwg is considered nominal.In summary, the upper and lower containment limits for differential pressure contain sufficient conservatism to either accommodate (high limit) or preclude consideration of(low limit) instrument uncertainty.
Ref. C-ICE-48.01-001 Surveillance Test Instrument(s)
Surveillance Limit DB-PF-3008 M&TE<0.15 La with a >20# test pressure DB-OP-03006 Section 4.3 PDI 645 (preferred) or local indicators PDI 645A, PDIS 624-14 iwg < Press. < +25 iwg Thursday, May 08, 2003 Page 30 of 48 Parameter Source Value Applicable Mode Containment Temperature TS 4.6.1.5 Est Inst Error Error Source Margin Margin Source+/- 2 deg.Thermocouple and indicators, averaged NMA N/A Notes</= 120 Deg. F 1,2,3 &4 The surveillance requirement is to determine the primary containment average air temperature as the arithmetical average of the inlet temperature(s) to the operating containment air cooler(s).
The limit of 120F is established so the overall containment average air temperature does not exceed the initial temperature conditions assumed in the accident analysis for a LOCA.TS 4.6.1.5 verifies primary containment average air temperature
(</= 120F) by reading indicators TI-1356 (57,58) or preferably computer points T298, T302, T306. The containment air is maintained between 75 and 120F by the Containment Air Cooling System during normal operation.
The location of the temperature elements is such that it is unlikely that a representative measurement can be made of the true average containment temperature.
A search of historical correspondence and Condition Reports concluded that this is an unresolved issue from a design standpoint.
At this time, Licensing has concluded that the current configuration meets the Tech Spee requirements and therefore constitutes a true measurement.
From an engineering perspective, this is difficult to justify.Historically, temperatures within 1 to 2 degrees of the limit have been observed.
This is not accomodatable by the instrument uncertainty present in the loops. A formal calculation will require development to determine appropriate instrument margin and establish a basis for surveillance procedure changes.Nominal Value SD-023, Reg. Guide 1.97, M-345 Hydrogen concentrations must be maintained at less than or equal to 4% following a LOCA (AEC Safety Guide 7 (RG 1.7) and IOCFR50, Appendix A). The analyzers have the capability of detecting, indicating, and recording hydrogen concentrations from 0 to 10 percent. The surveillance demonstrates operability of the hydrogen analyzers by calibration to a known gas concentration and adjustments to within established tolerances.
The Hydrogen Analyzers are required to be operable as RG 1.97 indications to monitor the hydrogen concentrations.
The instrumentation is qualified as Class IE equipment, in accordance with seismic requirements, and has other design and qualification requirements including accuracy.
Given these requirements, it is concluded that these instruments are capable of monitoring the level of hydrogen concentrations in containment.
Surveillance Test Instrument(s)
Surveillance Limit DB-OP-03007 T11356,T11357,T11358or computer points T298, T302, T306 (preferred)
</= 120 Deg. F Hydrogen Analyzers TS 4.6.4.2 2.5 +/- 0.5% H2 1&2 N/A N/A N/A N/A DB-MI-03729 DB-MI-03730 AI-5027, AI-5028 2.5 +/- 0.5% H2 Thursday, May 08, 2003 Page 31 of 48 Parameter Source Value Applicable Mode Shield Building Annulus Pressure TS 4.6.5.2.2>/= 0.25 iwg within 4 sec. After the fans attain a flow rate of 8000 CFM +/- 10%1,2,3 & 4 Est Inst Error Error Source Margin Margin Source 1% span 0-5 iwg DP cell Controller 0.45 iwg Setpoint Notes Surveillance Test Instrument(s)
Surveillance Limit CST Volume TS 4.7.1.3.1>/= 250,000 Gal.1,2 & 3 approx. 12 inches LT loop Significant Safety analyses The EVS provides drawdown of the Containment annulus and penetration spaces by the use of two 100%capacity fans with flow control provided by recirculation dampers. The dampers modulate to maintain a setpoint of -0.75 iwg upon actuation by a Level I signal. The various surveillances demonstrate fan performance, adequate drawdown (amount vs. time), and controller calibration.
Currently the surveillance criteria in DB-SS-3254 which verifies the time needed to reach -0.25 iwg does not provide for instrument indicator uncertainty and, therefore, should be adjusted in the future. However, the control system used during the operational lineup has sufficient margin to ensure that the performance requirements are met. Review of the inservice test data and of the control system tolerances determined that margin was present between the TS minimum of-0.25 iwg and the operating point of -0.75 iwg even when considering all applicable instrument tolerances.
In addition, review of the USAR revealed that significant margin exists in the assumed inleakage, backpressure , and outleakage which provide further assurance of the performance margin.In summary, the operating point of-0.75 iwg will ensure conservative margin to the TS limit and the associated dose analysis.The two redundant tanks contain a total volume of 250,000 gallons each with a collective minimum usable volume of 250,000 gallons of water for removal of reactor decay heat for a period of 13 hours and then to cool down the Reactor Coolant System to 280F (total of 19 hours).Review of the basis calculation Bech 69.11 determined that significant margins exist in the allowable limits for water volume: Hotter water temperatures are assumed than exist in the tanks (217F vs. 85F)No credit is taken for the Cond. Pump recirculation to the tank.The calculated requirement is 192,000 gal for the above scenario -this provides an additional 44,000 gal margin. The calculation estimates this additional quantity to provide 6 additional hours of HSD capability.
In summary, this system contains significant conservatism with respect to contained volumes and does not require evaluation of additional instrument uncertainty at this time.The activity limits for secondary radiation provide the threshold values for operation with degraded SG tubes as well as input conditions for SGTR dose analysis.
The nominal value of 0.1 uCi is consistent with industry practice as is the sample analysis methodology.
The current SGTR analysis has been shown to be within the acceptance criteria both for offsite dose and Control Room (GDC 19) habitability limits. Based on this, and the fact that the input activity level has a relatively minor impact on the dose results, itis concluded that sufficient conservatisms exist to preclude consideration of additional uncertainties on the sampling process.DB-SS-3254 PDI 5000/5014 PDC 5000/5014>0.25 iwg DB-OP-3006
'LT 512, 516>47 feet combined Secondary Coolant Specific Activity TS 3/4.7.1.4<= 0.10 Micro-curies/gram 1,2,3 & 4 DB-CH-1801, 1814, 3009 Various Thursday, May 08, 2003 Page 32 of 48 Parameter Source Value Applicable Mode UHS Level Est Inst Error Error Source Margin Margin Source+/--2 ft.Engineering judgement; 2 ft. =approx 6 times the string check Notes TS 4.7.5.1>/= 562 Feet IGLD 1,2,3&4 Oft 12501-M-001 LIR CR02-05356, 02-05351 UHS Level: Estimate of instrument error: 2 ft.Estimate of Instrument error source: Conservative engineering judgement:
The computer point string check tolerance is +/- 0.35 feet; 2 feet, or about 6 times the string tolerance is judged to be sufficiently conservative to account for potential process and environmental effects related to the computer point. Note, also, that annunciator alarms alert operators of UHS low level-around 564 feet IGLD; operators are then directed to take actions to preserve forebay level.Margin: 0.0 feet Margin Source: Calculation 12501-M-001 (Referenced-in USAR Section 9.2.5.1) determines the thermal performance or the UHS, and is used as design input to LOCA and MSLB analyses.
This calculation uses an ultimate heat sink level of 562 ft. IGLD as an initial condition.
For the determination of UHS level, there is no margin in the design basis accident analysis, or in the surveillance test procedure, to account for instrument error. Instrument uncertainty must be calculated, for each of the methods used to determine UHS level, and the surveillance procedure revised to conservatively incorporate the calculated uncertainties.
PAST OPERABILITY:
It must be determined whether the TS limit on UHS level might have been violated in the past, considering instrument error. It is estimated, based on engineering judgement, that 2 feet is conservative with respect to the uncertainty that would be calculated formally, thus will be used as bounding here. DADS data was examined to determine the forebay level, as indicated by computer point LA35, for the past four years. Based on this examination, at no time during the past four years did forebay level drop below 564 feet IGLD. Thus, there is reasonable assurance that the UHS was not inadvertently inoperable, in the past four years, due to level instrument uncertainties.
Surveillance Test Instrument(s)
Surveillance Limit DB-OP-03007 L435, L1902, Indication on intake structure 562 Ft. IGLD Thursday, May 08, 2003 Page 33 of 48 Parameter Source Value Applicable Mode UHS Temperature Est Inst Error Error Source Margin Margin Source 5 Deg. F Engineering judgement
-see notes 0.75 Deg. F DB-OP-03007, 12501-M-001 Notes Surveillance Test Instrument(s))
Surveillance Limit DB-OP-03007 T413, T11500, or M&TE 89.25 Deg. F TS 4.7.5.1 LIR CRs 02-08331, 02-06039, 02-05748, 02-05727, 02-05356, 02-05351UHS TEMPERATURE Estimate of instrument error: 5 Deg. F Estimate of Instrument error source: Conservative engineering judgement (A back-of-the-envelope look shows that 2.5 to 3.0 deg. F might be more realistic)*
Margin: 0.75 deg. F Margin Source: DB-OP-03007 acceptance criteria in Attachment I requires forebay temperature to be </=89.25 Deg. F. Calculation 12501-M-001 (Referenced in USAR Section 9.2.5.1) determines the thermal performance or the UHS, and is used as design input to LOCA and MSLB analyses.
This calculation uses an ultimate heat sink temperature of 90 deg. F as an initial condition.
</= 90 Deg. F 1,2,3 &4*Estimate of UHS RTD error: 100 Ohm Platinum RTD Deg. F Deg F A2 Process Effects 0.5 0.25 Sensor Effects-IEC 751 2.34 5.4756 RTD Calibration tolerance 0.75 RTD M&TE 0.15 RTD Temp Effects 0 RTD Pressure effects 0 RTD lead wire effects 0 RTD vib. Effect 0.075 RTD drift 0.5 Bridge Calibration tolerance 0.75 Bridge M&TE 0.15 Bridge Temp. Effects 0.1 Bridge Vib. Effects 0 Bridge Humidity Effects 0 Bridge Drift 0.5 A/D error 0.15 Concatenation error 0 0.5625 0.0225.0 0 0 0.005625 0.25 0.5625 0.0225 0.01 0 0 0.25 0.0225 0 SRSS 2.726486 Use 5 Deg. F as conservative For the determination of UHS temperature, there is no margin in the design basis accident analysis, to account for instrument error. The margin incorporated into the surveillance test procedure was recently added as a result of CR 02-03938.
This margin of 0.75 degrees F was based on the Instrument calibration accuracy (for the RTD only), and does not represent a total bounding instrument error. The 0.75 degrees is not conservatively bounding because it does not consider all possible error contributors such as calibration of the computer multiplexer bridge, drift and environmental effects. Total uncertainty must be calculated for each of the methods used to determine UHS level, and the surveillance procedure revised to conservatively incorporate the calculated uncertainties.
PAST OPERABILITY:
It must be determined whether the TS limit on UHS temperature might have been violated in the past, considering instrument error. It is estimated, based on engineering judgement, that 5 degrees F is conservative with respect to the uncertainty that would be calculated formally, thus will be used as bounding here. DADS data was examined to determine the forebay temperature, as indicated by computer Thursday, May 08, 2003 Page 34 of 48 Parameter Source Value Applicable Mode Est Inst Error Error Source Margin Margin Source Notes Surveillance Test Instrument(s)
Surveillance Limit point T413, for the past four years. TE738, which feeds T413, was inoperable from December of 2000 until September of 2001; a spot check of operator logs for the hottest part of the summer (July and August) shows UHS temperatures logged at around 77 degrees F maximum. Based on this examination, it appears that in the past four years forebay temperature did not rise above 85 degrees F. Thus, there is reasonable assurance that the UHS was not inadvertently inoperable, in the past four years, due to temperature instrument uncertainties.
CREVS Makeup Flow TS 4.7.6.1.e.3 300 +/- 10% cfm N/A N/A N/A N/A Nominal Value 1,2,3 & 4 Makeup air flow to the CREVS is required to maintain control room pressure at a positive value with respect to the surrounding spaces. Positive control room pressure is required in order to prevent the in-leakage of radioactive elements in the post-accident environment.
As described in the USAR, the CREVS is designed for 3300 CFM flow, including 300 CFM makeup flow.Calculations C-ME-028.01-003, "Control Room Cooling Load -Emergency Mode" and C-ME-028.01-007,"Analysis of Control Room Air Leakage" assume a nominal makeup air flow of 300 CFM.The surveillance Requirement of TS 4.7.6.1 .e.3 ensures that the system is capable of attaining the nominal design flow. The inclusion in the License value of +/- 10% tolerance is further evidence that this is a nominal value.Day Tank Volume -Value appears to be based on the ability to mitigate the initial stages of a LOCA without requiring makeup from the outside tank. The USAR capability statement of 20 hours of operation easily bounds this time period and appears to be arbitrarily based on the 6000 gallon tank delivered and the known vendor measured fuel consumption for the 4000 minimum level. There is no other analytical/accident basis for the 20 hour value.The Operations and Surveillance procedures require initiation of refill at 4750 gal., and margin is included in surveillance'procedures to account for instrument inaccuracies that might be introduced by the level .monitor. Margin is also included in the calculation (C-ME-024.01-005, Rev. 1), in the form of conservatisms.
Engineering judgment says that the combination of"margins" is sufficient to conservatively cover the existing instrument error.DB-SS-03710/3711 M&TE 300 +/- 10% cfm EDG Day Tank Volume TS 4.8.1.1.2.a.I/c.I
>/= 4000 gal.1,2,3 &4 N/A N/A 175 gal.(LI-l 121)/300 gal (LI-2787A or B)Surveillance Test DB-SC-3070, 3071, 3076, 3077 LI-I 121 LI-2787A/B 4000 gal.Thursday, May 08, 2003 Page 35 of 48 Parameter Source Value Applicable Mode EDG Fuel Storage Tank'Volume TS 4.8.1.1.2,a.2/c.2
>/= 32,000 gal.1,2,3 & 4 Est Inst Error Error Source Margin Margin Source N/A N/A 500 gal uncertainty 3000 gal operational Surveillance Test Notes Surveillance Test Instrument(s)
Surveillance Limit Outside Tank Volume -generally sized to allow a seven day supply though there are no specific requirements in the existing Reg. Guide for load carried or the duration to be carried. Since fuel economy is variable, but certainly less than the assumed 110% load factor, the sizing is a conservative design.Based on existing surveillance performance, the nominal 32K gal capacity is deemed adequate for LOCA/LOOP performance and when combined with the 4000 day tank, provides 7-day full load capability that is significantly greater than a design basis accident scenario requirements.
Margin is included in surveillance procedures to accomodate instrument inaccuracies and level is administratively controlled at greater than 34,500 gal. Margin is also included in the calculation (C-ME-024.01-005, Rev. 1), in the form of conservatisms.
Engineering judgment says that the combination of "margins" is sufficient to conservatively cover any instrument error introduced by the level indication loop.CR 02-03362.The 4160V frequency limits are based on the T-E 345 kV distribution system performance requirements of +/-1% and recent grid evaluations to ensure the operability of connected ESF loads during DBA conditions.
This results in an effective frequency operating band of 59.5 to 60.5 Hz.The 60 hz/900 rpm performance requirement set for the EDG is a nominal value and the Surveillance procedure tolerance limit of 59.9 to 60.1 (set by the MCB potentiometer) is acceptable to ensure the above limits are met. If found outside the band, the OP requires the operator to adjust back to the given tolerance.
The surveillance requirements do not include meter accuracy or governor performance.
During the emergency conditions ofa UV signal and/or Level 2 start, speed is controlled by the isochronous mode of the electronic governor.
The integral nature of the Woodward governor is such that frequency is driven back to setpoint continuously.
Therefore; the only real consideration is setup of the governor and the setting tolerance on this as-left value of the speed setpoint potentiometer.
Currently, this tolerance, which is conservative to the required frequency specification, and is, therefore, acceptable to ensure that connected load frequency remains within the prescribed limits.DB-SC-3070, 3071, 3076, 3077 LI-4891, 4892 35000 refill requirement EDO Speed TS 4.8.1.1.2.a.4/c.4 900 RPM 1,2,3 & 4 0.25% span equal to appro. 0.12 Hz Frequency transducer MCB meter 0.5 Hz Surveillance procedure DB-MM-09118 DB-SC-3070,3071 Frequency meter Woodward EGB Thursday, May 08, 2003 Page 36 of 48 Parameter Source Value Applicable Mode EDG Load TS 4.8.1.1.2.a.5/c.5
>/= 1000 kW 1,2,3 & 4 Est Inst Error Error Source Margin Margin Source N/A N/A 1100 kW, 1450 nominal Surveillance Test Notes Surveillance Test Instrument(s)
Surveillance Limit DB-SC-3076,3077 MCB kW meter 2100-2600 for 3 hours, 2450 nominal The basis for these requirements stems from Reg. Guide 1.9 and the older AEC Safety Guide 9 governing EDG test requirements.
The requirements for performance ofload-run tests contain many references to specific requirements at various test plateaus.
Statements such as "Demonstrate 90 to 100 percent of the continuous rating of the emergency diesel generator" and "operating at power factor between 0.8 and 0.9" are representative of the nominal nature of the requirements.
Likewise the load values presented in the refueling and semi-annual Tech Spec surveillances are arbitrary, rounded, values that are insensitive to minor variations based on any reading errors from the control panel meters. The intent of the testing is solely to demonstrate that the engine system can perform its design function through a simulated loading and unloading exercise.
For this to be valid, the engine needs to be loaded enough to reveal any mechanical/electrical deficiencies which might compromise performance under accident conditions.
This is done successfully in the referenced procedures.
In each case the required loading significantly exceeds the TS requirement (2450 kW nominal vs. 1000 kW) for load and meets all time requirements.
As such, minor variations in actual load which might be introduced into the acceptance criteria (due to meter/transducer uncertainty) have no impact on successful demonstration of the EDG system.The basis for these requirements stems from Reg. Guide 1.9 and the older AEC Safety Guide 9 governing EDG test requirements.
The requirements for performance of load-run tests contain many references to specific requirements at various test plateaus.
Statements such as "Demonstrate 90 to 100 percent of the continuous rating of the emergency diesel generator" and "operating at power factor between 0.8 and 0.9" are representative of the nominal nature of the requirements.
Likewise the load values presented in the refueling and semi-annual Tech Spec surveillances are arbitrary, rounded, values that are insensitive to minor variations based on any reading errors from the control panel meters. The intent of the testing is solely to demonstrate that the engine system can perform its design function through a simulated loading and unloading exercise.
For this to be valid, the engine needs to be loaded enough to reveal any mechanical/electrical deficiencies which might compromise performance under accident conditions.
This is done successfully in the referenced procedures.
In each case the required loading significantly exceeds the TS requirement (2450 kW nominal vs. 2000 kW) for load and meets all time requirements.
As such, minor variations in actual load which might be introduced into the acceptance criteria (due to meter/transducer uncertainty) have no impact on successful demonstration of the EDG system.EDG Load TS 4.8.1.1.2.d.3
>/= 2000 kW 1,2,3 & 4 N/A N/A 100 kW, 450 nominal Surveillance Test and Reg. Guide 1.9 criteria DB-SC-3070, 3071, 3076, 3077 MCB kW meter 2100-2600 kW 2450 nominal Thursday, May 08, 2003 Page 37 of 48 Parameter Source Value Applicable Mode EDG Load Est Inst Error Error Source Margin Margin Source N/A Notes Surveillance Test Instrument(s)
Surveillance Limit TS 4.8.1.l.2.d.4 N/A< 2838 kW 1,2,3 & 4 at least 238 kW Surveillance procedure 125 VDC Battery Float Charge Voltage TS 4.8.2.3.2.a.I/b.1, Table 4.8-1.0015 volts Voltmeter EDG Max load -To demonstrate that the total sum of connected emergency loads can be successfully carried by the EDG, the 2000 hour full load requirement of 2838 kW is selected to be conservatively bounding while not challenging the availability of the prime mover. The verification that this value is bounding is provided by engineering calculations which conservatively assess the DBA loadings.The surveillance procedure requirements verify that connected loads do not exceed this value, and therefore, remain within the capability of the engine during routine performance testing.In each case, the surveillance limits are well below the max rating of 2838 kW. As stated above, load tests are limited to 2100 -2600 kW with the target value being 2400-2500 kW. The refueling test in which actual ESF loading is verified is typically on the order of 1800-2000 kW and is, similarly, well below the limit.As such, additional consideration of transducer/meter accuracy is not warranted given the available margins.CR 02-06735, 02-08482, 02-06573, 02-06572, 02-06389, 02-06387, 02-05925, 02-05922, The periodic check of pilot cell voltage is done using precision M&TE (typically a Fluke voltmeter) with an accuracy that is an order of magnitude better than the measurement criteria.The error introduced by the M&TE is considered negligible compared with the allowed tolerance on the measurement.
As such, no additional adjustment to the surveillance limit for this uncertainty is required.The float charge voltage is measured at the battery terminals as part of the weekly battery/charger surveillance.
The limit includes a tolerance of+/-0.5 V and is set slightlyabove the TS minimum value.The error introduced by the M&TE is considered negligible compared with the allowed tolerance on the measurement.
As such, no additional adjustment to the surveillance limit for this uncertainty is required.No instrumentation is used for this determination.
As long as the water is nominally between the marks (the 1/4 overfill allows a slight margin from the target value of the upper mark during filling) the cell is acceptable.
The allowance of exceeding the upper level during equalization charge is consistent with the manufacturer's guidelines and the governing standard IEEE450.This determination is well within the skill of the technicians, and no additional compensation is requiied.DB-SC-3070, 3071, 3072, 3073, 3076, 3077 MCB kW meter 2100-2600 kW 2450 nominal DB-ME-3000 Fluke 8840 DVM (typical)>/= 2.13 V>/= 2.13 Volts/cell 1,2,3,4,5
& 6 125 VDC Battery Float Charge Voltage.015 volts DB-ME-3000 TS 4.8.2.3.2.a.2
>/= 129 Volts 1,2,3,4,5
& 6 Voltmeter 3 volts Surveillance Fluke 8840 DVM (typical)132 +/- 0.5 V 125 VDC Battery Electrolyte-Level TS 4.8.2.3.2.a.
I/b. 1, Table 4.8-1 Minimum level indication Mark<Level<I/4" above Maximum Level Indication Mark 1, 2,3,4,5 & 6 Thursday, May 08, 2003 N/A N/A DB-ME-3000 None Page 38 of 48 Parameter Source Value Applicable Mode 125 VDC Battery Specific Gravity TS 4.8.2.3.2.a.1, Table 4.8-1>/= 1.200 1, 2,3,4,5 &6 125 VDC Battery Specific Gravity TS 4.8.2.3.2.b.1, Table 4.8-1>/= 1.205 Average of all connected cells 1, 2, 3,4, 5 & 6 Est Inst Error-Error Source Margin Margin Source 0.001 g/cm2 Density meter 0.005 TS AV 0.001 g/cm2 Density meter Significant Load Calculation 0.2 deg.Thermometer Approx. 16 deg Operational-alarm Notes Surveillance Test Instrument(s)
Surveillance Limit 125 VDC Battery Electrolyte Temperature TS 4.8.2.3.2.b.3
> 60 Deg. F 1, 2, 3, 4, 5&6 The surveillance includes a manufacturer's nominal range of 1.205-1.225 within which no adjustment is required.
For low or high values a temperatureflevel correction may be made based on the test condition of the cell.The allowable value 1.195 (and the deviation from average spec) provides margin for individual cell variation.
In any case, the instrument uncertainty component is judged to be negligible based on the device accuracy and the measurement requirements.
The cell average spec ensures the overall health of the battery by maintaining the composite density in the range of the optimum manufacturer's criteria.
No additional special M&TE is required for this evaluation, since it is a combination of the individual cell measurements.
The battery load calculations which assess the design basis loading assumes a capacity of 80% and an aging factor of 1.32. These conditions would be indicative of a battery with a specific gravity (and cell voltage)considerably less than the Tech Spec allowable.
As such, it is concluded that significant margin exists in the surveillance of battery health and capacity relative to the assumed design basis performance.
The verification of battery cell temperature is performed during the quarterly check and involves measurement of every sixth cell. Though it is not credited-or required, cell temperature is checked during the weekly to determine the need to assess correction factors for specific gravity.During normal (and shutdown) operation, battery room temperature is maintained in the vicinity of 8OF and would not approach the lower limit of 60F unless there was a malfunction of the thermostat.
Annunciation of low temperature is provided by switches set at 76F with a +/- 1% tolerance.
Operation of these switches would pre-empt any drop in battery cell temperature prior to the lower limit being reached.It is concluded that sufficient margin exists between the operational condition and the lower operability limit, and that the alarm settings provide continuous protection for any degradation in HVAC conditions which might lead to the ambient temperature approaching the limit- Therefore, the negligible inaccuracy introduced during the quarterly surveillance by the M&TE need not be considered as a relevant in the evaluation of battery operability.
The DC bus voltage is evaluated during the weekly performance evaluation and is recorded by the Aux tour on the shiftly logs. The minimum value for the logsheet is 129 VDC and is taken from the local DC panel meters (which are similar in range and accuracy to the MCB meters).As long as the charger is connected, the expected voltage should be near the setting voltage of 134 VDC.Unless the charger is OOS or fails, the voltage cannot be below the logsheet range. Annuciation is provided to indicate conditions of charger failure. With the charger OOS, the battery must be declared inoperable until its standby is placed in service.The battery may be subsequently returned to service after voltage verification.
It is therefore concluded that the current surveillance limit and the monitoring frequency (provided by the logs) is sufficient to provide margin to the Tech Spec limit value without consideration of the small uncertainty introduced by the meter.DB-ME-3000 Density meter>1.200 DB-ME-3000 Density meter Average >1.205 DB-ME-3000/3001 Thermometer/Submersible probe>60 deg.125 VDC Bus Voltage TS 4.8.2.3.1>/= 125 Vdc 1,2,3&4+/- IV Panel Voltmeter 4v Surveillance DB-ME-3000 Ops logs P9-11656, 11743, 50015, 50014 or M&TE>129 V Thursday, May 08, 2003 Page 39 of 48 Parameter Source Value Applicable Mode Storage Pool Area Ventilation Pressure TS 4.9.12.1 Negative Pressure >/= 1/8 iwg Whenever Fuel is in the pool Containment Hydrogen Purge System -HEPA Filter Efficiency TS 6.8.4.f. I< 1% with flow at 100 cfm +/-10%Est Inst Error Error Source Margin Margin Source Notes Surveillance Test Instrument(s)
Surveillance Limit N/A N/A Significant Analysis Operability of the Fuel Bldg. Exhaust is required any time fuel is stored in the pool. However, consideration of a fuel drop event (on which the limit is based) is only required during movement of fuel/equipment in the pool or during Core Alts. This condition forms the basis of the specification and is reflected in the action statement.
Fuel reload has been completed at this time. Therefore, there is no immediate or near future challenge to the capability of this system. The negative pressure requirement was successfully demonstrated prior to the recent Mode 6 entry.Further evaluation of this parameter is not required for plant restart.The ANSI N5 10, Section 10, the In-Place HEPA filter leak test purpose is to verify acceptable system performance following system maintenance or filter replacement.
As stated in the USAR, "The filters are shop tested to measure the pressure drops across filter banks. The filters have a removal capability of`99.97 percent of the DOP smoke when shop tested with 0.3 micron diameter DOP. Also, the filters are in-place tested with DOP. Less than I% penetration and bypass leakage of DOP by each entire HEPA filter unit shall constitute acceptable performance for in-place testing.The 1% penetration requirement is not to be confused with the efficiency test of individual filters. That is, this test is intended to verify that the physical installation, if modified, does not permit bypass of the media.The overall 99% verified test performance of the unit provides conservative margin to the 95% efficiency credited in the analysis for the filter unit (charcoal/HEPA).
This test methodology and the M&TE used are consistent with NRC and industry practices, and demonstrate adequate performance.
No additional numerical adjustments are required.The ANSI N510, Section 10, the In-Place HEPA filter leak test purpose is to verify acceptable system performance following system maintenance or filter replacement.
As stated in the USAR, "The filters are shop tested to measure the pressure drops across filter banks. The filters have a removal capability of 99.97 percent of the DOP smoke when shop tested with 0.3 micron diameter DOP. Also, the filters are in-place tested with DOP. Less than 1% penetration and bypass leakage of DOP by each entire HEPA filter unit shall constitute acceptable performance for in-place testing.The 1% penetration requirement is not to be confused with the efficiency test of individual filters. That is, this test is intended to verify that the physical installation, if modified, does not permit bypass of the media.The overall 99% verified test performance of the unit provides conservative margin to the 95% efficiency credited in the analysis for the filter unit (charcoal/HEPA).
This test methodology and the M&TE used are consistent with NRC and industry practices, and demonstrate adequate performance.
No additional numerical adjustments are required.DB-SP-03318 M&TE 1%Shield Building Emergency Ventilation System -HEPA Filter Efficiency TS 6.8.4.f. 1< I% with flow at 8000 cfm +/-10%N/A DB-SS-3252/3253 M&TE< 1% with flow at 8000 cfm +/-10%N/A Significant N/A Thursday, May 08, 2003 Page 40 of 48 Parameter Source Value Applicable Mode CREVS -HEPA Filter Efficiency TS 6.8.4.f.1< 1% with flow at 3300 efn +/-10%Est Inst Error Error Source Margin Margin Source N/A Notes Surveillance Test Instrument(s)
Surveillance Limit N/A Significant N/A The ANSI N510, Section 10, the In-Place HEPA filter leak test purpose is to verify acceptable system performance following system maintenance or filter replacement.
As stated in the USAR, "The filters are shop tested to measure the pressure drops across filter banks. The filters have a removal capability of 99.97 percent of the DOP smoke when shop tested with 0.3 micron diameter DOP. Also, the filters are in-place tested with DOP. Less than 1% penetration and bypass leakage of DOP by each entire HEPA filter unit shall constitute acceptable performance for in-place testing.The 1% penetration requirement is not to be confused with the efficiency test of individual filters. That is, this test is intended to verify that the physical installation, if modified, does not permit bypass of the media.The overall 99% verified test performance of the unit provides conservative margin to the 95% efficiency credited in the analysis for the filter unit (charcoaUHEPA).
This test methodology and the M&TE used are consistent with NRC and industry practices, and demonstrate adequate performance.
No additional numerical adjustments are required.The HEPA filter efficiency and the charcoal filter efficiency are tested in-situ, using a challenge gas injected upstream of the filters. The test error is estimated by conservatively assuming 300% downstream concentration error(I.e., actual concentration is four times the measured concentration) and 20% upstream concentration error resulting in a penetration error: P=I00*Cd/Cu let Cd=.O0Cu P=I P+Perror = 100* (Cd+Cderror)/(Cu+Cuerror) 1-+ perror= 100*(.0lCu+.03Cu)/(Cu+/-.2Cu) maximum Perror=4/(I-.2)-l
=4%.Realistic expected errors might be 10% for both upstream and downstream concentrations, thus this is a very conservative estimate of error.Margin: Calculation 31.01"H2 Purge Doses After LOCA" assumes 95% filter efficiency.
The above shows that even with extremely conservative assumption of instrument error there is adequate margin to provide reasonable assurance that additional consideration or application of error is not required prior to restart.DB-SS-03145/3146 M&TE< 1% with flow at 3300 cfm +/-10%Containment Hydrogen Purge System -Charcoal Filter Efficiency TS 6.8.4.f.2< 1% with flow at 100 cfm /-i 10%Penetration
=4%DB-SP-03318
+ Eng.Judgment see notes 4%System Description/Calc.
31.01 DB-SP-03318 M&TE 1%Thursday, May 08, 2003 Page 41 of 48 Parameter Source Value Applicable Mode Shield Building Emergency Ventilation System -Charcoal Filter Efficiency TS 6.8.4.f.2< 1% with flow at 8000 cfm +/-10%Est Inst Error Error Source Margin Margin Source Penetration
= 4%see notes 4%SD-22C Notes Surveillance Test Instrument(s)
Surveillance Limit CREVS -Charcoal Filter Efficiency TS 6.8.4.f.2< 1% with flow at 3300 cfm +/-10%Penetration
= 4%see notes 4%C-NSA-028.01-005 The HEPA filter efficiency and the charcoal filter efficiency are tested in-situ, using a challenge gas injected upstream of the filters. The test error is estimated by conservatively assuming 300% downstream concentration error(l.e., actual concentration is four times the measured concentration) and 20% upstream concentration error resulting in a penetration error: P=-I00*Cd/Cu letCd=.OlCu P=I P+Perror = I 00*(Cd+Cderror)/(Cu+Cuerror) 1+ perror = 100*(.O1Cu+.03Cu)/(Cu+/-.2Cu) maximum Perror = 4/(1-.2)-I
= 4%.Realistic expected errors might be 10% for both upstream and downstream concentrations
[P=100* Cd/Cu]thus this is a very conservative estimate of error.Margin: Dose calculations and system design require only 95% filter efficiency.
The above shows that even with extremely conservative assumption of instrument error there is adequate margin to provide reasonable assurance that additional consideration or application of error is not required prior to restart.The HEPA filter efficiency and the charcoal filter efficiency are tested in-situ, using a challenge gas injected upstream of the filters. The test error is estimated by conservatively assuming 300% downstream concentration error(I.e., actual concentration is four times the measured concentration) and 20% upstream concentration error resulting in a penetration error: P=100*Cd/Cu let Cd=.OICu P=I P+Perror=
00*(Cd+Cderror)/(Cu+Cuerror) 1 + perror= 1000(.OlCu+.03Cu)/(Cu+/-.2Cu) maximum Peror = 4/(l-.2)-l
= 4%.Realistic expected errors might be 10% for both upstream and downstream concentrations
[P=-I00* Cd/Cu]thus this is a very conservative estimate of error.Margin: Calculation C-NSA-28.01-005 "Control Room Radiation Doses Following A LOCA" assumes 95%filter efficiency.
Other Control Room dose calculations(e.g., C-NSA-028.01-001 "Control Room radiation Doses without Isolation During a Waste Gas Tank Rupture and Fuel handling accident" &C-NSA-028.01-002 "Control Room Radiation Dose Due to the Fuel handling Accident Inside Containment")
take no credit for the CREVS.The above shows that even with extremely conservative assumption of instrument error there is adequate margin to provide reasonable assurance that additional consideration or application of error is not required Estimate of instrument error.Reference accuracy of PD15059B is 0.5%; assume a total accuracy of four times the reference accuracy = 2.0% or 5 iwg.Margin: From the System Description,'the Filter maximum design dP, including pressure drops across the filter vessel and hardware, is 11.11 iwg at 100 cfm.There is adequate margin to provide reasonable assurance that additional consideration or application of uncertainty is not required prior to restart.DB-SS-3252/3253 M&TE< 1% with flow at 8000 cfm +I-10%DB-SS-03145/3146 M&TE< 1% with flow at 3300 cfm 4/-10%Containment Hydrogen Purge System -Filter Delta-P TS 6.8.4.f.4<25 iwg with flow at 100 cfm +/-10%5.0 iwg EMPAC and engineering judgement 13.89" SD-022D, Table 2.4-1 DB-SP-03318 PD15059B< 25 iwc Thursday, May 08, 2003 Page 42 of 48 Parameter Source Value Applicable Mode Shield Building Emergency Ventilation System -Filter Delta-P TS 6.8.4.f .4 6 iwg with flow at 8000 cfm +/-10%Est Inst Error Error Source Margin Margin Source 0.96 iwg EMPAC See notes 2.1 iwg SD-022C, Table 2.4-: max clean filter dP=3.9 iwg 0.8 iwg Notes Surveillance Test Instrument(s)
Surveillance Limit CREVS -Filter Delta-P TS 6.8.4.f.4 EMPAC see notes 1.0 iwg see notes Estimated instrument error: The total pressure drop is calculated from the sum of the pressure drop indications from PDIS020A, B C & D. PDI5020A, C & D are Dwyer model 2002C (range 2 iwg) dp gauges with a reference accuracy of +/-2%. PDI5020B is a Dwyer model 2006C (range = 0-6 iwg)dP gauge with a reference accuracy of +/-2%. Take the measurement uncertainty to conservatively be 4 times the reference accuracy, or +/- 8%..08-(2.0 + 6.0 +2.0 + 2.0)iwg = 0.96 iwg The above is for Train 1, Train 2 is assumed to be identical.
There is adequate margin to provide reasonable assurance that additional consideration or application of uncertainty is not required prior to restart.Estimated instrument error: dP is measured by the sum of indications from three dP gauges DP15293, DP15294 & DP15295 indicating pressure drop across the Carbon Filter, HEPA filter and prefiter respectively(for Train 1. Train 2 is taken to be identical to Train 1 here). DP15293 and DPI5295 are Dwyer model 2002C dP gauges (0-2 iwg) with a reference accuracy of+/-2%. DP15294 is a Dwyer model 2006C dP gauge (0-6 iwg)with a reference accuracy of +/-2%. The measurement uncertainties are conservatively assumed to be 4 times the reference accuracy or 8%.0.08-(2.0
+2.0+6.0)iwg
= 0.8 iwg Margin: From SD-029B, Table 2.4-1, the design pressure drop across the HEPA filter is 1.0 iwg at 3300 CFM and the pressure drop across the charcoal filter is 1.0 iwg at 3300 CFM, assume 0.4 iwg for system losses and 1.0 iwg for prefilter pressure drop: 4.4-1.0-1.0-0.4-1.0 iwg = 1.0 iwg There is adequate margin to provide reasonable assurance that additional consideration or application of uncertainty is not required prior to restart.M&TE model/accuracy requirements are unspecified in the test procedure; 3% is the algebraic sum of 2%current measurement accuracy and 1% voltage measurement accuracy; this is conservative based on a)specification accuracies are smaller and b) the three-phase values are averaged, "This assumes that the acceptance criteria of +/- 20% can be utilized to offset measurement errors Based on the above, there is adequate margin to provide reasonable assurance that additional consideration or application of uncertainty is not required prior to restart.DB-SS-3252/3253 PDI5020A, PDI5020B, PDI5020C & PD15020D 6 iwg with flow at 8000 cfm +/-10%DB-SS-03145/3146 DP15293, DP15294 & DP15295 4.4 iwg with flow at 3300 cfm +/-10%4.4 iwg with flow at 3300 cfm +/-10%Containment Hydrogen Purge System Heat Dissipation TS 6.8.4.f.5 2000 watts +/- 20%3%*FLUKE Doc. 8 2156, Specifications for Fluke Model I 10 True RMS Multimeter
+/- 20%TS** I DB-SP-3318 M&TE -unspecified Thursday, May 08, 2003 Page 43 of 48 Parameter Source Value Applicable Mode Boric Acid Addition Tank Volume TRM 3.1.2.9.a.
In accordance with Figure 3.1-1 1,2,3&4 Boric Acid Addition Tank Concentration TRM 3.1.2.9.b 7875 ppm to 13,125 ppm 1,2,3 & 4 Boric Acid Addition Tank Solution temperature TRM 3.1.2.9.c Est Inst Error Error Source Margin Margin Source N/A N/A approx. 600 gal.Analysis Notes Surveillance Test Instrument(s)
Surveillance Limit DB-CH-03027 LI-MU-49-2, LI-MU-65-2
>/= curve limits The TRM provides a range of values for concentration based on the available volume in the BAAT.Per the TRM Bases, the TRM figure has been conservatively increased to account for instrument and chemical analysis tolerance.
Although not quantified here, this provides resonable assurance for restart.The estimate of boron sampling accuracy is based on discussions with Chemistry personnel and considers the concentration range in question.+/- 20 ppm DB-CH-03027 Titration system The TRM Bases provides the minimum indicated values in the Table, and it varies according to the contained volume. The analytical basis assumes very conservative conditions of fuel reactivity, xenon worth (none), approx. 300 ppm and low temperature.
Also, as stated in the TRM bases, the values have been conservatively increased to account for instrument and chemical analysis tolerances.
Sampling Titration analyzer 7875 ppm to 13,125 ppm Derived from TRM Figure 3.1-1 and Bases+/- 3 deg.Temperature cell Although not quantified here, the combination of the above conservatism provides a reasonable assurance that the BAT tank limits will ensure sufficient SDM capability.
The surveillance limit is to provide verification that the solution temperature of the Concentrated Boric Acid Storage System is >/= to 105 deg F. The temperature limit is established to preclude the precipitation of the boric acid solution and is based on maintaining the crystallization temperature for a solution of 7 weight percent boric acid (95 deg F).This is 10 deg F lower than the surveillance limit and provides an allowance for any measurement errors/margin associated with the process. It is concluded that the surveillance limit is acceptable for this parameter and contains sufficient conservatism to preclude the addition of temperature measurement penalties.
LIR CR 02-07293 re: CCW Surge Tank Level Switches LSLL3757A/
B/C & 3758A/B/C
*Engineering judgment based on +1-1/4" dimensional tolerance for switch and 2" calibration tolerance; 4" assumed to be bounding for other uncertainty effects DB-CH-03027 TI-MU-48-1, TI-MU-64-1I
>/= 105 deg. F.10 deg.>I= 105 deg. F.1,2,3 & 4 Surveillance CCW Surge Tank Level TS 3.7.3.l(CR 02-07293)A: 45" dec., B&C: 35" dec.1,2,3 & 4 t/- 4" M-362-48-3", Inst.Data Sheet 9", C-NSA-0I 6.04-002 (CCW Surge Tank Level Switches LSLL3757A/
B/C &Thursday, May 08, 2003 Page 44 of 48 Parameter-Source Value Applicable Mode Refueling Water Level TS 3.9.8.1>/= 23 feet 6 when water level above the Refueling Water Level TS 3.9.8.2>/= 23 feet 6 when water level above the Est Inst Error Error Source Margin Margin Source Notes Surveillance Test Instrument(s)
Surveillance Limit Significant Safety Analysis Significant Safety Analysis The current refueling cavity level is maintained at 23.6 ft. to support the core reload for Cycle 14. Though it is not clear whether or not instrument uncertainty was included in the safety analysis for the fuel drop accident-loss of cooling event, it is clear that significant conservatisms exist with respect to the iodine scrubbing analysis, gap activity released, bumup history, and enrichment.
In the case of the cavity, the actual level over the fuel is on the order of 45 feet when the cavity level is flooded to the nominal 23 ft. value. The SFP level does not contain this margin but is only a challenge to the analysis during the movement of irradiated fuel.Since this activity has been completed, no additional evaluation is required for this parameter prior to restart.The current refueling cavity level is maintained at 23.6 ft. to support the core reload for Cycle 14. Though it is not clear whether or not instrument uncertainty was included in the safety analysis for the fuel drop accident-loss of cooling event, it is clear that significant conservatisms exist with respect to the iodine scrubbing analysis, gap activity released, burnup history, and enrichment.
In the case of the cavity, the actual level over the fuel is on the order of 45 feet when the cavity level is flooded to the nominal 23 ft. value. The SFP level does not contain this margin but is only. a challenge to the analysis during the movement of irradiated fuel.Since this activity has been completed, no additional evaluation is required for this parameter prior to restart.The special test exemption (suspending shutdown margin) is not currently exercised during physics testing at DB. The current method of continuous boration of a full length CRAs maintains SDM throughout and does not require repositioning of APSRs.Therefore, evaluation of the 35% withdrawal limit for APSRs is not required.SD-023 The H2 Dilution System introduces air into the containment to the hydrogen concentration when hydrogen concentration reaches 2% by volume. The system is designed to provide approximately 100 scfm for a containment pressure range of 0 to 18 psig which envelops the requirement of 43 scfm for these pressure ranges. Any measurement uncertainties are accounted for in the design specifications/sizing for the hydrogen dilution blowers (I.e. rated flow of 100 scfm and design pressure of 25 psig). It is concluded that there is adequate margin to account for any measurement uncertainties and that the surveillance limit is acceptable for this parameter.
Shift Logs LII600 SFP L11627 Cavity 23.0 Shift Logs LI 1600 SFP LI1627 Cavity 23.0 APSR Position TS 4.10.4.3 N/A N/A>/= 6.5%2 Containment Hydrogen Dilution System Discharge Pressure TS 4.6.4.3.b N/A N/A DB-SP-3320 DB-SP-3321 PI-5147 PI-5069>=15 psig and <25 psig 15 psig I &2 Significant Surveillance Procedure Thursday, May 08, 2003 Page 45 of 48 Parameter Source Value Applicable Mode Control Room Air Temperature TS 4.7.6.l.a Est Inst Error Error Source Margin Margin Source 4 Deg. F VMAN G-ML-0471 5 Deg. F System Descriptions Notes Surveillance Test Instrument(s)
Surveillance Limit</= 110 Deg. F 1,2,3&4 125 VDC Bus Voltage+/- IV TS 4.8.2.4.1>/= 125 VDC Panel Voltmeter 4V-Surveillance Estimated instrument error: The Surviellance procedure DB-OP-03006 requires that control room temperature be recorded from the control room hygrothermograph or M&TE as specified by the Shift Manager.The hrgrothermograph has a reference indication accuracy of +/- 1 deg. C, or 1.8 deg. F. 4 Deg. F was estimated to conservatively bound any drift error and other measurement error effects.From System Description for RPS/SFAS/SFRCS:
NIIRPS, SFAS & SFRCS equipment located in the control room is designed to operate at temperatures up to 110 Deg. F. If control room temperature exceeds 110 Deg.F, then RPS & SFAS would have to be declared inoperable.
Procedure DB-OP-02533 provides instructions for the actions required when Control Room temperature exceeeds 90 degrees F. This procedure requires that IF control Room temperature exceeds 105 deg. F, THEN the a rapid reactor shutdown is performed.
This action is more conservative than the action that would be required by TS 3.0.3 if temperature exceeded 110 deg. F. Thus we can take credit for the procedural requirement as conservative margin (see definition of MARGIN in the guideline).
The DC bus voltage is evaluated during the weekly performance evaluation and is recorded by the Aux tour on the shiftly logs. The minimum value for the logsheet is 129 VDC and is taken from the local DC panel meters (which are similar in range and accuracy to the MCB meters).As long as the charger is connected, the expected voltage should be near the setting voltage of 134 VDC.Unless the charger is OOS or fails, the voltage cannot be below the logsheet range. Annuciation is provided to indicate conditions of charger failure. With the charger OOS, the battery must be declared inoperable until its standby is placed in service.The battery may be subsequently returned to service after voltage verification.
It is therefore concluded that the current surveillance limit and the monitoring frequency (provided by the logs) is sufficient to provide margin to the Tech Spec limit value without consideration of the small uncertainty introduced by the meter.Leakrate testing is performed in accordance with the guidance of Appendix J and its associated Regulatory Guides (1.163, NUREG 1493) and test procedures.
Accomodations are made in the testing methodology standards for the use of various types of M&TE (ANSI/ANS 56.8-1994) in the various battery of tests which make up the ILRT program. This combined with conservative acceptance criteria (<.75 La) provides adequate margin to demonstrate containment operability.
As stated in its 1995 revised rulemaking concerning performance-based leakrate testing,"The NRC has decided, in general, to maintain the present level of prescriptiveness in the proposed rule and, in particular, to not decrease further the test frequency for ILRTs. The NRC's position is guided by the desire to maintain some conservatism to address uncertainties and adopt an evolutionary approach wherein incentives remain for good performance." No additional adjustments are required at this time.DB-OP-03006 CR Hydrothermograph or M&TE</= 110 Deg. F DB-ME-3000 Ops logs P9-11656, 11743, 50015, 50014 or M&TE 5&6 Containment Leakage Rate TS 6.16 N/A N/A As specified Thursday, May 08, 2003 Page 46 of 48 Parameter Source Value Applicable Mode Internals Vent Valve Opening Force TS 4.4.10.1.b.3
</= 400 lbs.All Modes Est Inst Error Error Source Margin Margin Source 10 lbs.see notes I 1 lbs.C-NSA-062.01-004
+/-30 psig WR Pressure 0-3000#100 psig Heatup procedure Notes Surveillance Test Instrument(s)
Surveillance Limit RCS Pressure TS 3.5.1>800 psig 3 Estimated instrument error: The alternate method of measurement in DB-MM-03002 uses a spring scale, which will be more inaccurate than the preferred method, which uses a load cell. Assume a spring scale with 0.5% accuracy and a range of 0-1000 lbs, with a resolution of+/- 51bs. The algebraic sum of the errors is 5+5 10 lbs.Safety analyses assume that the valve will open fully at a differential pressure of 0.26 psid, which is equivalent to a vertical force of 125 lbs in air. Calculation C-NSA-062.01-004 determines that the 125 lbs.In air is equivalent to 109 lbs in water, and adds 11 pounds for measurement uncertainty.
This value is the minimum operability range for the Core Flood Tanks. Although cooling requirements may decrease as power decreases, the CFTs are still required to provide core cooling as long as elevated RCS pressures and temperatures exist. In MODE 3 with RCS pressure < 800 psig, and in MODES 4, 5, and 6, the CFT motor operated isolation valves are closed to isolate the CFTs from the RCS. This allows RCS cooldown and depressurization without discharging the CFTs into the RCS or requiring depressurization of the CFTs.This LCO is only applicable at pressures
> 800 psig. Below 800 psig, the rate of RCS blowdown is such that the safety injection pumps can provide adequate injection to ensure that peak clad temperature remains below the 10 CFR 50.46 limit of 22001F.The value of 800 # is not specifically modeled in the LOCA analysis.
Similiarly, the generous LCO time allowance indicates that this is a nominal value. During plant heatup, the CFT's are placed in service between 675 and 700 psig, which provides sufficient margin to the LCO limit to allow verification of leak performance and valve position.Without a specific analytical limit, the application of additional uncertainty is not appropriate.
The requirement for the Channel Functional Test of the AFW Inlet Steam pressure protection ensures that the actuation circuitry will be available during operation in the range of High Energy Line Breaks (>275#, >200 F). The requirement is nominal and is not explicitly modeled in the accident analysis.
This is further demonstrated by the allowance of 24 hours to perform the testing "after exceeding 275 psig during each plant startup, if the test has not been performed within the last 31 days." DB-MM-03002 M&TE</= 120 lbs DB-OP-6900 WR Pressure>800 #Steam Line Pressure 30#WR pressure loop TS 4.7.1.2.2> 275 psig 1,2&3 TS Surveillance It should be noted that the OPERABILITY of the Auxiliary Feed Pump Turbine Inlet Steam Pressure Interlocks is required only for high energy line break concerns and does not affect Auxiliary Feedwater System OPERABILITY.
As such, no consideration of additional instrument uncertainty is required to support this surveillance value.Thursday, May 08, 2003 Page 47 of 48 Parameter Source Value Applicable Mode Fuel Handling Building Crane Loads TS 4.9.7</= 2430 lbs.With Fuel Assemblies in the Spent Fuel Storage Limits TS 4.9.13.1 Per Figures 3.9-1, 3.9-2 & 3.9-3 With Fuel Assemblies in the Est Inst Error Error Source Margin Margin Source N/A N/A Notes Surveillance Test Instrument(s)
Surveillance Limit The operation of the handling cranes ensures that heavy loads carried in the vicinity of the SFP do not exceed the limits of the Fuel Handling Accident.
However, this analysis is not related to unit operation and should not be considered as required for restart evaluation.
N/A Not Mode Related -While important to safe plant operation, this parameter is not specifically critical to plant restart. This issue will be revisited post-restart.
N/A N/A N/A Thursday, May 08, 2003 Page 48 of 48 Parameter CTS/ITS Margin Assessment Barrier BWST Temperature 4.5.4.b The Tech Spec values for low and high limit are Calc C-ICE-049.01-001 stipulates the required SR 3.5.4.1 those assumed in the respective analyses.
instrument uncertainty compensation.
Daily surveillance procedure DB-OP-03007 provides the corrected verification limits.Core Flood Tank 3.5.1.b The Tech Spec values for CFT low and high limit are Calc C-ICE-051.01-001 stipulates the Volume SR 3.5.1.2 conservative with respect to the analytical limits, but additional required instrument uncertainty not include all applicable instrument uncertainty, compensation.(Note: Davis-Besse has agreed to change the Tech Spec limits such that all instrument uncertainty is Shiftly surveillanceprocedure DB-OP-03006 included).
provides the corrected verification limits.Core Flood Tank 3.5.1.d The Tech Spec values for CFT low and high limit are Calc C-ICE-051.01-002 stipulates the Pressure SR 3.5.1.3 conservative with respect to the analytical limits, but additional required instrument uncertainty do not include all applicable instrument uncertainty, compensation.(Note: Davis-Besse has agreed to change the Tech Spec limits such that all instrument uncertainty is Shift surveillance procedure DB-OP-03006 included), provides the corrected verification limits.Ultimate Heat Sink 4.7.5.1.a The Tech Spec value for UHS level is the level Calc C-ICE-009.01-002 stipulates the Level SR 3.7.9.1 assumed as the initial condition for the respective additional required instrument uncertainty analyses.
compensation.
Daily surveillance procedure DB-OP-03007 provides the corrected verification limits.Ultimate Heat Sink 4.7.5.1 .b The Tech Spec value for UHS temperature is the Calc C-ICE-009.01-001 stipulates the Temperature SR 3.7.9.2 temperature assumed as the initial condition for the additional required instrument uncertainty respective analyses.
compensation.
Daily surveillance procedure DB-OP-03007 provides the corrected verification limits.
NRC ITS Tracking Page I of 16 Return to View Menu e RAI Screening Required:
Yes Status: Closed This Document will be approved by: Gerald Regulatory Basis must be included in Comments Waig; Tim Kobetz section of this Form This document has been reviewed and- Yes information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING NRC Reviewer IDD 200801021633 Conference Call Requested?
No Category Other Technical Challenge ITS Section: TBP.OC:. JFD..Number.
Page.Numb-er(s)-:
ITS 3.5 Ross Telson None 83 Information ITSNunib.er:
OS1: DOC-Number:1 Bases.JFD Number: None None None None REF: Question ID 200711161716; Licensee Response by Bill Bentley on 12/18/2007; item (2):----- ACTIONS NEEDED: 1. Provide appropriate markups to the ITS Conversion Section 3.5.4 to correct not having adequately addressed the term "available" in CTS LCO 3.5.4.a (Page 83) as part of the conversion.
: 2. Provide CR 02-04514 and applicable calculations to aid 'staff review and understanding of the.adequacy of the margin provided by the proposed BWST volume limits.----- BASIS FOR REQUEST: 1. Davis-Besse has identified the need for a change to ITS Specification 3.5.4.CTS LCO 3.5.4 (Volume 10, Page 83) states "The Borated Water Storage Tank (BWST) shall be OPERABLE with: a. An AVAILABLE borated water volume of between..." ISTS SR 3.5.4.2 (Page 87) verifies "BWST borated water volume is..." The values associated with the term "AVAILABLE" in CTS LCO 3.5.4.a (Page 83) are not consistent with the ITS bases statement (Page 91): "Note that the volume limits refer to TOTAL, rather than USABLE, volume required to be in the BWST..." The CTS BWST water volume limits are based on the water volume that is AVAILABLE above the top of the discharge line penetration into the bottom of the tank. The top of the discharge line is 4 inches above the bottom of the BWST.Davis-Besse requests that an RAI question be posted against ITS Specification 3.5.4, so that we may provide appropriate markups to the ITS Conversion to address these two changes.http://www.excelservices.com/eXceldbs/itstrack davisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 16 2. CR 02-06407, provided by licensee, documented an adverse condition regarding insturment uncertainty for Non-LSSS applications.
The adverse condition was deemed to be sufficiently significant to warrant a root cause determination.
Corrective Action 6 concluded that adequate TS margin could not be demonstrated without detailed calculations or engineering evaluations to quantify the applicable instrument uncertainty and establish appropriate surveillance guidelines for BWST Volume (as well as 9 other plant parameters).
Pre-existing CR 02-05157 was referenced with regard to BWST Volume but was not provided with CR 02-06407.---- REGULATORY REQUIREMENT:
&sect; 50.36 Technical Specifications (a) Each applicant for a license ... shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications
... shall also be included in the application, but shall not become part of the technical specifications.(b) ... The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to &sect; 50.34. The Commission may include such additional technical specifications as the Commission finds appropriate.(c) Technical specifications will include items in the following categories:
(2) Limiting conditions for operation.(i) Limiting conditions for operation are the lowest functional capability or Comment performance levels of equipment required for safe operation of the facility.(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria: (C) Criterion
: 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.(3) Surveillance requirements.
Surveillance requirements are requirements
...to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.&sect; 50.46 Acceptance Criteria For ECCS Requires, in part, that uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated.
This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of this section, there is a high level of probability that the criteria would not be exceeded.&sect; 50.92 Issuance of amendment.(a) In determining whether an amendment to a license ...will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses or construction permits to the extent applicable and appropriate.
Issue Date 01/02/2008 IF http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 3 of 16 Close Date 07/02/2008 Logged in User: Anonymous'vResnonses NRC Response by Ross Telson on 01/15/2008
: 1. This is a correction to ACTIONS NEEDED No. 1. CR 02-04514 was listed in error. Please replace CR 02-04514, which addresses establishment of adequate TS margin for HPI and LPI flow, with CR 02-05157 and CR 03-02644, which appropriately address the establishment of adequate TS margins for BWST Volume and Boron Concentration.
CR 02-04514 and the establishment of adequate TS margins for HPI and LPI flow will be addressed in a separate Q&A Thread. 2. Revise ACTIONS NEEDED No. 1 into the following sub-parts:
: a. Confirm that CR 02-05157 and CR 03-02644 have been resolved.
: b. Briefly summarize how adequate margin was established between the licensee-controlled surveillance procedure, the TS SR acceptance criteria, and the analyses and evaluation included in the safety analysis report, such that (i) meeting the associated surveillance procedure requirements assures the TS SR acceptance criteria are met and (ii) meeting the TS SR acceptance criteria assures the necessary BWST qualities (Volume and Boron Concentration) are maintained, that BWST operation will be within safety limits, and that the limiting condition for operation of the BWST will be met.c. Provide CR 02-05157 and CR 03-02644.
: d. Provide additional existing calculations, assumptions, or analyses you deem necessary or appropriate to support a. and b., above.Licensee Response by Jerry Jones on 02/11/2008 Action 1 Response:
CTS 3.5.4.a and 4.5.4.a (Volume 10, Page 83)has been changed to relocate the word "available" to the Bases.Additionally, a new Discussion of Change (DOC) LAO 1 (Page 84)and a new Bases Justification for Deviation (JFD) 9 have been added. Furthermore, changes have been made to Bases markups (Pages 91, 92, and 94) to address the relocation of "available." A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
Action 2 Response 2.a: CR 02-05157 and CR 03-02644 have been resolved.
2.b: An overview of CR 03-02644 regarding Boron concentration is as follows: For BWST boron concentration, the two values are 2600 and 2800 ppm. The low limit of 2600 ppm is based on Framatome calculation 5 1-501665 1-01, DB Cycle 14 Boron Calculations.
The minimum value to maintain 1% dK/K was 2312 ppm. Since this would result in 288 ppm of margin, no calculation was deemed necessary for the low concentration value (i.e. any instrument errors are bounded by the margin). The 2800 ppm value is based on Framatome calculation 86-5006059-00 for boron precipitation analysis.
The analysis used 550,000 gallons injected from the BWST at 2800 ppm, 2140 cu ft at 3500 ppm from the Core Flood Tanks, and the RCS at 11500 cu ft at 1900 ppm. The uncertainty assumed for titration of the samples was 1%. Sincethere is never 550,000 gallons available and/or injected into the core/containment, the http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 4 of 16 amount of available boron would be lower by more than the 1%uncertainty.
Based on that, it was determined that no instrument uncertainty adjustment was required.
An overview of CR 02-05157, Corrective Action 2, regarding BWST Volume is as follows: 550,000 gallon Tech Spec Limit The 550,000 gallon limit is not instrument uncertainty adjusted.
This is the maximum volume that is used in analyses for tri-sodium phosphate neutralization of the boric acid and for flooding level in containment.
The revision to Calculation C-ICE-048.01-004 provided a maximum allowed indicated level that is adjusted for level indication uncertainty and readability.
This maximum allowed indicated level of 41 feet is stipulated in the surveillance procedure.
500,100 gallon Tech Spec Limit The revision to Calculation C-ICE-048.01-004 demonstrated that when accounting for the below listed items, a surveillance that ensures at least 500,100 gallons is available in the BWST ensures the analytical limit is protected.
This minimum allowed indicated level of 39 feet is stipulated in the surveillance procedure:
-Existing permissive setpoint -Level required for 360,000 gallons -Level Indicati6n uncertainty
-Readability of the Level Indication The 500,100 gallon limit is instrument uncertainty adjusted.
For background, a summary of how the limit was obtained is provided:
Starts with a minimum level above the discharge pipe to ensure adequate protection from vortexing.
Adjustment is made for Operator response time to initiate the ECCS pump suction transfer from the BWST to the CTMT emergency sump Adjustment is made for instrument uncertainty to provide the permissive setpoint.
This permissive setpoint is established to provide protection' against vortexing if the transfer is started too late due to instrument error.The Allowable Values on either side of the trip setpoint are 101.6 and 115.4 inches, and are provided in ITS 3.3.5, Parameter 5.Adjustment is made for the minimum volume assumed in the analyses of 360,000 gallons injected into the core/containment.
Adjustment is made for instrument uncertainty.
This additional uncertainty is included to ensure the transfer is not started too early due, to instrument error, and thus not meeting the 360,000 gallon requirement.
The result of these adjustments is calculated to be a minimum value of 500,051 gallons. As stated in the CTS Bases, the Tech Spec limit of 500,100 gallons was conservatively rounded up from the calculated value. Note that the draft markup to the ITS Bases (Page 92) has been altered accordingly.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
2.c: A copy of CR 02-05157 and CR 03-02644 are provided with the names redacted.
2.d: The values listed in the Technical Specifications for BWST Temperature are 35 F and 90 F. Neither value includes instrument uncertainty, and as stated in the ITS Bases draft markup, the' instrument uncertainty adjustment is provided in the surveillance procedure.
The BWST Temperature is monitored by procedure DB-OP-03007, Daily Instrument Check. The temperatures used for surveillance are http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS. Tracking.Page 5 of 16 38.4&deg;F and 86.6 0 F. These values are instrument uncertainty compensated to account for 3.4&deg;F of instrument uncertainty when using computer points T064 and T065. The 3.4&deg;F value is derived in calculation C-ICE-049.01-001.
NRC Response by Ross Telson on 02/12/2008
: 1. Please confirm the range of "available" gallons of boric acid in the BWST represented by an indicated level of 39 feet, after accounting for uncertainty/drift/etc. (e.g. 500,100 --< X --< ???)Confirm lower limit & provide ???. 2. Please confirm the range of"available" gallons of boric acid in the BWST represented by an"indicated" level of 41 feet after accounting for uncertainty/drift/etc. (e.g. ??? :5: x --< 550,000) Confirm upper limit & provide ???. 3. Please correct Action 2 Response 2.b.which states, in part, "...there is never 550,000 gallons available and/or injected into the core/containment..." Based on SR 3.5.4.2 and associated bases, a correct statement would be "...there is never more than 550,000 gallons available and/or injected into the core/containment..." 4. Please amend B-JFD 9 to justify why it is necessary to deviate from NUREG BASES APPLICABLE SAFETY ANALYSES and SURVEILLANCE REQUIREMENTS associated with SR 3.5.4.2. (e.g. Why the values for maximum and minimum allowed water volume can not be computed to "refer to total, rather than usable, volume" as stated in the NUREG. B-JFD 9 states "Changes have been made to explain that the minimum and maximum BWST volumes are the available volume." B-JFD 9 is a statement
-not a justification.
NRC ADMIN. LTR. 96-04: Efficient Adoption Of Improved Standard Technical Specifications states, in part, "... the NRC continues to believe that total adoption of the improved STS will substantially improve the efficiency of the regulatory process, and ensure that licensee and NRC resources are applied to significant safety matters..." and "The major objective of converting from plant-specific technical specifications to the improved STS is to achieve as much consistency in the license requirements as possible, to the extent that the plant-specific design basis can conform with the related typical plant design reflected in the improved STS." Licensee Response by Bill See attached file for responses to Items #1 -#4 in the NRC Bentley on 02/20/2008
]{response dated 2/12/08.Licensee Response by Bill During a 3/5/08 phone call with the reviewer, additional Bentley on 03/06/2008 information was requested.
The attached file is believed to contain the requested information.
Licensee Response by Bill During a 3/7/08 call with the reviewer, additional information was Bentley on 03/08/2008 requested.
The attached file includes portions of the calculation for vortex formation with ECCS pump suction from the BWST, and is believed to contain the requested information.
NRC Response by Ross Telson on 03/21/2008 Thank you for the additional information.
To close out this question thread, please confirm the reviewer's understanding as summarized below, or provide corrections as needed...
and address question 9: Staff understand the licensee has stated that proposed ITS SR 3.5.4.2, as implemented by licensee-controlled procedures, http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1fddcea1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 6 of 16 assures and will continue to assure (a) that the necessary quality of the BWST is maintained, (b) that the BWST will be operated in a manner that supports facility operation within safety limits, (c) that the BWST limiting conditions for operation will be met, and (d)that the proposed SR acceptance criteria are consistent with analyses and evaluation included in the UFSAR. Specifically, staff understand, based on the licensee-provided information, that: 1.The ITS SR 3.5.4:2-specified minimum borated water volume of 500,100 is assured through implantation of a licensee-controlled surveillance procedure which ensures that the BWST minimum indicated level will not be allowed to drop below 39 feet. This level corresponds to an indicated volume of 505,440 gallons and, given instrument loop uncertainty, assures that the ACTUAL AVAILABLE BWST volume will be no less than 488,160 gallons, measured from the top of the 4-inch BWST discharge pipe. 2. The ITS SR 3.5.4.2-specified maximum borated water volume of 550,000 gallons corresponds to the ACTUAL AVAILABLE BWST volume at which spill-over is expected to occur. To assure that this volume is not reached or exceeded, the licensee-controlled surveillance procedure ensures that the BWST maximum indicated level will not be allowed to exceed 41 feet. This level corresponds to an indicated volume of 531,360 gallons and; given instrument loop uncertainty, assures that the ACTUAL AVAILABLE BWST volume will be no greater than 548,640 gallons, measured from the top of the 4-inch BWST discharge pipe. 3. The bottom 3 feet of ACTUAL AVAILABLE BWST volume (38,880 gallons) are not assumed to be available for ECCS injection following a LOCA based on BWST vortex / NPSH analysis constraints associated with drawing boric acid from the BWST. Thus the TS minimum USEABLE BWST volume shall be no less than 449,280 gallons.4. ECCS switchover from injection to recirculation is a manual operation .that occurs in two time-critical intervals:
: a. The first interval is from time of ECCS injection actuation until receipt of an automatic switch-over permissive which occurs when BWST level has diminished to between 8.5 and 9.6 feet (109,728 gallons to 124,632 gallons remaining).
Based on the maximum allowable ECCS injection rate of 14,239 gpm, field operators will have at least 25 minutes to accomplish this task: Minimum operator training and qualification requirements (e.g. JPM's) and procedures assure that this task can be reliably accomplishedin less than 20 minutes. b. The second interval is from receipt of the switch-over permissive until flow control valves manipulated by main control room operators have fully redirected ECCS pump suction from the BWST to the Emergency Containment Sump, thus terminating BWST drawdown.
Based on the maximum allowable ECCS injection rate of 14,239 gpm, main control room operators will have at least 5 minutes to accomplish this task.Operator training and qualification requirements (e.g. Simulator Performance Requirements) and procedures assure that this task can be reliably accomplished in less than 4 minutes. 5. ECCS http://www.excelservices.com/exceldbs/itstrack.davisbesse.nsf/I fddcealOd3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 7 of 16 pump design, flow-control valve limits, and surveillance requirements assure that the maximum allowable ECCS injection rate of 14,239 gpm, upon which the above time-critical intervals are dependent, will not be exceeded.
: 6. Based on ECCS pump NPSH / vortex analyses constraints associated with drawing water from the Emergency Containment Sump, switchover from ECCS injection to recirculation may not occur prior injecting at least 360,000 gallons into the reactor coolant system / containment envelope.
Based on allowable values for the switch-over permissive and BWST TS minimum volume, no less than 363,528 gallons will be injected prior to switchover.
: 7. Licensee-controlled procedures assure, and will continue to assure, that BWST level instrumentation relied upon to meet SR 3.5.4.2 shall be no less accurate than the +/- 16 inches assumed above. 8. The licensee understands that, in specifying an ANALYTICAL (actual) upper BWST volume limit and an instrument uncertainty-corrected lower volume limit in gallons (as opposed to consistently specifying appropriate OPERATIONAL (indicated) limits in the units displayed to the operators by the associated instrumentation, e.g.41 feet and 39 feet, respectively) the BWST volume limits, while adequate, are more difficult to understand and not fully in keeping with the Commission's TS Policy Statement expectation that conversion to ITS improve the safety of nuclear power plants through the use of more operator-oriented TS. 9. Given the time-critical nature of 4.b., above, (the second time-critical event associated with ECCS switchover from injection to recirculation which must be completed in less than 5 minutes, please justify the proposed relaxations associated with the proposed changes to CTS SR 4.5.2.d.2.b)
[Att. 1, Vol. 10, Pg 30 of 98]. The relaxations are discussed in LA04 and LA05 [Pg 35 and 36 of 98] as Type 1 -Removing Procedural Details but appear to effectively constitute removal of these surveillance requirements:The modified requirement is mapped to INSERT 1 [Pg 45 of 98] which creates a proposed new ITS SR 3.5.2.8. -addressed as JFD 4 [Pg 46 of 98].The CTS requires that, within 75 seconds of receiving a BWST Low-Low Level interlock trip (permissive interlock?)
that the BWST Outlet Valves HV-DH7A/B automatically close in less than or equal to 75 seconds after the operator manually pushes the control switch to open the Containment Emergency Sump Valves HV-DH9A/B and that these valves open in less than or equal to 75 seconds. The reviewer's understanding of this SR is that it must be performed with the motor operators for the valves energized but that the ability of the valves, when energized and commanded, to stroke in less than or equal to 75 seconds must be met at all times while in the TS-applicable MODE. A note in the associated ITS SR 3.5.2.8, which the licensee proposes, states that the SR is required to be MET ONLY when the motor operators for the valves are energized (while in the TS-applicable mode). The terms MET and PERFORMED have particular importance when used in the context of SR's. It appears that INSERT 1 has incorrectly http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 8 of 16 applied the term MET in a manner such that, the inability of the valves to stroke, when energized and commanded, in less than or equal to 75 seconds would not constitute an failure to meet the SR as long as the valves were maintained deenergized
-as is the case most of the time. The SR also eliminates the 75-second stroke acceptance criteria from the TS and provides no associated SR 3.5.2.8 Bases discussion.
This proposed change is a Beyond-Scope change because it deviates both from the CTS and the STS as evidenced by concurrent DOCS (LA04 and LA05) and associated JFD4. Because this BSI was not explicitly identified by the licensee, it was not explicitly reviewed by the cognizant technical branch earlier in the ITS conversion.
LA04 states that removal of the 75-second acceptancecriteria and limiting SR applicability constitutes removal of a type of information not necessary to be included in TS to provide adequate protection of public health and safety. It further states that this type of procedural detail will be adequately controlled in the ITS Bases. Please explain how this is possible and indicate where, in the ITS Bases, these types of procedural details are adequately controlled, LA05 states that the stroke timing detail to assure that the BWST Outlet Valves HV-DH7A/B automatically close in less than or equal to 75 seconds and that the Containment Emergency Sump Valves HV-DH9A/B open in the same time-frame are moved to the ITS Program. Please indicate where this is stated in the associated TS and provide that portion of the ITS program that contains this requirement.
JFD 4 is inaccurate/incomplete in that it states ITS SR 3.5.2.8 has been added consistent with the current licensing basis. CTS SR 4.5.2.d.2.b);
which is part of the CLB, requires that these valves stroke within 75 seconds. The reviewer interprets the CTS to require that the valves meet the SR at all times when in the applicable mode; the proposed ITS SR does not. It is the reviewer's understanding that while in the TS-applicable mode: (a) these valves are energized only about 5 minutes before they must be stroked following a LOCA, (b) 75 seconds constitutes a significant 25% of the approximate 5 minutes budgeted for the time-critical switchover to recirculation, (c) failure of the valves to stroke or to stroke within the required time could result in loss of safety system function, (d) there would likely be insufficient time to resolve any identified valve OPERABILITY issues in the 5 minutes following valve energization, and after which they must be stoked. Thus, the valves must be capable of being energized and stroking in less than or equal to 75 seconds at all times in the TS-applicable mode. The NOTE in INSERT 1 could be interpreted differently.
Please explain how the proposed change is consistent with the analysis above, the current licensing basis and, in particular, with 10 CFR 150.36(d)(3).
Licensee Response by Jerry Everything in the reviewer post on 3/21/08 has been confirmed Jones on 04/03/2008 with the exception of the below corrections/clarifications to the specified item numbers: 1. It is true that an indicated level of 39 feet corresponds to 505,440 gallons. It is true that given instrument http ://wwwexcelservicesI.com/exceldbs/itstrack.davisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008-NRC ITS Tracking Page 9 of 16 uncertainty, the actual available BWST volume would be no less than 488,160 gallons at the indicated level of'39 feet. However, a summary of the response, posted on 2/20/08 is provided below to clarify the rest of the item #1: Permissive setpoint = 108.5 inches Level for 360,000 gallons = 333.3 inches Uncertainty for indicator string = 16 inches 108.5 + 333.3 + 16 = 457.8 inches Rounding this for the readability of the indicator gives 462 inches (38.5 feet).Since 462 inches is less than 463.1 inches (500,100 gallons) -a surveillance that ensures at least 500,100 gallons available ensures that the analytical limit is protected.
The surveillance stipulates 39 feet (468 inches) -which is above 463.1 inches and therefore ensures 500,100 gallons available.
Stated in terms of gallons: 39 feet -indicator uncertainty
= (468 -16) = 452 inches = 488,160 gallons Permissive setpoint = 108.5 inches = 117,180 gallons Level for 360,000 gallons = 333.3 inches = 360,000 gallons 488,160 -117,180 -360,000 = 10,980 gallons of margin to the analytical limit 3. The purpose of the. reviewer statement "Thus the TS minimum USEABLE BWST volume shall be no less than.449,280 gallons" is not clear. There is no requirement for a TS minimum USEABLE volume. 4.b. A second time interval is described by the reviewer as "from receipt of the switch-over permissive until flow control valves manipulated by main control room operators have fully redirected ECCS pump suction from the BWST to the Emergency Containment Sump, thus terminating BWST drawdown.
Based on the maximum allowable injection rate, operators will have at least 5 minutes to accomplish thisThe attachment provided with the response on 3/08/08 included information from Calc C-NSA-049.01-004, Vortex Formation with ECCS Pump Suction from the BWST. This calculation stipulates that the transfer from the BWST to the emergency sump be initiated within 2 minutes after the low level permissive is received at 75 inches. 6. Based on thevarious margins in the calculation for minimum required BWST level, there will be more than 360,000 gallons injected into containment before completion of the manual suction transfer.
As described in the attachment to the 3/6/08 response, the calculation starts with the minimum required level to receive the permissive (75 inches), and applies various uncertainties and margins to arrive at the minimum required level of 500,051 gallons (rounded to 500,100 gallons for the Techp I ec value). The reviewer states "Based on allowable values for the switch-over permissive and BWST TS minimum volume, no less than 363,528 gallons will be injected prior to switchover." It is not believed that knowing this precise volume amount is important.
However, to confirm or correct the value of 363,528 gallons, the bases behind the number must be understood.
: 7. Continued assurance of the instrument uncertainty.
assumptions are provided by the fact that procedure and calculation changes are governed under the 10 CFR 50.59 process.8. The reviewer statements in item #8 are understood.
However, changing the BWST Tech Spec values for volume would be a http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/1 fddceal 0d3bdbb585256e....
7/17/2008 NRC ITS Tracking Page 10 of 16 Beyond Scope Issue (BSI). The Davis-Besse goal has been and continues to be to minimize the number of BSI's for the ITS Conversion.
: 9. As stated in the Description section of ITS 1.4 (Volume 3,Page 53), a Surveillance is "met" only when the acceptance criteria are satisfied.
Known failure of the requirements of a Surveillance, even without a Surveillance specifically being"performed," constitutes a Surveillance not "met." When the valves are de-energized, they cannot meet the requirements to open or close, as applicable, in less than 75 seconds. Thus, the term "met" in the Note to ITS SR 3.5.2.8 (Volume 10, Page 45) is the correct term to use, not "performed," which implies that the valves have to open or close within the limits even between performances of the SR (i.e., when the valves are de-energized).
The removal of the time limits to close or open the valves, 75 seconds, is consistent with the ISTS. This removal is justified in Discussion of Change (DOC) LA05 (Page 35) only (the NRC reviewer implies that the removal is part of DOC LA04 also -it is not). DOC LA05 states that the specific time limit is being relocated to the IST Program (not the ITS Bases). The ISTS does not include stroke times for various components in the ISTS. For example, the valve actuation SR, ISTS SR 3.5.2.5, and the pump start SR, ISTS SR 3.5.2.6, requires the valves to actuate to the correct position and the pump to start on a actuation signal and no valve stroke times or pump start times are provided.
Another example is ISTS SR 3.6.6.5, which verifies that the containment spray system valves actuate to the correct position on an actuation signal, and no time for the valves is provided.
ITS 5.5.7 provides the requirements for the Inservice Testing Program, and these valves are currently tested in this program. The specific time, 75 seconds, is the acceptance criteria in the procedures that implement the IST Program for these valves. Therefore, Davis-Besse believes that we are consistent with the intent of the ISTS (i.e., we are attempting to be consistent with the ISTS), and that the times for valves of this nature are allowed to be in the IST Program. However, based upon the NRC reviewer's concern, Davis-Besse reviewed other ISTS SRs to see how isolation times were covered in other SRs. It was noted in ISTS SR 3.7.2.1 (Volume 11, Page 41) for MSIVs and ISTS SR 3.7.3.1 (Page 67)for MFSVs, that these SRs required the isolation time to be verified to be "within limits." Therefore, Davis-Besse will modify ITS SR 3.5.2.8 to verify the valves actuate to the correct position"within the time limit" on a manual actuation signal. A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
In addition, JFD 4 states that ITS SR 3.5.2.8 is consistent with the current licensing basis. It does not state that it is identical to the words used in the CTS. The wording in the ITS differs from the CTS wording, but only where necessary to be consistent with the format and details of what should be in the ITS, as described above (e.g., specific time limits are part of the IST http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/
lfddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page I I of 16 Program and are located in that document).
Furthermore, the actual type of signal that is used to alert the operators of the need to manually realign the BWST valves is also not normally included in the ISTS SRs. For example, as stated earlier, the valves actuation SR, ISTS SR 3.5.2.5, and the pump start SR, ISTS SR 3.5.2.6, do not list the specific signal to actuates the valves or start the ECCS pumps. The types of start signals are described in the ISTS Bases (in the Background section and in the individual SR).Licensee Response by Jerry Jones on 04/26/2008 This response modifies the 4/3/2008 response for question 9.Based on a phone conversation with the NRC reviewer subsequent to posting the 4/3/2008 response to question 9, Davis-Besse understands that the NRC reviewer believes that the valve stroke time requirement of SR 3.5.2.8 (Volume 10, Page 45 of the draft markup provided in the 4/3/2008 response) should not have the Note applied to it. Davis-Besse concurs with the NRC reviewer and is providing a new draft markup to the SR. The proposed draft markup regarding this change is attached, and supersedes the previous draft markup provided in the 4/3/2008 response.
The draft markup includes a requirement to verify the valve actuation time, and moves the actual time to the TRM in lieu of the IST Program, since the valve actuation time is analogous to the ECCS Response Time requirements for other ECCS components (and the ECCS Response Times are currently located in the TRM). This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
During the phone conversion described above, the NRC reviewer also requested information concerning the purpose of the SR -i.e., is it related to ECCS Response Time? The SR is related to ECCS Response Time testing. The original design of Davis-Besse included an automatic swapover on BWST-Low Low Level However, this automatic swapover was changed to a manual swapover, as approved in License Amendment 36, dated January 24, 1981. License Amendment 40, dated June 1, 1981, approved additional changes with respect to the manual swapover:
(1) Surveillances were added to verify the interlock between the BWST outlet valves and Containment Emergency Sump valves, and to verify the valve stroke time; and (2) The SFAS Response Time requirements for the BWST Low Low Level Function were deleted, since the valve Surveillances described above included a stroke time. At the time of License Amendment 40, the SFAS Response Time requirements were in CTS 3.3.5, Table 3.3-5. These SFAS Response Time values were removed from the Davis-Besse CTS and placed under Davis-Besse control as documented in the NRC Safety Evaluation Report for License Amendment 225, dated July 7, 1998. The valve interlock Surveillance described above is maintained as ITS SR 3.5.2.8. Davis-Besse has reviewed the ITS SR 3.5.2.8 Bases description and believes clarification is needed concerning the BWST Level -Low Low signal, since the signal only provides a permissive to allow manual transfer of the ECCS suctions.
The proposed draft markup includes these changes and will also be http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/lfddcea1Od3bdbb585256e..:
7/17/2008 NRC ITS Tracking Page 12 of 16 reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Ross Telson on 05/07/2008 Ref: Question 9, posted by staff on 3/31/08 and responded to by the licensee on 4/3 and 4/26, and discussed in a teleconference on 4/30. The reviewer examined and discussed with technical staff the reasoning behind license amendments 36, 40, and 225, and Generic Letter 93-08, in the context of the B&W STS model and Davis-Besse's site-specific design for ECCS switch-over.
Based on the above, portions of D-B's proposed revisions to CTS SR 4.5.2.d.2.b.
appear to be beyond the scope of the ITS conversion and inconsistent with the D-B current licensing basis, as documented in license amendments 36 and 40 and associated Safety Evaluations.
In license amendment 36, the licensee proposed to permit manual switchover (as opposed to an automatic switchover) of ECCS pumps from the BWST to the Emergency Sump during a loss of coolant accident (LOCA). Staff approved the modification as an interim solution to address concerns following a December 5, 1980 event during which an inadvertent actuation of the SFSAS occurred with an automatic alignment of the ECCS pumps to a dry sump. The SE documented the licensee's commitment to incorporating the BWST and Sump valve actions into the CTS. Staff required that sufficient time be allowed for operator action in the analyzed sequence of events. Based on conservative assumptions at the time, including (but not limited to)the assignment of a dedicated operator, maintenance of revised minimum BWST levels, and valve stroke times no greater than 90 seconds, the operator would have about 4 1/2 minutes to complete the switchover procedures from the time he is alerted to commence.
Staff concluded that this was acceptable onan interim basis because it provided greater than the three minutes required for two operator actions necessary to transfer one train of the ECCS. License amendment 40 approved CTS SR 4.5.2.d.2.b): "Each ECCS subsystem shall be demonstrated OPERABLE at least once per 18 months by verifying that on a Borated Water Storage Tank (BWST) Low-Low Level interlock trip, the BWST Outlet Valve HV-DH7A (HV-DH7B) automatically close in ! 75 seconds after the operator manually pushes the control switch to open the Containment Emergency Sump Valve HV-DH9A (HV-DH9B) which should be verified to open in S 75 seconds." The SE stated: "TECo's proposed TSs would impose an 18-month surveillance requirement to verify that each BWST outlet valve moves to its closed position when the operator opens the respective sump valve. All valves are required to complete their movements in less than 75 seconds. As a result of the inclusion of the response time in this new requirement, TECo has proposed that the current response time requirements of Table 3.3-5 be deleted." CTS SR 4.5.2.d.2.b) was subsequently revised to: "Each ECCS subsystem shall be demonstrated OPERABLE at least once each REFUELING INTERVAL by verifying that on a Borated Water Storage Tank (BWST) Low-Low Level interlock trip, with the http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/lfddcealOd3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 13 of 16 motor operators for the BWST outlet isolation valves and the containment emergency sump recirculation valves energized, the BWST Outlet Valve HV-DH7A (HV-DH7B) automatically close in  75 seconds after the operator manually pushes the control switch to open the Containment Emergency Sump Valve HV-DH9A (HV-DH9B) which should be verified to open in -<75 seconds." The reviewer finds that CTS SR 4.5.2.d.2.b) is a plant-specific SR that was previously determined necessary to asssure adequate surveillance requirements following plant-specific modifications.
It need not be substantially modified, nor should its 75-second acceptance criteria be relocated from TS, as part of the ITS conversion scope. The ITS Bases should, however, be appropriately modified to include an appropriate summary statement of the bases or reasons for the SR, in accordance with 10 CFR 50.36(a).Licensee Response by Jerry Jones on 05/19/2008 Davis-Besse notes that the NRC reviewer quoted the NRC Safety Evaluation for Amendment 40, and that one of the quotes states that the new 75 second requirement is a response time and is replacing the ECCS RESPONSE TIME in CTS Table 3.3-5.Subsequent to this amendment, the ECCS RESPONSE TIME Table was allowed by the NRC to be relocated to the Technical Requirements Manual. Thus, Davis-Besse continues to believe that the 75 second valve stroke requirement in CTS 4.5.2.d.2.b is analogous to the ECCS Response Time -i.e., if it were automatic, the time for the valve stroke would not be in the CTS. However, Davis-Besse will agree to not remove the 75 seconds at this time, and will add it back into the ITS submittal.
A draft markup regarding this change is attached and supersedes the previous draft markup attached to the response on 4/26/08. This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Ross Telson on 05/20/2008"rhe reviewer considers the proposed change to CTS 4.5.d.2.b., as reflected by the marked-up ITS SR 3.5.2.8.a NOTE, to constitute an inappropriate relaxation in the applicability of SR 3.5.2.8.a.
Specifically:
CTS 4.5.d.2.b.
Requires:
Each ECCS subsystem to be demonstrated OPERABLE at least once per REFUELING INTERVAL by "Verifying that on a Borated Water Storage Tank (BWST) Low-Low Level interlock trip, with the motor operators for the BWST outlet isolation valves and the containment emergency sump recirculation valves energized, the BWST Outlet Valve HV-DH7A (HV-DH7B) automatically close in -- 75 seconds after the operator manually pushes the control switch to open the Containment Emergency Sump Valve HV-DH9A (HV-DH-9B) which should be verified to open in --< 75 seconds. If at any time a performer of this SR (or a subsequent reviewer)determines any of the valves did not (or could not be relied upon, under conditions specified in CTS 4.5.d.2.b., to) stroke to the correct position in -. 75 seconds, the above SR would require the licensee to declare the SR AND the LCO NOT MET -regardless http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 14 of 16 of whether or not the valves had been subsequently de-energized.
This is the current licensed (and NRC-accepted) condition of operation.
The proposed-ITS SR 3.5.2.8 Requires:
Two ECCS trains to be demonstrated OPERABLE once per 24 months by verifying: "a) each BWST outlet valve and containment emergency sump valve actuate to the correct position on a manual actuation signal; and.b) the actuation time of each BWST outlet valve and containment emergency sump valve is within the limit." The above is modified by NOTE: "SR 3.5.2.8.a is only required to be met when the motor operators for the borated water storage tank (BWST) outlet valves and containment emergency sump valves are energized." Unlike CTS 4.5.d.2.b., should a performer of this SR (or a subsequent reviewer) determine that "each BWST outlet valve and containment emergency sump valve" did not or could not be relied upon to "actuate to the correct position on a manual actuation signal" AFTER "the motor operators for the borated water storage tank (BWST) outlet valves and containment emergency sump valves are" no longer "energized," the proposed ITS would require the licensee to conclude SR 3.5.2.8.a NOT MET but would NOTrequire the licensee to determine the LCO NOT MET. 1. Consider retaining the content of CTS 4.5.d.2.b without the ITS 3.5.2.8.a.
NOTE or with a NOTE more like the content of CTS 4.5.d.2.b.
: 2. Also consider amending the SR 3.5.2.8 Bases to include, in accordance with 10 CFR 50.36, "a summary statement of the bases or reasons for such specifications" (a) as discussed in LARs 36 and 40 and the associated Safety Evaluations that established the SR, and (b) as described in 10 CFR Part 50 "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," 58 FR 39132 (pp 39132-39139), Published 7/22/93. E.g. (excerpts follow): Clarification of the scope and purpose of Technical Specifications has provided useful guidance to both the NRC and industry and has served as an important incentive for industry participation in a voluntary program to improve Technical Specifications.
Each ... Surveillance Requirement should have supporting Bases. The Bases should at a minimum address the following questions and cite references to appropriate licensing documentation (e.g., FSAR, Topical Report) to support the Bases.What are the Bases for each Surveillance Requirement and Surveillance Frequency; i.e., what specific functional requirement is the surveillance designed to verify? Why is this surveillance necessary at the specified frequency to assure that the system or component function is maintained, that facility operationwill be within the Safety Limits, and that the LCO will be met? Note: In answering these questions the Bases for each number (e.g., Allowable-Value, Response Time, Completion Time, and Surveillance Frequency), state, condition, and definition (e.g., operability) should be clearly specified.
As an example, a number mighlt be based on engineering judgment, past experience, or PSA insights; but this should be clearly stated.FI http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 15 of 16 NRC Response by Ross Telson on 06/01/2008 During informal dialogue, the licensee inquired into whether the staff would consider CTS LCO 3.5.2.d to be MET with the valve motors de-energized.
Reviewer response:
CTS LCO 3.5.2.d is MET when those SSCs required to meet the LCO are OPERABLE AND all applicable SR are MET. In this discussion, the applicable CTS SR is 4.5.2.d.2.b).
It requires that, on a Borated Water Storage Tank (BWST) Low-Low Level interlock trip, with the motor operators for the BWST outlet isolation valves and the containment emergency sump recirculation valves energized, the BWST Outlet Valve HV-DH7A (HV-DH7B) automatically close in --< 75 seconds after the operator manually pushes the control switch to open the Containment Emergency Sump Valve HV-DH9A (HV-DH-9B) which should be verified to open in -75 seconds. To satisfy the above, the valves must be capable of being energized through closure of OPERABLE supply circuit breakers.The reviewer considers this capability to be implicit in CTS SR 4.5.2.d.2.b).
If this capability is NOT present, then SR 4.5.2.d.2.b) and LCO 3.5.2.d are NOT MET; however, the act of opening the supply circuit breakers to satisfy Appendix R does not, in and of itself, prevent CTS SR 4.5.2.d.2.b) and LCO 3.5.2.d from being MET. This interpretation differs slightly from that proposed by the licensee and contributes to the staff's concern regarding the proposed ITS SR applicability NOTE. One acceptable resolution might be to an applicability-NOTE-free ITS SR 3.5.2.8 and BASES that appropriately convey the above. I. E.: ITS SR 3.5.2.8: Verify: a) that the motor operators for the valves listed in b), below, are capable of being energized through closure of their respective OPERABLE supply circuit breakers; and b) that on a Borated Water Storage Tank (BWST) Low-Low Level interlock trip, with the motor operators for the BWST outlet isolation valves and the containment emergency sump recirculation valves energized, the BWST Outlet Valve HV-DH7A (HV-DH7B)automatically close in --< 75 seconds after the operator manually pushes the control switch to open the Containment Emergency Sump Valve HV-DH9A (FIV-DH-9B) which should be verified to open in :-< 75 seconds.Licensee Response by Jerry Jones on 06/02/2008 Davis-Besse has reviewed the newest response by the NRC reviewer dated 6/1/08 and believes that the following proposal might resolve the issue. 1) The Note to SR 3.5.2.8 in the last draft markup will be deleted. 2) A new Note will be added to the LCO statement stating "The BWST outlet and containment emergency sump valves may be considered OPERABLE when the associated valve motors are de-energized, provided the valves are not otherwise inoperable." 3) This'new Note will be added for LCO 3.5.3 also. Davis-Besse notes that NUREG-1430 contains a Note (Note 1) for LCO 3.5.3, which is a similar approach as being suggested above. If this approach is acceptable to the NRC reviewer, a draft markup of the specific changes will be provided so that the specifics of the change can be reviewed.II http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsfl1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 16 of .16 Licensee Response by Jerry Jones on 06/05/2008 Davis-Besse is posting a new proposed markup based on our response of 6/2/08. The proposed markup includes the following:
: 1) The Note to SR 3.5.2.8 in the last draft markup will be deleted.2) A new Note will be added to the LCO statement stating "The-BWST outlet and containment emergency sump valves may be considered OPERABLE when the, associated valve motors are de-energized, provided the valves are not otherwise inoperable." 3)This new Note will be added for LCO 3.5.3 also. Davis-Besse worded the proposed Note as described above based on the precedence set in NUREGs-1433 and -1434, ISTS 3.5.1. The LCO statements for NUREGs-1433 and -1434 ISTS 3.5.1,which require each ECCS subsystem to be OPERABLE, include a Note that allow's the low pressure coolant injection (LPCI) subsystems to be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal cut in permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.
This Note is needed since another Specification (ISTS 3.4.8 for NUREG-1433 and ISTS 3.4.9 for NUREG-1434) requires the RHR System to be in operation, and doing this would render the LPCI subsystems inoperable.
Thus, the NRC allows the LPCI subsystems, to be considered OPERABLE when in this condition.
Davis-Besse believes that this approach best matches the CTS.allowance and ITS format. A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
This draft markup supersedes the previous markups for the ITS 3.5.2 issues in this RAI.Licensee Response by Jerry Jones on 07/02/2008 During a recent phone conversation with the NRC, the NRC reviewer requested minor changes to the draft markup provided in the 6/5/08 Davis-Besse response.
A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
This draft markup supersedes the 6/5/08 markup for the ITS 3.5.2 issues in this RAI.Licensee Response by Jerry During a recent phone conversation with the NRC, the NRC Jones on 07/02/2008 reviewer requested changes to the draft markup provided in the.previous response.
A draft markup regarding this change is attached and supersedes the previous draft markup.NRC Response by Ross Telson The item is closed out for Ross Telson by Carl Schulten on 07/02/2008_
Date Created: 01/02/2008 04:33 PM by Ross Telson Last Modified:
07/02/2008 02:17 PM http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1fddcealOd3bdbb585256e...
7/17/2008 Item #1 -the 'below information was obtained from the section of Calculation C-ICE-48.01-004 that determined appropriate surveillance limits to place in the applicable procedure.
(1) 500,100 gallons = 463.1 "WC (1080 gal/inch)(2) Existing setpoint = 108.5 "WC (3) Level required for 360,000 gallons = 333.3 "WC (1080 gal/inch)(4) Uncertainty for indicator string = 16.0 "WC Margin exists if 2 + 3 + 4 is less than 1.108.5 + 333.3 + 16.0 = 457.8 "WC The readability of the indicator is k of the smallest division, and the smallest division is one foot. Therefore, the readability is 6 inches.Rounding 457.8 to the next highest 6 inch value~gives 462" (38.5 feet).Since 462" is less than 463.1", a surveillance that ensures at least 500,100 gallons is available in the tank ensures the analytical limit is protected.
Since the surveillance stipulates 39 feet (468"), the surveillance ensures that at least 500,100 gallons is available.
Item #2 -the below information was obtained from the section of Calculation C-ICE-48.01-004 that determined appropriate surveillance limits to place in the applicable procedure.
The 550,000 maximum volume level is to preclude overflow of the tank, to ensure that the amount of tri-sodium phosphate is adequate for the maximum amount of water that can be injected, and for flooding level in containment.
550,000 gallons = 509.3 "WC .(1080 gal/inch)Uncertainty for indicator string = 16.0 "WC 509.3 -16.0 = 493.3 "WC Rounding down to the next 6" level for readability results in 492 "WC (41.0 ft) -which is equivalent to 531,360 gallons.Note also for both Item #1 and #2 that annunciators are provided for both low and high BWST level. The setpoints for these annunciators operationally restrict level even tighter within the specified surveillance procedure limits.
Item #3 -The statement "there is never 550,000 gallons available and/or injected into the core/containment" is a correct statement.
Only a summary of the CR 03-02644 evaluation was provided in the response.
Based on the ECCS pump suction switchover from the BWST to the Emergency Sump, 75 inches of BWST level is not injected.
75 inches corresponds.
to 81,000 gallons.Item #4 -A clarification is requested for this item. It was believed that the new LA01 provided the justification for relocating the term "available" to the Bases. JFD #9 merely explains the relocation that was documented in LA01. Would sufficient justification be to add the following statement at the end of JFD #9: "since the term available was relocated to the Bases as described in Discussion of Change LA01." Or, is the concept.of relocating the term available to the Bases being challenged?
For background, the analyses are based on available BWST volume, which is the volume in the BWST above the 4 inch discharge penetration into the bottom of the tank. The BWST level transmitters are calibrated such that they read zero when water level is at the top of the 4 inch penetration.
In other words, they are, calibrated to indicate available volume. This means that the sump switchover setpoint in ITS 3.3.5, Parameter 5, is also based on available volume. Changing the current BWST volume limits to reflect total volume would be a beyond scope change- which we are trying to avoid. Simply changing the BWST level limits to reflect total volume would likely lead to confusion, after considering the preceding information.
If relocating the term available is not acceptable, another option would be to maintain the term available in SR 3.5.4.2. This would maintain the current license basis.
: 1. Calc C-NSA-049.01-004 is titled Vortex Formation with ECCS Pump Suction from the BWST. This calculation is 639 pages long. Some key points from the analysis:-The analytical limit for the BWST-to-Emergency Sump transfer permissive is 75 inches.-3 feet in the BWST is determined to be the worst case level. 3 feet in the BWST is used as the input for Calc C-NSA-049.02-048
-titled LPI, CS and HPI Pump NPSH with suction from the BWST.2. The results of Calc C-NSA-049.02-048 are an input to UFSAR 6.3.2.14.UFSAR Pages 6.3-11, 6.3-12 and 6.3-13 seem to have all the information of interest concerning NPSH.3. In the response to question 200801021633 posted on 2/11/08, a summary of the bases for the 500,100 gallon Tech Spec minimum limit was provided.
Calc C-ICE-48.01-004 determined the setpoint and allowable values for the transfer permissive, and also documented the bases for the 500,100 gallon Tech Spec limit. A restatement of the applicable portions of the previous response is given below:-Starts with the analytical limit for the transfer permissive (75 inches).-Adjustment is made for Operator response time to initiate the ECCS pump suction transfer from the BWST to the CTMT emergency sump (12 inches of additional margin)-Adjustment is made for instrument uncertainty to provide the permissive setpoint (108.5 inches) and allowable values. This permissive setpoint is established to provide protection against vortexing if the transfer is started too late due to instrument error. The Allowable Values on either side of the trip setpoint are 101.6 and 115.4 inches, and are provided in ITS 3.3.5, Parameter 5.-Adjustment is made for the minimum volume assumed in the analyses of 360,000 gallons injected into the core/containment.
-Adjustment is made for instrument uncertainty.
This additional uncertainty is included to ensure the transfer is not started too early due to instrument error, and thus not meeting the 360,000 gallon requirement.
-The result of these adjustments is calculated to be a minimum value of 500,051 gallons. As stated in the CTS Bases, the Tech Spec limit of 500,100 gallons was conservatively rounded up from the calculated value.
Selected Items from Calc C-NSA-049.01-004, Vortex Formation with ECCS Pump Suction from the BWST OBJECTIVE OR PURPOSE: This calculation determines a Borated Water Storage Tank (BWST) level that provides an acceptable analytical limit for the BWST-to-Emergency Sump transfer permissive.
In order to ensure an acceptable analytical limit, the following are evaluated:
: 1) Potential for air ingestion into the Low Pressure Injection (LPI), and Containment Spray (CS) pumps during the transfer from the BWST to the Emergency Sump. Both large-break and small-break LOCA scenarios are evaluated.
: 2) Acceptable pump NPSH requirements are maintained at all times.SCOPE OF CALCULATION/REVISION:
Revision 2 incorporates:
(1) a PROTO-FLO analysis that considers the flowpath between the BWST and Emergency Sump, (2) evaluation of terminating HPI flow before the BWST transfer permissive, (3) removal of the 4 inch pipe extension included in the Froude number calculation, (4) the reduction of the air volume in a fluid due to a decrease in elevation and, (5)numerous format changes have also been incorporated.
Due to the incorporation of several technical changes and a large number of format changes, Revision 2 is considered a complete update and no change-bars are provided.Revision 2 incorporates PIN 1 issued against Revision 1.Revision I of this calculation has been design verified.DIN 1, DIN 2, DIN 6, DIN 33, DIN 36 and DIN 42 have been designed verified.
 
==SUMMARY==
OF RESULTSICONCLUSIONS:
The acceptance criteria are satisfied for an initial BWST level of 75 inches. Therefore, an analytical limit of 75 inches for the BWST-to-Emergency Sump transfer permissive is acceptable.
IMPACT ON OUTPUT DOCUMENTS:
DIN 6: The BWST low level analytical limit is used by DIN 6 to determine an allowable value.The analytical limit has not been changed by Revision 2. Therefore, no changes are needed to DIN 6 as a result of Revision 2.DIN 17: This DIN discusses that the transfer from the BWST to the Emergency Sump is initiated at a BWST level of 9 feet. This value is not being changed by Revision 2. Therefore, no changes are needed to DIN 17 as a result of Revision 2.DIN 43: A minimum BWST level of 3 feet (during the swapover from the BWST to the Emergency Sump) is used by DIN 43. The analyses performed by Revision 2 do not change the minimum BWST level. Therefore, no changes are needed to DIN 43 as a result of Revision 2.DIN 44: This DIN lists time critical operator actions. No changes to operator aclions are required as a result of Revision 2. t Introduction At the onset of all accidents, the Borated Water Storage Tank (BWST) would begin at a. level of over 35 feet (500,100 gallons, DIN 20). At this level, there is no potential for vortex formation.
However, as the BWST level decreases, the potential for vortex formation and air entrainment'in the suction piping of the ECCS pumps could occur if the transfer to the Containment Vessel Emergency Sump is performed at a sufficiently low level and the fluid velocity is sufficiently high. Once the transfer to the Emergency Sump is completed, the BWST is no longer used as a suction supply.With air entrainment in the range of 0 to 15 percent air by volume, the pump performance will fall below the head flow curve which would be expected when pumping a single phase liquid of the same average density. Information on air entrainment in the range of 0 to 15 percent has been compiled in DINs 4 and 13 for pumps with specific speeds of approximately 800 to 1600.At the low end of this range (below 4 percent).
the pump performance tends to be as predicted by the head flow curve any time the flow rate is at least half the flow rate at the pump's best efficiency point. At the upper end of the range (approximately 15 percent), the pump head/flow curve may be essentially fully degraded.
Performance of the pump with increasing air entrainment produces a curve very similar to curves which are based on reductions in suction head below the NPSH requirement, where cavitation is increasing.
Normally, within this range, degradation in the pump head flow curve is temporary, and performance will immediately return to the normal head-flow curve as soon as the air entrainment is sufficiently reduced. However, small amounts of air entrainment also have been noted to increase NPSH requirements; e.g. air ingestion of 2 percent will approximately double the required NPSH requirement.
Therefore, long-term operation of a pump with significant air entrainment (over approximately 2 percent) is not generally desired-For long-term pump operation with marginal NPSH available, even minor air ingestion can be detrimental to pump operation since erosion of the impeller and fatigue of components due to vibration can occur. Erosion processes lead to long-term permanent damage to the pump internals.
Long-term operation, up to a year or longer, might be required in some post-accident scenarios with suction being provided from the Emergency Sump. Therefore, DIN 4 and 13 are entirely focused on describing the air entrainment effects with respect to long-term suction from the Emergency Sump. In the case where suction is provided from the BWST, greater NPSH is available.
More importantly, the operational period when air entrainment can occur is well defined andbrief; no longer than a few minutes. By the time the BWST levels are sufficiently low that air entrainment could potentially occur, the flow requirement for any of the pumps (LPI, HPI, or CS) is low. Therefore, minor temporary degradation from the nominal head-flow curves, as detailed in DINs 4 and 13 for the CS and LPI pumps, would not be of concern.
When a pump is provfing flow at near the best efficiency point, it is unlikely to become air bound for Small amounts of air entrainment.
When flow is reduced, internal recirculation current may force 1h1 air cavities to accumulate near the center of The impeller, resulting in air binding of the pump, In the absence of specific testing, DIN 13 Indicates that approximately 20 to 50 percent of the best efficiency flow rate should be maintained if air entrainment is present. However, since air entrainment is less likely to occur oI lower flow rates, air binding is not expected, Therefore, while air entrainment should be prevented or minimized, there is no reason for concern over sufficiently limited, temporary degradation from the nominra head-flow cwve, From DIN 5, small amounts of air ingestion, i.e., 2 percent or less, will o ea to severe pump degradation re uired NP5H from the pump manfac ,I s Curves i inesed based on the cal ulated, ar Ing lion. in this .aton, the air i n and NPSr i reuirements wIl be evaluated.
The Revision of this calculation used 75 inches as anr an.-ytical limit, because the nominal transfer permlsse vel was 96 inches, and after adverse instrument error and drift, the initial level of 75 inches was :sed for conseratism (DIN 8). Prace.durally (DIN 7), the transfer perrniss e Ic -h s been raised to 108 inches (as a'result of DIN 23), in this calculation, 75 es IA contie t o tv used because t is more conservative and the additional 12 Inches is ao~unted for as margin in DIN 6.Depending on the BWST level prior to the swapover and the valve positions (i.e., DH9NB and DHT7AB} during the swapover, flow from the BWST to the Emergency Sump could occur during the swapover du& to gravity. In order to determine the worst-case combinatin of BWST level and vylve positions, a sensitivity study will be performed with diffrent initial BWST levels and vIlve position combinations during the transfer, Design Inputs Pump Flowrates Pump flowrates will be based on PROTO-FLO predictions.
The following flowrates will be used as a guide to ensure the PROTO-FLO predicted flowrates are reasonable.
Maximum LPI Flow: 2*4168 gpm = 8336 gpm (max allowable flow per DIN 7, large-break LOCA, 4000 gpm, with 168 gpm uncertainty from DIN 1)Maximum CS flow: 2070+1.933 gpm = 403 gpm (DIN 2)Maximum HPI flow: 2* 950 gpm =1900 gpm (Pump runout, DIN 24)BWST Level to Volume Conversion The BWST contains 1080 gallonslinch (DIN 6).
Operator Actions (1) Timing of HPI Pump Termination and Initiation of Transfer Based on DIN 26, operator action to terminate HPI pump operation during a large-break LOCA is assumed to be within one minute following the receipt of the BWST transfer permissive.
Initiation of the transfer to the Emergency Sump is assumed additional one minute later. These times are specified by DIN 44, In April 1997, a simulator exercise was conducted in order to determine Operator response times for making this transfer on a LBLOCA. The crew had no knowledge to the purpose of this exercise.
The results indicated that the H-PI pumps were shutdown within the first minute and the transfer was completed in 2.1 minutes (included in Attachment 2).This response time was compatible with the requirements of this calculation.
Results If the transfer from the BWST to the Emergency Sump is initiated within two minutes after the low level permissive is received at 75 inches, air ingestion of 2 percent is the maximum expected.
Additional air will not be entrained due to placement of the BWST and HPI recirculation penetrations, since they are not located directly over the BWST outlet. NPSH requirements of the pumps will be satisfied with 2 percent entrainment.
For LBLOCA, at the anticipated level of air entrainment, pump head will not be substantially degraded and essentially full flow will be maintained.
Since flow rates are very high, air binding will not occur. For SBLOCA, loss of subcooling margin would have compelled the operator to place HPI in piggyback mode, maintaining significant LPI flow, per DIN 7. If CS pumps had started, the LOCA should be sufficiently large to demand reasonably high HPI flow. Since HPI is supplied by LPI when piggybacked, the LPI pump flow would be well above minimum recirculation flow and should be sufficient to avoid air binding. Under these SBLOCA conditions, where only the HPI pumps and CS pumps are injecting, air entrainment is expected to be nearly zero.The guidance given in DINs 4 and 13 is based on long-term
(> 1 year) protection for pumps taking suction from the Emergency Sump. Suction for the BWST at levels which could result in some degree of air induction will occur for only a few minutes. Early in plant life, a condition briefly existed where both sources of water could be simultaneously valved-out at the time of pump start. Until such time the valves would open, severe cavitation might occur. In DINs 15 and 16, the CS and LPI pump vendors evaluated the effects of this condition.
Essentially, both pumps are sufficiently rugged that failure would not occur even with extreme cavitation for a short period of time. Likewise, a small amount of air injection for a short period would not be expected to result in any permanent damage.Conclusions The acceptance criteria are satisfied for an initial BWST level of 75 inches, Therefore, an analytical limit of 75 inches for the BWST-to-Emergency Sump transfer permissive is acceptable.)
NRC ITS Tracking Pagel1 of 3 Return to View Menu, Print Do R.ATI Screening Required:
Yes Status: Closed This Document will be approved by: Gerald Regulatory Basis must be included in Comments Waig; Tim Kobetz; Carl Schulten section of this Form This document has been reviewed and Yes information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING NRC Reviewer ID 200801151040 Conference Call Requested?
Yes Category Other Technical Challenge ITS Section: TB POC: JFD !NMum-ber:
Page Number(s).
ITS 3.5 Ross Telson None 43 Information
[TS..Number;:
OSI:. DOC Number: Bases. J.FD.Number;.
None None None None REF: Question ID 200801021633
----- ACTIONS NEEDED: 1. Regarding SR 3.5.2: a. Confirm that CR 02-04514, addressing the establishment of adequate TS margin for HPI and LPI flow, has been resolved.b. Briefly summarize how adequate margin was/is established between the licensee-controlled surveillance procedure, the TS SR acceptance criteria, and the analyses and evaluation included in the safety analysis report, such that (i)meeting the associated surveillance procedure requirements assures the TS SR acceptance criteria (AC) are met and (ii) meeting the TS SR AC assures the necessary HPI and LPI flow qualities are maintained, that HPI and LPI operation will be within safety limits, and that the limiting condition for operation of the ECCS will be met.c. Provide CR 02-04514.d. Provide any additional calculations, assumptions, analyses, or references you deem necessary or appropriate to support a. and b. above.----- BASIS FOR REQUEST: 1. CR 02-06407, provided by licensee, documented an adverse condition regarding instrument uncertainty for Non-LSSS applications.
The CR was deemed by the licensee to be sufficiently significant to warrant a root cause determination.
Corrective Action 6 concluded that adequate TS margin could not be demonstrated for LPI and HPI flow without detailed calculations or engineering evaluations to quantify the applicable instrument uncertainty and establish appropriate surveillance guidelines.
CR 02-04514 was referenced by CR 02-06407 in regard to resolving HPI and LPI flow TS margin but was not http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 3 provided along with CR 02-06407.Eight (8) other plant parameters were similarly identified in CR 02-06407.Three (3) of the eight (8) are directly related to ITS Section 3.5 SR AC. These are under review in questions 200710032123, 200711161716, and 200801021633.
While TS SR acceptance criteria (AC) that are not changed during ITS conversion might normally receive only cursory review, in this instance, licensee-provided information has raised unresolved questions regarding the adequacy of these TS SR AC.It has,'thus, become necessary to revisit the adequacy of these TS SR AC in accordance with &sect;50.92, which directs that (staff) be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate, and &sect; 50.36, which requires (a) that TS be derived from the analyses and evaluation included in the safety analysis report, and (b)that SR's assure the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.---- REGULATORY REQUIREMENT:
&sect; 50.36 Technical Specifications (a) Each applicant for a license ... shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications
... shall also be included in the application, but shall not become part of the technical specifications.
Comment (b) ... The technical specifications will be derived from the analyses and...........
evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to &sect; 50.34. The Commission may include such additional technical specifications as the Commission finds appropriate.(c) Technical specifications will include items in the following categories:
(2) Limiting conditions for operation.(i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria: (C) Criterion
: 3. A structure, system, or component that is part of the primary'success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge:
to the integrity of a fission product barrier.(3) Surveillance requirements.
Surveillance requirements are requirements to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.&sect; 50.46 Acceptance Criteria For ECCS Requires, in part, that uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated.
This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of this section, there is a high level of probability that the criteria would not be exceeded.http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1fddcealOd3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 3 of 3&sect; 50.92 Issuance of amendment.(a) In determining whether an amendment to a license ...will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses or construction permits to the extent applicable and appropriate.
Issue Date [01/15/2008 Close Datel 03/21/2008 Logged in User: Anonymous vResponses Licensee Response by Bill See attached file Bentley on 01/24/2008 NRC Response by Ross Telson Thank you for your response.
Regarding'item 1, you are correct..on 01/25/2008 Inquiry was directed at SR 3.5.2.2. Staff understand the following from your response and attached documentation.
1.b.(i) meeting the associated surveillance procedure requirements (IST Acceptance Criteria) assures the TS SR Acceptance Criteria (AC)are met and 1.b.(ii) meeting the TS SR AC assures the necessary HPI and LPI flow qualities are maintained, that HPI and LPI operation will be within safety limits, and that the limiting condition for operation of the ECCS will be met. If the above understanding is correct, please confirm so this question thread can be closed.Licensee Response by Bill See attached file for response to the reviewer response posted on Bentley on 02/20/2008 1/25/08. 1 Licensee Response by Bill The staff understanding posted in the 1/25/08 response is correct.Bentley on 02/28/2008 Additional supporting details were provided in the Davis-Besse
__ _ _ _ _ _ _ response posted on 2/20/08.NRC Response by Ross Telson Thank you for confirming the reviewer's understanding.
of your on 03/21/2008 responses.
The reviewer has no further questions regarding this question thread at this time. Should unanticipated questions arise, a new question may be opened at that time.Date Created: 01/15/2008 10:40 AM by Ross Telson Last Modified:
03/21/2008 10:34 AM http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 200801151040 Response 1. Reviewer states "Regarding SR 3.5.2." Davis-Besse believes this was imeant to be regarding SR .3.5.2.2.
If not, advise otherwise.
L.a. CR 02-04514 has been resolved.L.b. A basic summary of CR 02-045.14 as it relates to SR 3.5.2.2. is as follows: There was an issue with the interface between the IST Program and Design Basis Information.
-,Appropriate corrective actions were taken, for the HPI and LPI pumips to ensure that: IST Acceptance Criteria defines the minimum design required.system flow values IST Acceptance Criteria are Updated for any pump changes or calculation changes IST Acceptance Criteria include instrmentncwertainty values t .c A copy of CR 02-045 14 is provided, with names redacted.lI.d No additional information is deemied necessary to be provided.
NOP-LP-2001-01 Site: G201 CONDITION. CR Number tl " IL., " II02-04514 TITLE:. INADEQUATE INTERFACE' BETWEEN THE IST PROGRAM AND DESIGN BASIS INFORMATION DISCOVERY DATE WIME EVENT DATE TIME SYSTEM / ASSET#8/19/2002 N/A 8/19/2002 N/A .011-01 N/A.,EQUIPMENT DESCRIPTION N/A __" .FLOC System FLOC DESCRIPTION OF CONDITION and PROBABLE CAUSE (if known) Summarize any attachments.
Identify what, when, where, why, how..R During NQAs parallel review of the IST Program, NQA found that the data in the Pump Performance I Curves Procedure DB-PF-06704 Rev. 2 is not used within the IST program. The IST engineer does G not forward new baseline curves to the Pump Performance Curves procedurepreparer for inclusion I in the procedure.
IN A Paragraph ISTB 4.4 Effect of pump Replacement, Repair, and Maintentance on Reference Values T requires that "Deviations between the previous and new set of reference values shall be evaluated, and verification that the new values represent acceptable pump operation shall be placed in the'record of tests."'0 N .Design Engineering uses the:Pump Performance Curves Procedure DB-PF-06704 curves as input in t , heir hydraulic analyses.
In general, pump curve data for calculations is taken from pumptcurve books0 as calculation input. This is an acceptable means of obtaining information from controlled documents.
Over the years, the Service Water Pumps have been modified and/or overhailed.
On some pumps, where the pump has been replaced-with non-vendor tested impellers and/or bowls, the plant test performance data has not been added to the DB-PF-06704 procedure.
The Service Water Pump curves in the IST procedures are different than the curves in DB-PF-067-04 and in various calculations.
For the three Service Water Pumps: 1. The P3-1 IST Curve (DB-PF-03017 Secondary Baseline curve) is above the Pump Performance-Curves procedure curve.2. The P3-2 IST curve about equal to the Pump Performance Curves procedure curve..3. The P3-3IST curve (DB-PF703030 Baseline curve) is below the'Pump Performance Curves procedure curve.The higher curve may have impacts on the diesel loading calculation/condition.
The lower curve impact may be that the curve is below analyzed conditions.
The lower IST curve is almost at the unacceptable level if the DB-PF-06704 curve was the baseline curve for the P3-3 pump. The unacceptable level identified in the DB-PF-03030 Baseline curve is almost 6% lower than an unacceptable curve if based on the DB-PF-06704 curve.The IST- Basis Document for the Service Water Pump was reviewed.
The baseline curve for the Service Water Pump in the Basis Document does not agree with the Service Water Pump curve in the Pump Performance Curves procedure..
For the Service Water pump, a Design Basis analysis of the Service Water System with a 93%degraded curve could not be found. It is not known whether the ASME OM Code 93% acceptance criterion is acceptable relative to the analyzed conditions for the Service Water Pumps.Since, in general, the overhauled pumps are not previously tested pumps by the vendor, it is not known whether additional horsepower is required to operate the rebuilt pumps. Although required horsepower is not an IST requirement, confirmation that a rebuilt pump does not require more horsepower than the original pumps is an item that needs to be tested as required by 10CFR50 Page 1 of 3 NOP-LP-2001-01 Site: G201 COCNDITION REPORT CR Number 02-04514 TITLE: INADEQUATE INTERFACE BETWEEN THE IST PROGRAM AND DESIGN BASIS INFORMATION Appendix B Criterion XI Test Control. When new pump curves show increases in the head capacity curve, a check should be made to assess whether the pump.horsepower requirements have increased and whether there are any impacts on the die sel loading calculations.
It is not clear that this was done for the higher P3-3 pump curve.When the IST Engineer develops a new reference pump curve, he sometimes makes a mathematical curve-fit based on the reference test data. The curve-fit may be higher and/or lower than the tested data. If lower, the acceptance criterion is not corrected for the lower starting point. In other words, 93% of a curve-fit reference curve is not corrected to 93% of the test value or to any .limiting analytical requirement.
There is inadequate interface between the IST Program and Design Basis. Information for pump test*acceptable level determinations.
IMMEDIATE ACTIONS TAKEN / SUPV COMMENTS (Discuss CORRECTIVE ACTIONS completed, basis .for closure.)Discussed the condition with the IST Program Owner and with.the Design Basis Engineering'.
personnel.
Recommerid this as a CAQ at the Apparent cause level, due to the missing Communication between the IST program, to the pump :performance curves procedure.:
The pump performance curves are used.for analytical calculations&#xfd;,.
____._......, QUALITY, ORGANIZATION USE ONLY IDENTIFIED BY (Check one) .Self-Revealed' ATTACHMEN NTS Quality Org. Initiated.
i'Yes .IndividuallWork Group : is Internal Oversight
: .Qualily.Org.
Follow-up
[V Yes E No Suplervision/Management.
External Oversight I .'J Yes t__, No' 6 ORIGINATOR " " " ORGANIZATION DATE SUPERVISOR
' .DATE
* PHONE EXT."QA 8/19/2002 8/19/2002G 7554 Page 2 of 3 NOP-LP-2001-01 Site: G201~r CONk O IEPOR CR Number L... 0 Sit02-04514 TITLE: INADEQUATE INTERFACE BETWEEN THE.1ST PROGRAM AND DESIGN BASIS INFORMATION SRO EQUIPMENT OPERABILITY ORG. IMMEDIATE ORG. MODE CHANGE p REVIEW OPERABLE ASSESSMENT NOTIFIED INVESTIGATION NOTIFIED RESTRAINT L 'Yes * , " "Ys " REQUIRED REQUIRED A 'E'N .Yes* No F .] N Yes i'No. -Yes N/A i YesE] No IM-DEI ASSOCIATED TECH SPEC NUMBER(S)
ASSOCIATED LCO ACTION STATEMENT(s)
-N 4 IN/A' #1 NIA~T:#2.. ... .... ........ .. ..... .. .... ... ...I --. --- ----- ------- -DECLARED INOPERABLE' REPORTABLE?
One Hour NIA " .APPLICABLE UNIT(S)(Date /Time) ,For Hor N/A E 8-19-02/17M35
] Eight Hour U1 Ul j U2 Both R. ..* ." *
* i EvalRequired Other N/A ' -*A,: ......... ........ .. ...........
..... ... ...... .... .'... ..... ......... ... ...........
........ ..... .-.. .-.....T. COMMENTS I Discussed this issue with the originatorand supervisor.
These calculations for new baseline data O0' have the possibility.of not having been:analyzed per the original design. It appears that #3 SW N' PumP is acceptable,'
but this' is not provable from data'supplied, There is a concern with #1 Service Sl, Water Pump from the possibility of mor0 loading than previously analyzed for on the EDG. The #2 pump is the only pump not really in question from looking at the graphs. Due to the programmatic problems of not having the data properly:analyzed from a design standpoint; I declare #1 and #3 SWPs to be INOPERABLE forthis conditionbbut still funti6nal.
,,For this reason, I'consider this to be a mode 4 restraint for TS 3.7.4.1. (SW pumps). Referred to theOlperability Determination procedure.an~dTech.
Specs for this situation.
Sincethe" #2 SWPnearly matches the graph as shown in DB-PF-06704, Operability ofthis pump is not in question.
However, since this pump does notexactly.
match the graph " request an Operability Evaluation for #2 SWPump in additidn to #3 and #1 SWPs to be-delivered 24 hours from the declaration time. This is assigned, # 2002-38 and is aqsinned to Plant engineering at.this time.-currentaode
-UniIi Power Level-Unit I Current Mode- Unit 2 Power Level.- Unit 2 DEF 0 4 N/A N/A.SRO -UNIT 1 SRO UNIT 2 DATE-.8/19/2002 CATEGORY /EVAL ASSIGNED ORGANIZATION DUE DATE ' REPORTABLE?
A DBPE 9/30/2003 Yes I'No HLER No... .... ......-.....-.--......-.-........
G --_ .........
.. .... .. ....* ' A A ' .D B P E .' 'i 9130120 0 3 ..... ...........
...... .E.. .... .........
.........
.... .TRENOCODES CompType/ID Cause If CRPA L REPORTABILITY REVIEWER C Ai .TREND CODES *Compe Typ.I Caus W) Or Process / Activity / Cause.Code(s) (if Cause T or W) Org A Jordan, P SUPV ER2 3950 B04 DES -. ........ ...........
.............
..............................
....... -- ------D T.... 3950 F04 -R DATE M R B ...........
....5 .....................
.........
;.... ...........
................
..... ....... I t .' " ........ 0 3 /0 4 /0 3 MRB 3950. F04 DES 0/40 INVESTIGATION OPTIONS CLOSED BY 'PDATEEvaluation
[-]GnrcIplications
{. Pirt 21 5/420 Page 3 of 3 Attachment Site: G201 CONDITION REPORT CR Number 02-04514 REPORTABILITY DETERMINATION:
This CR identifies a prograqmmatic deficiency with communication bewteen IST engineer and DBE engineer responsible for pump performance curves. IST Engineer apparently does not forward new baseline curves to DBE for includsion in the Pump Performance Procedure.
These curves are used by DBE as, input intohydraulic analyses.'
Based on Ops comments, this apparent:deficiency is focused on the Service Water Pumps. SW Pumps#! and #3 are identified as questionable relative to the analysis of new baseline, data against original design. However, the pumps are considered.
to be operable by Ops. This is a documentation concern.This is an administrative issue which is not reportable.
Page 1 of 1
-qit &#xfd; 1nl CAUSE ANALYSIS CR Number NOP-LP-2001-03 0.2 04.5114 Category Eval Code: :AA: ""..:If Yes,.slel-Iec ne " Condition Description and Cause Basis: .'. .. .e... one 1`Hardwamri Condition Resolution Req No Repar I YSap."". " ' .." .Rework .Use-As~ s PROBLEM STATEMENT Current pump performance curves for safety-related pumps are: not located in the Pump Perf0rmance Curves Procedure DB-PF-06704.
Design.basis inputs from DB-PFr06704 could be effected by not using the most current flow characteristics of safety related pumps.ADDITIONAL ISSUES TO BE EVALUATED PER THIS CAUSE ANALYSIS: As part of. the Corrective Actions associated with this Cause Analysis, other programmatic issues will be addressed for various CRs, including CR 02-05993 and 02-05887.For CR 02-05993, the HPI pump flow acceptance criteriac used to meet the new OM Code requirements-does. not include instrument errors. Also,% the flow acceptancecriterion;is the minimum value used in the LOCA analysis, which .is non-conservative.
The programmatic issuesassociaited with acceptince criteria and instrument.errors are addressed by the.Corrective Actions associated with this Cbndition Report. The problems noted in CR 02-05993 falls under the Human.lPerformance causal factor analysis that has been performed as part of CR 02-04514 Cause Analysis.
s For.CR 02.05887, the current Makeup Pump Surveillance Testing pump degradation acceptance criteria.in DB-SP-03371
:does not agree With the~degradation and.:assum~ptionis inB&W calculation 32-1167143-01, which were. used to demonstrate compliance with General Design Criteria (GDC)33 requirements.
The programmatic issues associated with this accepiance'critriion are addressed by the Corrective Actions associated with this Condition Report; The problem noted in:.CR 02-05887 falls under the Human* Performance.
causal factor analysis that was perormed.as part, of CR.02-04514 Cause Analysis.
CR 02-..05887 will address IST Acceptance Criteria for the Makeup.Pumps.
Programmatic issues addressed by thisCondition Report are referenced in CR 02-07271 for the CCW System. These same programmatic issues were evaluated and identified Under this Cause Analysis.INVESTIGATION RESULTS.TapRoot methodology identified that there were four basic causes that led to problems associated with.pump curves.1). There was no procedure or procedural requirement for includingthe current curves in DB-PF-06704.
: 2) Adequate training was not provided to the IST Engineer in order.to prepare him for OM Code requirements for pumps when he assumed responsibility for the IST Program.3) No means or method of communication was established for IST and Design Engineering to control changes to pump curves.4) There are no Standards, Policies, or Administrative Controls required Plant Engineering to update the pump curve procedure and notify Design Engineering of changes to pump curves.a.) SEQUENCE OF EVENTS The testing of IST Pumps (safety-related) has been performed since the implementation of Safety Guide 26 in the late 70s. A majority of the IST Pumps were never properly base-lined, as required during the first 10-Year Interval of the IST Program. Davis-Besse is currently in its 3rd 10-Year Interval and as many as Page 1 of 10 0.itpa Q.n CAUSE ANALYSIS CR Number NOP-LP-2001-03 02-04514.:five IST pumps still have not been baseline tested.Davis--Besse'Test Performance Engineers within the Plant Engineering Group never understood the requirements for maintaining the design basis of the plant, as it applies to IST Pump Test acceptance criteria.Information and guidance was provided in the issuance of NUREG-1482 in 1995 and Information Notice'IN 97-90 in 1997. In each of these documents, the requirements for having design inputs to IST pump.test acceptance criteria were *learly'identified.
No changes were made to the IST procedures at this time ,(1.997) to include design basis' requirements for differential pressure percentages as it applied to the acceptance criteria to maintain the Davis-Besse design basis.This CR was generated due to conditions found during NQAs parallel review of the IST Program in August of 2002. NQA found, du ring .this. reyiew, that the data in the Pump Performance Curves Procedure DB-PF-;06704 Rev. 2.is notused within the IST Program. The IST Engineer does notforward new baseline.curves to the Pump Performance Curves procedure owner for inclusion in the procedure. -According to this review, over the years, the Service Water Pumps have been modified and/or.overhauled.
On some PUmps wher6 the pump has.been replaced with hon-vendor tested impellers and/or bowls, the plant test p~erformance data has not been added to the DB-PF-06704.procedure.
In relation to compliance with NUREG-1482 and information Notice IN 97-90, the original issue that generatedthis CR was minor in:...nature...
Design basis requirements not being provided in IST.acceptance.criterion is a significant issue to be resolved.,,.%From initial start up of the.plant, until now, Design Engineering has used the Pump Performance Curves-Procedure DB-PF-06704 curves as input in theirhydraulic analyses.
In general, pump curve data for calculations is taken from the pump. cLirve. books as calculation input. This is considered an acceptable
.,means of obtaining information from controlled documents, by Design Engineering.
b.) DATA ANALYSIS The NQA parallel review:6f the iST Program uncovered the information shown below and analyzed Service Water Pump(s) test data and existing, curves. This review has been included in this Basic Cause Analysis due to its thoroughness.
The Service Water Pump curves in the IST procedures are different than the curves in DB-PF-06704 and'*in various calculations.
.1. The P371 IST.curve (DB-PF-03017 Secondary Baseline curve) is above the Pump Performance Curves Procedure curve.2; The P3-2 IST curve is about equal to the Pump Performance Curves Procedure.
curve.3. The P3-3 IST curve (DB-PF-03030 Baseline curve) is below the Pump Performance Curves Procedure curve.'The highercurve may have impact on the diesel loading calculation/condition.
The lower curve maybe below analyzed conditions and could allow the pump to be considered operable due to meeting the IST requirements.
The lower IST curve is almost at the'unacceptable level ifthe DB-PF-06704 curve was the baseline curve.for the P3-3 pump. The unacceptable level identified in the DB-PF-03030 baseline curve is almost 6%lower than an unacceptable curve if based on the DB-PF-06704-curve.
* The IST Basis Document for the Service Water Pump was reviewed.
The baseline curve for the Service', Water Pump in the Basis Document does not agree with the Service Water Pump cujrve in the Pump Performance Curves Procedure..
..Page 2 of 10 Sitpe G201 CAUSE ANALYSIS CR Number NOP-LP-2001-03 02-04514 For the Service Water pump, a Design Basis analysis of the Service Water System witha 93% degraded.curve could not be, found. It is unknown whether the OM Code 93%.acceptance criterion is acceptable relative to the analyzed conditions for the Service Water Pumps.Since the overhauled pumps are typically not tested by the vendor, it is uncertainwhether additional-horsepower is required.
to operate the rebuilt pumps. However, since the pumps are overhauled with original manufacturer parts and to the manufacturer specifications, it is not expected there would be a difference in operating horsepower.
Discussionwith the Service Water system engineer indicated he, checks the motor amps periodically after the pumps are overhauled and has not.observe.d an increase in: running amperage.
Although required horsepower is not an IST requirement, confirmation that:a rebuilt, pump does not require more horsepower thari the original pumps~is.an item that.needs.1to be tested as" required by 10CFR50 Appendix B Criterion XI;Test Control. 'When new pump curves show-increases-in:
the head capacity curve,'a check should be made to assess whether the pump horsepower requirements:
have increased and if there is any .impact on the diesel, loading calculations..
When the IST Engineer develops a~new reference pump curve, he sometimes.
makes a mathematical.
curve-fit based on the referencetest data. The curve-fit may be higher and/or lower than tl-e tested. data.'.If lower, the acceptance.
criterion is notcorrected for the lower startingpoint.
In9other words, 93% of.a curve-fit, reference curve is not corrected to 93% ofthe test value or: to any limiting analytical requirement.
There is' inadequate interface between the IST Programand Design BaSiS Informationfor pumrp test acceptable level determinations.
c.) EVALUATION OF DATA: The initiator.of this Condition Report performed a very th6rough review of the Service Water Pump curve information-for IST and Design. All of the data.appears to be-accurate in relation to the different pump: durves and information listed in the Origination Section .of the Condition Report.: Design BasisEngineering is responsible for.establishing the Minimum Design Basis Flow.requiremenitsfor all of the IST Pumps at Davis&#xfd;Besse..
The IST Program is responsible formaintaining this Design:Basis.
It is the responsibility, of Plant and DesignEngineering to interface so that.'all of the Design flow.requirements for the IST pumps is known so that it can be included in the IST- test acceptance criteria,.as ,required.
An interview-with the IST Engineer indicates that the correlation of.the IST.Pump Curves to the Design Basis were unknown. Much'of the'"tribal knowledge".of the Plant Engineering Group was'not -available at the time that the current IST Engineer took over the program. Over the last year or so,.the IST Engineer has become more aware of.what part IST plays in maintaining the Design Basis for pumps.The Design Flow limit was. thought to be 10,000 gpm, but may be 10,250 gpm. The actual design limit is in question at this time. The IST test and acceptance criteria must conservatively bound the ASME OM 'Code for Operation
& Maintenance of Nuclear Power Plants', Licensing basis and design basis .requirements:
Information Notice IN 97-90 was issued addressing this issue for various pumps based on NRC findings at other plants. Davis-Besse must establish 1sT acceptance criteria for all of the pumps tested in the IST program with input from Design Engineering to ensure the most restrictive requirement is in place. Otherwise, pumps operable per IST.criteria could. be below minimum design requirements., As stated by the Originator of this Condition Report, the "intent of the Code" is to test-and trend'data in order to determine when a pump is degrading so it can be rebuilt prior to being below the minimum design requirements.
The OM Code allows a pump of this design (Service Water is a vertical in-line pump) to be degraded 93% and still be considered operable.
However, the minimum design requirement must be known so that the 93% degradation limit does not allow the pump to be operable below the minimum design requirement..
.Per IN 97-90, "Several recent inspections in the area of safety-related pump performance -have resulted in'the issuance of Notices of Violation of Appendix B, Criterion Xl; "Test Control," of Part 50 of Title 10 of the.Page 3 of 10 Site: G201 CAUSE ANALYSIS CR Number NOP-LP-2001-03 02-04514 Code of Federal Regulations (10 CFR. Part 50) because licensees have concentrated on inservice testing (tST) requirements without ensuring that design requiremeInts were met. There are two applicable primary requirements for safety-related
*pump testing::
(1) to ensurethat-Criterion Xlis met in that each safety--related pump achieves its minimum design-required'performance and (2) to ensure that each safety-related pump meets the requirements of Section XI, "InserviceTesting," of the American Society of Mechanical Engineers Boiler and PressureVessel Code (ASME Code).* Criterion XI of Appendix B to 10 CFR Part 50 requires that a test program be established with written test procedures that incorporate the requirements and acceptance limits contained in applicable design documents.
Although licensees have established IST acceptance criteria that meet the requirements specified iin the ASME Code, the criteria at some plants allowed safety-related pumps to degrade below the performance assured in the accident analyses." Davis-Besse currently.uses the 1995 Edition of the Code with 1996 Addendum (OM Code).Some examples are provided in the Information Notice where plants were not limiting the acceptance criteria for IST pump testing to the most restrictive flow requirements.
The OM Code does not require that'pumps be tested at design-basis conditions.
Many licensees use the OM Code test:to verify compliance
*with the OMCode and the pump design requirements contained, in plantdes ign-basis-documentation and the FSAR.' The OM Code allows a specific percentage of degradation of pumphydraulic performance from an established reference value before:action'must bO taken. If theminimum design performance as specified in the plant design documentation is more stringent than theOM Code acceptance criteria, the-test acceptance criteria must be adjusted to avoid the actual pump performance being allowed to degrade below the minimum acceptable design performance.
In addition, licensees need:,to ensure. that original plant lesign-basi' calcu lations,, or, revisions ito these calculations,:
are properly integrated into surveillarice.
test procedures acceptance criteria.
The NRC published guidance on this issue in NURFEG-1482."Guidelines for Inservice Testing atNuclear Power Plants." Section 5.6. "Opeirability Limits of Pumps,".*states that operabilitylimits&#xfd; must always meet, or be consistent~with, licensing-basis assumptions in a plant's safety analysis.Review of Davis-Besse'ss response to Information Notice 97-90 is required to determine why the IST and:Design Engineers were not communicating as to pump operability requirements.
PCAQR 98-0055 was.retrieved from records and reviewed.
The IST Engineer was the Evaluator for the PCQAR as it applied to IN 97-90. The responses given were based on information about theedesign requirements as stated in the original Bechtel designinformation.
Calculations were identified from the original plant design and many references were made to the 10% pump degradation allowed by the OM Code, and in many cases it was stated that a more conservative design:limit was used. However, during discussions with, Design Engineering,;
HPI System Engineer, Nuclear, Engineering and Plant Engineering it was idenrtified that*Design and Analysis must review changes to:the pump curves and compare the changes to, the Accident Analysis and design limitations for flow of safety-related pumps. Design/Analysis Engineering has agreed to provide the acceptance criteria for all IST Pump testing in order to establish and maintain the design basis for pumps at Davis-Besse.
Additionally, Plant Engineering has agreed to a short term fix to administratively control the re-baselined pump.curves so that Design is notified in advanceas to when* rebuilt or-replaced pumps are to be tested, as notification by Work Control and Maintenance allows. Plant Engineering has agreed to generate a CR requiring review of pump curves when pumps are re-baselihed.
Flow rates for Service Water and Component Cooling water are to be reviewed in an attempt to obtain more margin for pump degradation.-
A new analysis/calculation is being created for High Pressure Injection and the pump tests will be revised to include the new criteria.
All existing IST pump test.procedures will have to be revised to reflect the new criteria and address the administrative controls for pump curves. Some full flow pump rebuild procedures are already in development to allow for pumps to be rebuilt and new 5-point curves to be created outside of the normal quarterly or refueling tests now performed.
According to the interview of the IST Engineer, conducted by the Evaluator, a Relief Request is in place for the Service Water System testing to allow for the use of variable reference values for flow rate and differential pressure during pump testing. This method of testing is discussed in NUREG-1482, Section Page 4 of 10 Site: Q2011 CAUSE ANALYSIS CR Number NOP.LP-2001-03 02-04514&#xfd;5.2. The NRC accepts the use of pump curves for reference values of flow rate and differential pressure if the licensee clearly demonstrates in a relief request the impracticality.
of establishing a fixedset of reference values: To obtain approval for a proposed method of evaluating these pump parameters to detect hydraulic degradation and determine pump operability, the licensee must demonstrate that the acceptance criteria are equivalent to the OM Code requirements in the respective Code Table(s) for allowable ranges using reference values. To use this test method, the licensee must plot a valid pump characteristic curve from an empirical data or obtain one from, the pump manufacturer and verify it with.measurements taken when the pump was known to be in good operating condition.
As an added note, it is unclear to the evaluator how Davis-Besse is using. this method of pump testing when the design basis flow requirements for Service Water have not taken instrument uncertainty into consideration.
Per NUREG-1482, in order to use pump curves as reference values: the licensee must perform the following elements in preparing pump curves for the relief request: (.1) Prepare pump curves, or validate the manufacturer's pump curves, when the pumps are known to be operating acceptably.
(2) When measuring the reference points for plotting or.validating the'curve, use instruments at least as accurate as required by the Code, (3) Construct each curve with a minimum of five points, though few points may be acceptable for a narrow range.(4) Construct the curve with only those.points beyond the "flat" portion .(low.flow rates) of the curves in a range. which includes or is as close as practicable to design basis flow rates.(5) Establish acceptancecriteria for the pumps that do not conflict-with the operability criteria for flow rate and differential pressure inthe'technical specifications or the facility safety analysisreport.
(6) If vibration levels vary significantly over the range of pump conditions, prepare a method for assigning appropriate vibration acceptance criteria for regions of the pump curve.(7) When the repair, replacement, or routine service may have affected the reference curve, plot a new reference curve or revalidate the previous curve by conducting an-inservice test..It is unknown from this requirement how a relief request was submitted and approved without: Design...Engineering input to develop the acceptance criteria.Per NUREG-1482 Section 5.10, when pump test procedures are developed, limits in the safety analysis cannot be ignored. The requirements for inservice testing are written generally.
If specific plant limits are more conservative, to ensure compliance with design basis assumptions, such limits must be clearly indicated as the "operability" limits and used for acceptance criteriaof IST as well. For example, see Section 5.2 (shown above): item (5) of the elements.listed for usingpumpcurves.
The references listed above are from NUREG-1482, published in 1995 discuss ASME Code Section XI test requirements.
The current Davis-Besse IST Program is under the 1995 Edition of ASME with 1996 Addendum (OM Code). Most references apply to the reasoning to be used for pump testing as it.relates to Design Basis Minimum Pump Flow requirements, but Comprehensive Pump Testing is a new test methodology for this 10-Year Interval.
Minimum Design Flow. Requirements are still required to be part of the IST Pump Test Acceptance Criteria, but foP Davis-Besse the pump tests are performed with a specific flow rate that is set and the allowable degradation percentages are applied, to the D/P.CAUSE Four Causes have been identified for the problem of current pump performance curves for safety-related pumps not being shown in DB-PF.06704.
: 1) Procedures
-Situation not covered because: (a) there was no design configuration hold point in the IST pump test procedures for notification and request for design engineering to review and evaluate new pump curves; (b) design engineering was unaware of the minimum design flow requirement (including instrument error) being needed to evaluate pump curves effectively for IST acceptance criteria.Page 5 of 10 ISite: G201.CAUSE ANALYSIS, CR Number NOP-LP-2001-03 02-04514 Reference CorrectiveActions 7, 8 and 10.2) Training-Task not analyzed.
This cause was selected because the engineer that was selected.for the review of Information Notice 97-90 had limited knowledge of IST and Designrrequirements at the time this review was performed.
No training was provided, using NUREG-1482 to bring the engineer Up to speed'on current IST requirements for pump testing. Section 5.2 .and 5.10 of.NUR.EG-1482 clearly explain the requirements for-design basis requirements as it applies to IST acceptance..criteria
,.For many "new" IST.Engineers,.the learning curve to become familiar with ASME Section. Xl,"or OM Code, testing requirements is quite steep. This. training'is not normally performed on site using utility personnel, but is performed offsite using consultants in this-area of expertise.
Therefore he was not equipped with the knowledge or experience needed to properly evaluate the Information Notice to the depth required in order to uncover the need for design input into the IST acceptance crit6ria.
:There was no interface.with experienced personnel to perform an adequate review. The requirement for-des!gn flow requirements was interpreted to mean the'"original" design requirements.
ReferenCe -CAF 02-07525-024 and .discussion in Corrective Action Sectionfor benchmarkingand..training.
: 3) Communications
-No Communication (communication system was\ not in' place). (a) DEEC Design.(DBE) was unaware of the requirementto properly identify instrurnent unceriainty and its impact on'the minimum design basis flow requirements and'how the lack of this information could:impact the iST ,Program and Pump Operability. (b) NED (DBE) was not required to include the instrument uncertainty and degraded-voltage (Diesel Loading) into IST acceptance'criteria.
Reference.
Corrective Actions 6, 7, 9, 10, 11, 12, 1314, 15, 16,:17; 18, 19 and20 forcommunication between Design "andPlant.Engineering.
4): Management System -Standards, Policies, or Administrative Controls (sPAc) not in place. (a) No Standards,:
Policies or Administrative Controls were. in place.in.the IST Procedures or Design Engineering
...Calculation procedures requiring verification of the current pump.curves or updates of existing curves, based on re-baseline.of pump curves, (b) No requirement was in place for Desigrn.Engineering topreview new pump curves.and compare them to the~a'cident analysis :or other designm limiting requirements.
Reference -Corrective Actions 6, 7, 8, 9 and'10 for Design Engineering involvement in"devel6pment of, acceptance criteria and measures put in place within procedures"to notify design, ofcha.nges to pump curves and/or updates of existing 'curves based on re-basetin'efo pulmrp curves. Design Engirneering review required for all new pump curves based on these corrective actions.CORRECTIVE ACTIONS The Corrective Actions listed below are shown as written with a brief explanation for each CAF.1. Determine if Davis-Besse should issue an Operating Experience(OEL)
Report according to NG-NA-* 00305, step 6.7.3.(ThisCorrective*Action, generated by PI)2, Per Operability Evaluation 2002-38 (see attached), place a" Mode 4 festraint on the SW system pending resolution of this condition report. This will involve determining the acceptability.
of service water pumps to provide flows that meet TS requirements.(This Corrective Action generated by DES (Nuclear) to address SW. flow and TS)3. Per Operability Evaluation 2002-38 (see attached), place a Mode 4 restraint on the EDGs system pending resolution of this conditioh report. This will involve a re-evaluation of:the EDG load tables using current pump loading.(This Corrective Action generated by DES (Nuclear) to address EDG Loading issues).4. Roll-over CAF -Ensure that the issues identified in 02-05887 are addressed in this Cause Evaluation.-Page 6ofl1O Site: G201 CAUSE ANALYSIS' CR Number NOP-LP-2001-03 02-04514 ReferenceCAF 02-05887-04.
Recommend Mode 4 restraint...
Assign to Enercon.: (This Corrective Action was generated by PES for rollover of CR 02-05887 to address programmatic issues between Design and Plant Engineering) .The MakeupPump programmatic issues.have been addressed, but additional corrective actions from 02-05887 and 02-5756 arelto addrbess acceptance criteria for Makeup Pump Test requirements.
: 5. PE/SYME ensure that the Basic Cause Analysis for CR 02-045,14 addreses the cause for CR 02- .05993 and that Corrective Actions developed in CR 02-04514 address.the common, generic issues f.or the two.CRs. Assign to Enercon (This Corrective Action was generated by PES for rollover of CR 02-05993 to address programmatic issues between Design and Plant Engineering)
: 6. NED provide IST Acceptance Criteria which defines the minimum. design required system flow value(s),:
that will govern pump testing to PE for the following pumps.The IST Pumps are: Component Cooling Water-Pumps Decay Heat Removal Pumps ;High Pressure Injection Pumps Service Water-Pumps The lST and System Engineer(s) recommends that this is a Mode 4 Restart action..(This Corrective Action was generated by PES and selected only the pumps listed at the request.of Nuclear Design. This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4)Management System.)7 ".,Revise existing IST procedures to require a.Condition Report to be generated forNED to review all new pump curves created as a result of a rebuilt, refurbishedor replaced Program. This.,Condition Report will be used by NED to evaluatei the new pump curve based on Design Flow-Requirements and Assumptions to maintain the System Design Basis.The IST Engineer considers this Corrective Action to not be required for Restart.(This Corrective Action was generated by PES for including new IST acceptance criteria into procedures.
This CAF satisfies Basic Cause (1) Procedures.)
: 8. Revise existing IST procedures (pump testing) to include the calculations for Minimum Design Flow Requirements as Technical References so that changes can be linked for updates when Calculations are..revised that effect IST Pump Testing Acceptance Criteria..
The IST Engineer considers this Corrective Action to not be required for Restart.(This Corrective Action was generated by PES for including new IST acceptance criteria into procedures.
This CAF satisfies Basic Cause (1) Procedures.)
: 9. NED provide IST Acceptance Criteria which defines the minimum design required system flow value(s), that will govern pump testing to pE for the AFW Pumps, The IST and System Engineer(s) recommends that this is a Mode 3 Restart action, (This Corrective Action was generated by PES and selected only the pumps listed'at the request of Nuclear Design. This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4)Management System.)10. IST Engineer tocreate an administrative control within the IST Administrative procedure to govern the Page 7 po 10 Site 0G201 CAUSE ANALYSIS cR Number NOP-LP-2001-03 02-04514 changes to pump curves and notification of Design Engineeringforf support of evaluations and review to, ensure the design basis is maintained.
The IST Engineer has indicated this action is required Post Restart...(This Corrective Action was generated by PES for creation of a mechanism to evaluate new pump curve.test data: This CAF satisfies Basic Causes (1) Procedures, (3).Communications and (4) Management System.)11. Add a prerequisite to the pump baseline tests for the Service Water pumps to.gather the following data at a point(s) near the maximum hydraulic horsepower:
(.1) Head (2) Flow (3) Perform energized motor testing per DB-PF-05064 to obtain Motor Voltage and Current data Note: The PdMA report is recommended for voltage and current The IST Engineer recommends this action be addressed Post Restart (This Corrective Action, as well as 12 through 16, was generated by.PES for evaluating changes to pump* flow characteristics that could affect calculation inputs and assumptions.
These CAFs satisfy.Basic Causes (1) Procedures, (3) Communications and (4) Management System.)12.. Add.a prerequisite.to the pump baseline tests for the Decay Heat pumPs to gather the following data at,:: a pOint(s) near the maximu'm hydraulic horsepower:.Same requirements listing as.CAF #11.13. Add a prerequisite to:the pump baseline tests for the High Pressure Injection pumps to gather.the.
following data at.a point(s) near the maximum hydraulic horsepower:
Same requirements listing as CAF #1&#xfd;1 14. Add a prerequisite to the pump baseline tests for the Component Cooling Water pumps to gatherr the following data ata point(s) near the.maximum h&#xfd;,draulic horsepower:
Same requirements listing as CAF #11* 15. Add, a prerequisite to the pump baseline tests for the Makeup pumps to gather the following data at a.point(s) near the maximum hydraulic horsepower:
Same requirements listing as CAF #11 16. Add a prerequisite to the pump baseline tests for the Containment Spray pumps to gather the following data at a point(s) near the maximum hydraulic horsepower:
Same requirements listing as CAF,#1 1 17. Design Electrical.to provide Instrument Uncertainty values to Nuclear Design and Mechanical Design for calculation input for IST Acceptance Criteria for the following pumps, as required.Service Water .Pumps Decay Heat Pumps (LPI).High Pressure Injection Pumps Component Cooling Water Pumps Containment Spray Pumps The IST Engineer has determinedthat this action should be addressed prior to Mode 4, .*(This Corrective Action was generated by PES at the request of Nuclear Design, in order to allow Electrical Design to supply the necessaryinputs for the calculations required for the above listed pumps.Page 8of 10 Site: G201 CAUSE ANALYSIS CR Number NOP-LP-200103.
02-04514:This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4) Management System.)18 Design Electrical to provide Instrument Uncertainty values to Nuclear Design and Mechanical Design:for calculation input for IST Acceptance Criteria for the AFW pumps, as required,.The IST Engineer has determined that this action should be addressed prior to Mode 3.;* (This Corrective Action was generated by PES at the request of Nuclear Design, in order to allow* Electrical Design to supply the necessary inputs for the calculations required for the above listedpumps.
This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4) Management System.).19. Design Electrical to provide Motor Frequency (RPM) values to Nuclear Design, Mechanical Design and Plant Engineering for calculation input for IST Acceptance Criteria for the following pumps, as required.Service Water Pumps (Nuclear)DecayHeat Pumps (LPI) (Nuclear)High Pressure Injection Pumps (Nuclear)Component Cooling Water Pumps (Nuclear)Containment Spray Pumps (Mechanical)
Makeup Pumps (Test and Performance)
The IST Engineer has determined that this action should.be addressed prior-to Mode 4.(This Corrective Action was generated by PES at the request of Nuclear Design, in order to allow**Electrical Design to supply the.necessary inputs for the~calculations required for the above listed pumps.This CAF satisfies Basic Causes:(1)
Procedures, (3) Communications and (4) ManagemenrtSystem.)
-20. Mechanical Design to provideIST Acceptance Criteria which defines the minimum design required ,* system flow value(s), thatwill govern pump testing for the Containment Spray Pumps to PES.The 18T Engineer considers this action to be required for Mode 4.(This Corrective Action was generated by PES at the request Of Nuclear Design, inworder:to allow.Mechanical Design to supply the necessary calculation required for the Containment Spray pumps. This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4) Management System.)'In relation to Basic Cause (2) Training -Task not Analyzed.* Corrective Action 02-07525-024 has been created to: Develop a plan to increase engineering personnel involvement and knowledge of industry activities" including:
: a. Benchmarking of other utilities b. Participation in industry working groups, and c. Assessments at other utilities.Set the expectation that best practices and incorporated into station conduct of engineering, activities.
No further action is required for Basic Cause (2) of this Condition Report for Plant Engineering personnel to benchmarking and participation in workshops (CR.02-07525-024..credited)..
CR 02-05887 Corrective Actions will address Acceptance Criteria for the Makeup Pumps. Long term Motor Frequency data acquisition will be addressed under *Corrective Actions for 02-04514.
" In summary, the (4) four areas shown in the Investigation Results Section, the four areas are resolved as follows: (1) A special procedure or additional steps in administrative procedure is addressed under Corrective Action. Number 10.(2) IST Engineer was new to his position and needed more skill/knowledge to complete the task that could Page 9 of 10 Ritp., r2f0l CAUSE ANALYSIS CR Number NOP-LP-2001-03 02-,4514 have identified a need for design requirements to be included in IST acceptance criteria is addressed under Corrective Action .02-07525-024, as discussed above.(3) Communication has been established across organizational boundaries for interface between IST and Design.(4) Administrative controls (generation of CR and s8T Admin procedure change) have been put in place to require interface of the tST Engineer and Design Engineering for evaluation'of pump data during re-baseline testing. Additional experience will be gained EXTENT OF CONDITION As part of the Cause Analysis for this Condition Report, all of the pumps in the IST program were evaluated.
The only pumps in the program. that received the support of Design Engineering were the High Pressure Injection Pumps.. The marginfor flow in the system and the LQCA analysis were known by design to be tight, so emphasis was given to these pumps with design input to IST Pump acceptance criteria.As Extent of Condition relates to CR 02-05993, which was a CA category Condition Report written on the.HPI pump flowacceptance criteria for OM-1 Code testing not including instrurfnenterror clearly identifies the lack of good engineering practices by design engineering and plant engineering as it relates to "ST a pump acceptance.criteria.
Even where the margiIns were known to be tight, instrument uncertainty was not included in the acceptance criteria.
The Corrective Actions for this Basic Cause Analysis will prevent reoccurrence of this deficiency.
INFORMATION ADDED A RESULT OF QA REJECTION.
QA rejection of this CR is documented in CR 03-03603.
Refer to CR 03-03603 for all followup actions!,necessary to resolve -the QA concerns.
CA 02-045,14-22 created for documentation.
.-Process Code Trend Codes.ER2 .-.c.useisT.., ".. .causeComponent Code Activity Code Cause Code -Type. IDp Cause Org* 3950 Primary. B04 Design analysis DES DesignEngineenng 3950 Secondary F04 Config/design changes PES Plant Engineering 3950 Tertiary F04 Config/design changes DES Design Engineering Completed By: D ate -:. I D:a -____ _____ ____ _ ": 3/10/2003.
Page 10 of.10 Site: G201 GENERIC IMPLICATIONS CR Number: : .:..:.:,ii
: ..:.i , '0 2-04 5 1.4 NOP-LP-2001-02" 0-41.Past occurrences of the issue at the site.Document Number:
 
== Description:==
" Previous Response: Various See continuation sheet See continuation sheet.... ..... .... .. ........ ....... .. .7 ... .. .. ..........
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*.. continued on attached Past occurrences of the issue in the industry.Document Number:
 
== Description:==
 
.. Provious Response: LER 315-99032
..See continuation sheet N/A OE 14444 See continuation sheet None L ER 275-.97012 SeeO& cntinu ation sheet .N/A OE 13659 See continuation sheet Not complete ,: .... ......._ ...... .. ...:. .._ ':. .. ...'. : ..... ... ... .... ........... .. ..... .. , : .: .. .. ..,= ... ..... ..............
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:.... ------. .. ..-- -........ .* .. ........ .... ........ ... .... .... ........ ..... .....*. continued on attached Experience Review*Questions:
1'.,Do past occurrences of similar conditions (as identifie~dabove) indicatea generic or broader scope issue?Condition reports.identified for check valve testing; heat exchanger.pertiomance teajing and relief valve testing indicate that a similar issue may exist withother lST program components'and other sURveillarice/performance testing. Specifically that design requirements for safety related equipment are not properly identified and translated to surveillance/peiformance test procedures and that testing results are not available for review and use by clesign engineering.
: 2. Discuss theeffectiveness of prior corr ective actions-for similar identified conditions (if applicable).
''How are currently proposed preventive action(s) different so as to be more effective (if applicable)?
Conditions identified for pump performance testing have been resolved by.CR 03-03603 corrective actions which required verification that IST program pumps are capable of meeting.desigrn requirements and providing appropriate acceptance criteria to the iST program.Corre&tive action 23 to CR 02-04514 will address training and experie.nce issues for the IST program.* Condition report 03-07765 initiated to address issue for other IST components or surveillancelperformance test procedures.
Extent Of Condition Questions:
: 3. Based on your knowledge and the results of the database review, is the condition present in other identical or similar.equipment, processes, programs or applications?
Condition may exist in other IST program components or with other surveillance/performance testing.4. Was a new CR initiated?
Yes 5. Why / Why Not?Recommend that this issue be evaluated as part of CR 03-06909:
Design Control Collective Significance Review,..... ....~ ~ ~ ~ ~ ~- -- -- --- .. ......L ....... .. ... ....... .... .... .... ... ... ... .........
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.... .Completed By: DATE: q911612003 Page 1 of 3 Sitp- G21Q1 GENERIC IMPLICATIONS CR Number*NOP-LP-2001-02 02-045.14 Past Site Occurrences (Continuation Sheet)Search of the CREST database for the two year period prior to the initiation of this CR (08/1912002) for cause code B4;'design analysis, identified 165 condition reports. A similar search of the CREST database for cause code F4. configuration/design ch!anges identified 1.7 condition reports.A search of CREST going forward from the initiation of condition report 02-04514 identified additional condition reports related to-the!ST program. Five with cause code of B4 or F4 and additional cause codes.Condition reports related to.the IST program and the basic causes of. CR 02-04514 were categorized as follows: Pump Curves and related calculations and acceptance criteria: CR 01-00518 CR 02-04585 CR 02-07524 CR 02-09150 CR 02-04485 CR 02-07380 CR 02-06863 CR 03-00550 CR 02-06996 CR 02-05993 CR 03-03603 CR 02-07006 Instrument accuracy/analytical methods for design requirements related to pump performance CR 02-06422 CR 02-07271 CR 02-08342 CR 03-06487'..These condition reports are concerned with the design basis, acceptance criteria and maintenance of the associated calculations and pump curves for pump, performance verification by testing: Included in this is assuring that the stated acceptance criteria*
..includes appropriate allowance for instrument inaccuracy, pump degradation and:operating conditions; that the acceptance criteria is properly translated to testing procedures (including compliance with. ASME Section X); and that testing results are available for review and use by design engineering.
Design basis flow rates for check valve testing CR 02-08635 CR 02-07657 CR 02-09024 Calculation/configuration control for CcW heat exchanger performance testing CR 03-02231 CR,03-02202 CR 02-06342 CR 02-06342 * ..CR'02-06303 Also related is OR 02-00529 concerning the, expanded tolerances for relief valve settings.These condition reports are related in that.they identify issues with the acceptance criteria utilized for.component performance testing and maintenance of the design basis,. .I Training and industry activities CR 02-04306 OR 02-04324 .These condition reports concerned the IST program owners reduced industry involvement and the TSM-413 Qualification Ca .rd being out of date: Past Industry Occurrences (Continuation Sheet)Search of the INPO web page for "inservice testing program" returned 145 records.Search of the INPO web page for "ist program" returned 300 records.Search of the INPO web page for "design analysis" returned 300 records.Search of the INPO web page for "design analysis" and "inseriJice testing program" returned 35 records.Search of the INPO web page for "pump curves" returned 81 records.A survey of the returned records identified that the majority were LERs submitted for individual components or groups of components which were not include in the IST program or not properly tested. No specific incident was identified that is similar to this event where testing results and design basis criteria were not properly correlated and evaluated.
LER 315-99032 D, C. Cook Unit 1: submitted January 2000 identified that while in extended shutdown deficiencies in the IST program including failure to verify stroke timing of certain valves, failure to include valves in the program and failure to verify position'indications. "The station determined that the causes of the ISTSprogram deficiencies were the lack of knowledge on the American Society of Mechanical Engineers (ASME) codes and licensing and design basis of the plant." IST managemen!
received industry training on the IST program development and implementation responsibilities and experienced qualified IST personnel were'employed to correct technical deficiencies in the program. .OE 14444 identified that at the Perry Plant engineering calculations used to quantify degraded emergency service water pump performance had been done improperly.
Degraded performance was determined by reducing pump head at all points by ten percent. The proper methodwould be to reduce flow rate by seven percent and calculate the corresponding head characteristics.(August 2002)Page 2 of 3
-itp r,901 GENERIC IMPLICATIONS CRNumber." " .... ."-". ' ii ..,:::0-04514 NOP-LP-2001-02 02.-04 14 LER 275-97012 Diablo Canyon 1: the station determined that testing of CCW pump 1-3 had been-performed using referencevalues of flow rate and differential pressure rather than a fixed reference flow rate,'The regulatory relief to use reference curves had not* been obtained LER was submitted August of 1997.*OE 13659 engineers at the Perry. Plant identified during testing of a rebuilt reactor. atei- clearup purnm that pump performance
* characteristics were significantly better than expected.
Performance did not:match the. vendor:supplied pump curves. The cause .was that the vendor had polished the pump internalswithout performing pump characteristic testing again.. The remaining reactor water cleanup pump .was also rebuilt in a similar manner. Replacement pump curveswere obtained and the station will require that pump curves be developed following future rebuilds. (April 2002)LER 271-99001 Verrnont Yankee:. the station determined that.the pumrip curves used to assess, pump performance for the residual heat removal service water had not been correct ly.applied.to the station.configuration..
'he curves used were obtained trom the...manufacturer and were specific to the pump design for instailation in a standing poo"However,the Pumps Supplied to the station were a canned design. .Page 3 of 3 Site: G201.10CFR21I Decision Applicability Checklist CR Number_NUOP4.pLP2001_04*
* -.. 02-0451.4 Does the Condition Report involve: lnto:mation obtained or an observation made of a BASIC COMPONENT that could I .Yes ..No comn~promise safety. ;" .,".(See loic tlow dialram defining tersnd ane ppiibaiifornmatibn on the next page.)if the answer is No, Stop here da rate on thwOrnoinator Signature Tab)If t he answe'r is Yes, Items A& B mist be anewered, rParl,) A & B tabi A. Does the Condition Report involve a-i3AbIC COMPONENT cf a .ian structure.
system, come.Onenli or )art thorcof necessary* Iassure:" Yes LiNo*. 1 ' Ti 'I lnlem ty of the reactor eool~int pressur e beundarvy.
"...- Ys [:}N 2. 'The capability to shutdowh the mehctor and naintain it in safe shutdown ,condition. .y. .No.3 The capabilit, to prevent. or m;itieate the ccnercuences of dccdent,.hict) could result .Yes i Noyes: ipdeitn (irslte puslus nYpar ble to those refrendltlo iMt&#xfd; 1 IFit0l. 1. , B Does the potential issue or defect involve *" 1. I :A deviation in a dlelivened cnipnel *. -" yes No 2- Deviation in a portion of a f"cilit, offereod for a-cep ce' .N D3 es IQii I. intala I tion .tc so, r o r''atforl Ofai dfe o,,,e s tructare.
svstem or ' es I No.." coh aenent?4. A cornitioni circumtance that could cor-tribet' to exceedine a Techi, " " Yes No', .*Yes ..pe.t i If anryitemrs in A are marked 'Yes' AND any items in B are marked 'Yes', contact Regulatory Personnel immediately to discuss and determine if a SUBSTANTIAL SAFETY HAZARD may exist, or if the issue is reportable.
Based on discussions with Regulatory Personnel that a SUBSTANTIAL SAFETY HAZARD or reportability issue does not exist, provide explanation I justification below: Although the requirements of "Minimum Design Basis Flow Requirements", as stated in NUREG-1482 and Information Notice IN 97-90, as it applies to IST Pump testing acceptance criteria, was improperly implemented, no pumps have been identified as being below their minimum design requirements.
Interviews and discussions with Plant Engineering (TP and SYME) and Design* Engineering (NED) indicate that Where margins were known to be tight (HPI), NED was involved in revising calculations and stop valves were adjusted to obtain more margin. However, failure to properly implement the requirements put the plant "at risk". The current understanding of the requirements and the corrective actions associated with this Condition Report should address all of the issues needed in order to become compliant with regulatory guidance and-preclude reoccurrence of the "at risk" situation.
Based on the determination that a SUBSTANTIAL SAFETY HAZARD or reportability issue may exist, draft a Corrective Action Form (CAF) to be accepted by the Regulatory Personnel to complete the 10CFR Part 21 requirements forthe CR.CAF Generated?
L- Yes i' No (if no, provide explanation I justification above)If Yes, CAF#,..............
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Completed By: DATE: 2/4/2003 Page 1 of 1 Site: G201 CORRECTIVE ACTION cRNumber: NOP-LP-2001-05 02-04514 CR Category:
Action Type: Schedule Type: ... .CA*Number:
AA " ) (A) Owner Assigned/Controlled 1 Corrective Action Type: Cause Code: ? .Resp 0g: ES ) Evaluation Support *(NA) Not a Deficiency
..." PEs'R
 
== Description:==
 
... .........1. Determine if Davis-Besse should issue an Operating Experience (OE). Report according to NG-G NA-00305, step 6.7.3. (The OE coordinator, John Johnson at 8345:can provide assistance)
" 2. If no.E report should.be issued, document the reasonswhy.
in.the CAF,.Implementation N Response section.A 3 If an DE report should be issued arrange to have an OE report issued by Davis-Besse within 50:-T ,.days after the event. Notification of the OE coordinator,.
John Johnson at 8345 or O jjjohnson@firstenergycorp.com is suggested to assist in arranging the action.R This assignment was made by the MRB and is due at the completion of the CR evaluation.
The 50-.day target is from an INPO recommendation.
C m pleted By: ."" " ........". ....co" Organiza~tion:
D te:. -T .. Phone: Attachments::r". , i -PI 8121/2002 8Y9d ,. ... s Ye"If a.Refueling Outage is-required,.
*'. ' ...' *..OterTracking
#.: :Corrective'Action D66e DaW&#xfd;e:i .[-AC- A Enter the Refueling Outage number: NA AN/A ..1200:-P ---- -__- ...:.. ..... ..........
....__.._. ...L ...... .._ -- __'" -- ..... :.: :: EPT Approval:.(Enter Name and Sign) " Section::.
Date: PES, .8/21/2002 QUAL- Quality Organization Approval:.
.Date: :* ITY -8/28/2002-I Response:. (J Completed as written .0-RevisedlAlternate.Solution , Not Performed M This Condition ReportwasWritten based on the fact:that the Pump Performance Curves Procedure P is not used within the IST Program. Discussions with Design Engineering:and Plant Engineering L indicate no differential pressures.
were below theminimum design. requirements for thesepurnm s', E Lack of Design input to IST Acceptance Criteria is being addressedfunder CAF #61.M E The lack of Communication between Design Engineering and :Plant Engineering in..relation to.the, N requirements for updating pump performance curves has been addressed by Regulatory Guidance T published since 1995.. NUREG-1482, Section 5.2 and 5.10 published in 1995 and Information I Notice IN 97-90 clearly address the requirements for design basis requirements and instrument.
N uncertainty to be included in IST Acceptance Criteria.
Per Section 5.10 of NUREG-1482, "When G pump test procedures are developed, limits in thesafety analysis cannot be ignored. The requirements for inservice testing are written generally.
if specific plant limits are more 0 conservative, to ensure compliance with design basis assumptions, such limits must be clearly.R indicated as the "operability" limits and used for acceptance criteria of IST as well." Due to G .guidance on this issue being published in.NRC guidance, and no events found that led to serious degradation of operating safety margin or affected core reactivity, core cooling or decay heat removal, no OE Report should be issued for this Condition Report.This Corrective Action is.considered closed.Alternate Corrective Action or Justification if Corrective Action not performed:
Corrective Action Implementation Date: 2/20/2003.J Signature indicates Corrective Action cQMplete: Page 1 of 59 Site: G201" ~CORRECTIVE ACTION .. CR Number: NOP-LP-2001-05.
02:. .... '-.
Completed By:. ".ll __ .Date: 2/20/2003 j Signature indicates verification for:SCAQ'.CRs
...Verified By: Date:.j .n e .a ............ .i , ." .....: '--. .. .--......-
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J Enler Namne and Sig tn Implementing Organization Approval:
Date: 21&#xfd;25/2003 Q U A L I"T Y V E R E Comments: The information is correct and the conclusion is acceptable.
The iss'ue of confirming that the analytical requirements are not more limiting than, the ASME'allowed criteriais alfeady known and published in both NUREG 1482 and NRC IN 97-90 As such it is: unnecessary to issue an OE on the condition;.
Approval: Date: 2/26/2003 Page 2.of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05
._02-04514 CR Category:
Action Type: Schedule Type: I CA Number: AA ( (). ( E ) Refueling/Forced Outage 2 Corrective Action Type: Cause Code: .Resp Org: (CA) Corrective Action (B06) Prog/process weak DES"0 -D e c r p t o n ..... ..... .... .. ._ ... ..............
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== Description:==
 
I Per Operability Evaluation 2002-38 (see attached), place a Mode 4 restraint on the SW system.G pending resolution of this condition report. This will involve determining the acceptability of service.I water pumps to provide flows that meet TS requirements.
N A T 0 R , ...... ...,..Completed By:, Organization:
Date Phone ,,Ahments:
DBE 8/29/2002 " 8567. jY No If a Refueling Outage is required, Other Tracking # Corrective Action Due Date: ACCt Enter the Refueling Outage number: 13RFO ... N/A 3/28/2003 EPT Approval: (Enter Nam e and."iS n '........
...... S.gn) ...... ...Sec.tion : .
DES 8/292002 QUAL Quality Organization Approval:
Date:.ITY Date: ""-.9/4/2002 I Response:
.
written (' Revised/Alternate Solution C.&#xfd; Not Performed M Called the mode restraint team (Greg Estep) and reviewed this CR Mode restraint category to -P verify that this CR has a Mode 4 restraint placed on it.. Although the SW system itself. does~not have L a Mode 4 restraint, the plant.cannot go to Modle ..4without resolving thiScondition report.. Therefore, E in effect there is a Mode 4 restraint on the SW system: Steve Martin of the cR.Mode restraint team.M agreed with this determination.
E. Alternate Corrective Action or Justification if Corrective Action not performed:
N ,._,_*_,__
T N G " " Corrective Action Implementation Date: __3/21/2003..., SigitUllre indicates Corrective Action complete: 0 Completed By: Date: 3/2i/2003 R j Signature indicates Verification for SCAQ CRs: G Verified By: .Date: j Enter Name and Sign: Implementing Organization Approval:
Date: 3/24/2003 O V Comments: u E Verified that condition report 02-04514 is listed as a Mode 4 hold on the computer S drive. The A R desired outcome of this corrective action will be satisfied by the Mode 4 restriction placed on CR 02-L I 04514. Nuclear Quality Assurance accepts the corrective action resolution based on the I F TI implementation of the implementation of the discussion text.Y E Implementation concurred by Approval: Date: 417/2003 Page 3 of59 -.
Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-04514.CRCategory:
Action Ty.e: .ScheduleType:
CA Number: AA :(E) Refuehn rced Outage T:3 Corrective Action Type: Cause Code: Resprg:.(CA) Corrective Action j. (806) Prog/process weak DES.R
 
== Description:==
 
I Per Operability Evaluation 2002-38 (see attached), place a Mode 4 restraint on the EDGs system'G pending resolution of this condition report. This will involve a re-evaluation of the EDG load tables I using.current pump loading..N'A T 0 R CuA pletea By: Organization Appro oa::.. ..Date : DBE , 92. .I Y/4/ 20 If detuermined Othatte EDs wequred oealfrM Ode ,thereforeathisg ModerestrineAt isnlongerDt:
LCC. applicabe and t ane. (mE 03R 009 was don for CR 03-0-- ----------.-.. ................
C0 3 has bee ct I o e EDGI 2 Load Tables and 1042 and EI arerevisd E andr the reyisio usaeis the pump/motor
: brk h ,orsp er asdtrmac ined by,. -ISTretiestiongThi Dacteion' ' :~DES !! : 8/30/2002 QUAL-. T ,".Quality Organization
.. .*. " * ..". : ,... .". .:[.,"" L -Date.: .*'" 9 ,20 2 : "..,. , .. .. Responise.., ..'. (' Completedlas w te ..ilY r e ,...' .iM ' 'Operabiliiy&#xfd;Evaluatiion OE 2003-009 using the increased service water motor
,.:" , P determined that-the&#xfd; EDGs were operable for Modes l-6,: therefore th .is, Mode. restraintiS no longer ." L..applicable, and this. CA Can be closed. (QE .2003-009 was done, for C R:03-00949)
..,. :,-: 101 CA 02-&#xfd;05385-19 has been created to ensure the EDG Load Tables.E-1042 and E7.1043 are-revised
.E and the revision uses the pump/motor brake horsepower as determlined by IST testing. This~actiOn N can be performed post restart due to the approved operability evaluation 2003-009 done for CR03-T 00949..N 9/8/03 Update.G Service Water Pump 1.Measurements in Baseiine Test on 9/3/03 (See scanned attachment to CR 03-07279 for PdMA data)" KW 442.75 KVAR 259.06 KVA 513.07 Calculated BHP 558.5 based on Efficiency of 94.1%R " G MPR EDG Transient analysis, Calc C-EE-024.01-008 (Used for EDG Operability Evaluation 2003-009 for.CR 03-00949)
.BHP 571.ETAP Horsepower Calculation, C-EE-015,03-007 BHP 576
 
== Conclusion:==
 
For Service Water Pump 1 the BHP value used by MPR in EDG Transient analysis.Calc C-EE-024.01-008 (Used for current Operability Evaluation for EDG) of 571 Hp and ETAP Horsepower Calculation, C-EE-015.03-007 of 576 Hp conservatively envelope the measured BHP of 558.5 Hp.Page 4 of 59 Site: G201*CORRECTIVE ACTION cR Nu mber: NOP-LP-2001-05.0-41 CR 03-07425 has been initiated to ensure Calculation C-EE-015.03-006 and EDG Load.Table E-1042 are revised to reflect the information observed in this baseline test. .9/10/03 Update This update is to provide basis for demonstrating that the increase in hydraulic horsepower observed during baseline testing of Service Water Pump 2 (SWP-2) is enveloped by current analysis of loading on EDG 2.As indicated on page 4 of Operability Evaluation 2002-38, the BHP requirements of SWP-3 have increased to 554 BHP according to pump affinity laws.MPR EDG Transient analysis, Calc C-EE-024.01-008 (Used for EDG'Operability Evaluation 2003-009 for CR 03-00949) was performed using the loading requirements of EDG71 because it enveloped the loading of EDG-2 which.would include SWP-2. In the MPR analysis the BHP:for service pump motor is BHP 571. This would envelop the increase to 554 for ,SWP7-2..In ETAP Horsepower.Calculation, C-EE-015.03-007, page 26, the BHPfor SWP-2 IS 98% of600 Hp or 588 BHP.Conclusion:.
BHP value used by MPR in EDG Transient analysis,.Calc C-EE-024.01-008 (Used for.current Operability Evaluation for EDG) of 571 Hp and ETAP Horsepower Calculation, C-EE-01 5.03-007.6f 588 Hpconservativel y envelope the calculated BHP of 554 Hp.":"i -CR 03-07425 has been initiated to ensure the EDG Load Table E-1043 is revised to include the most accurate available information.
Alternate Corrective Action or Justification if Corrective Action not performed:
., .*. ..... ..Corrective Action Implementation'.
Date: 9W10/2003'J Sign, to indicates Corrective Actio6'complete:
Completed By: ' Date: 9/10/2003_J Signature indicates verification for SCAQ CRs: Verified By: Date:, Enter Name ancI Sign" Implementing Organization Approval:.
=0. Date: 9/10/2003 Q v Comments: U E Operability Evaluation 2003-009 references two calculations that supports its position.
Calculation C-A R EE-015.03-006 Rev. 0 and C-EE-024.01-008.
Calculation C-EE-015.03.006 does not directly L I. identify the increased brake horsepower service water was used but references calculation C-EE-'I 015.03002 that lists the service water pump loads of 540, 520 and.550 horsepower for pumps P3i,1 Y E P3-2 and P3-3 respectively.
These loads are from the original service water pump curves.R Section 5 of calculation C-EE-024.01-008 identifies that. a sevice water brake horsepower of 571 was used. This reference was based on an assumption that the increased loads matched a condition that could be evaluated per the Pump Affinity Laws. Quality Assurance has previously issued condition report 03-04835 (6/19/03) that addresses the usage of pump affinity laws. While the use of the affinity laws may not be acedemically appropriate it provides a conservatively high value for horsepower~that in this case is acceptable for the intended application.
This corrective action is acceptable Concurred by NQA.-Page 5 of 59 (201 CORRECTIVE ACTION CR Number:.02-04514 ,NOP-LP-2001-05..
::.Removed 3 signature and 8/23/03 date on 9/8/03. Kb The 9/8/03 update on the Brake Horsepower (BHP) test data identified that the test BHP was 558.5.Since the MPR calculation used 571 HP and the Electrical Transeint Analysis Project conservatively used 576 HP, the resolved situation is acceptable.
Nuclear Quality Assessment (NQA) previosuly stated that the 571 HP value determined by applying the pump affinity laws was a'conservative method to determine the horsepower.
The test data i558.5 HP) demonstrated that this was a correct approach.
Concurred by -Nuclear Quality Assurance.
This corrective action remains acceptable.
Approv~~ Date: .9/112003..
Approval:
9/11/2003..
bt..',' ., ...-; ,, .;.,. .: ' I.p-J Page 6 of 59 Site: G201 CORRECTIVE ACTION cR Number: NOP-LP-2001 05 02-04514 CR Category-Action Type: Schedule Type: CA Number: AA .) " (E )Refueling/Forced Outage 4 Corrective Action Type: Cause Code: Resp Org: O PR) Preventive Action (NA) Not a Deficiency " PES R
 
== Description:==
 
Roll-over CAF -Ensure that the issues identified in 02-05887 are addressed in this Basic Cause G Evaluation.
Reference CAF 02-05887-04, N Recommend Mode 4 restraint.
A Assign to R' Completed By: --........
Date: Phone: Attachments:
,1/23/2003.
8 Yes No If a Refueling Outage is required, Other Tracking # Corrective Action Due Date: ACC_ Enter the Refueling Outage number:. 13RFO N/A 3/7/2003-O D a e ..:...... .....P Approval: (Enter Name and Sign) section:.
e:** E .2.3,3PES 1/23/2003 QUAL Quality Organization Approval:
j Date: 1TY .3/19/2003 Response:
..() Completed as written "(, RevisedlAlternate Solution C) Not Performed M For CR 02-05887, the current. Makeup Pump Surveillance Testing pump degradation acceptance P criteria in DB-SP-03371 does not agree with the degradation and assumptions in B&W calculation L 32-.1167143-01, .which Were used to demonstrate compliance with General Design Criteria (GDC)E 33 requirements.
.The programmatic issues associated with this acceptance criterion are addressed M by the Corrective Actions associated with this Condition Report. The problem noted in CR 02-E11 05887 falls under the Human Performance causal factor analysis that was. performed as part of CR N *02-04514 Basic Cause Analysis..I Four Basic Causes were identified for this Condition Report. Three of those causes were related to N CR 02-05887.G 1) Procedures
-Situation not covered because: (1) there was no.design configuration hold point in o the IST pump test procedures for notification and request of design engineering to review and R evaluate new pump curves; (2) design engineering was unaware of the minimum design flow G requirement (including instrument error) being needed to evaluate pump'curves effectively for IST acceptance criteria.3) Communications
-No Communication (communication system was not in place). (1) DEEC Design (DBE) was unaware of the requirement to properly identify instrument uncertainty and its impact on the minimum design basis flow requirements and how the lack of this information could impact the IST Program and Pump Operability.
(2) NED (DBE) was not required to include the instrument uncertainty and degraded voltage (Diesel Loading) into IST acceptance criteria.4) Management System -Standards, Policies, or Administrative Controls (SPAC) not in place. (1)No Standards, Policies or Administrative Controls were.in place in the IST Procedures or Design Engineering Calculation procedures requiring verification of the current pump curves or updates of Page 7 of 59 Site: G201 CORRECTIVE ACTION cR Numbe~r.NOP-LP-2001-05 0-41 existing curves based on re-baseline of pump curves. (2) No requirement was in place for Design -Engineering to review newipump curves and compare.them to the-accident analysis or. other design.limiting requirements.
The Corrective Actions listed below are shown as written with a brief explanation for each CAF.6. NED provide IST Acceptance Criteria which defines the minimum design required system flow value(s), that will govern pump testing to PE.The IST Pumps are: Component Cooling Water Pumps Decay Heat Removal Pumps High Pressure Injection Pumps Service Water Pumps The tST and System Engineer(s) recommends that this is a Mode 4 Restart action.(This Corrective Action was generated by PES and selected only the pumps listed at the request of Nuclear Design. This CAF satisfies Basic Causes.(1)
Procedures, (3).Communications and (4)* Management System.).7. Reviseexisting IST procedures.to require a Condition.Report to be generated for NED to review.all new pump curves created as a result of a rebuilt, refurbished or replaced pump within the IST*Program: This Condition Report Willbe used by NED to evaluate.the new pump curve based on.Design Flow Requirements and Assumptions to-maintain the System Design Basis..The.IST Engineer considers.this Corrective Action to not be required for Restart.(This Corrective Action was generated by PES for including new ISTacceptance criteria into* procedures..
This CAF satisfies Basic Cause (1) Procedures.)
: 8. Revise existing IST procedures (pump testing) toeinclude the calculations forMinimum Design Flow Requirements as Technical References so that changes can be'linked for updates when Calculations are revised that effect IST Pump Testing ACceptance Criteria:
.The IST Engineer considers this Corrective Action.to not be required for Restart.(This Corrective Action Was generated byPES for including new IST acceptance criteria into procedures.
This CAF satisfies Basic Cause (1) Procedures.)
: 9. NED provide IST Acceptance Criteria which defines the minimum design required system flow value(s), that will govern pump testing to PE for the AFW. Pumps.The IST and.System Engineer(s) recommends that this is a Mode 3 Restart action.(This Corrective Action was generated by PES and selected only'the pumps listed at the request of Nuclear Design. This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4)Management System.)10. IST Engineer to create an administrative control within:the IST Administrative procedure to govern the changes to pump curves and notification of Design Engineering for support of evaluations and review to ensure the design basis is maintained.
The IST Engineer has indicated this action is required Post Restart.(This Corrective Action was generated by PES for creation of a mechanism to evaluate new pump.curve test data. This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4)Page 8 of 59 Site: G201 CORRECTIVE ACTION -CR Number: NOP-LP-2001-05 02-04514 Management System.), 11. Add a prerequisite to the pump baseline tests for the Service Water pumps to gather the following data at a point(s) near the hydraulic horsepower:
(1) Head (2) Flow (3) Perform energized motor testing per, DB-PF-05064ito obtain Motor Voltage and Current data Note:.The PdMA report is recommended for voltage and current The IST.Engineer recommends.this action be addressed Post.Restart (This Corrective Action, as well as 12 through 16, was generated by PES for evaluating changes to.pump flow characteristics that could affect calculation inputs and assumptions.
These CAFs satisfy Basic Causes (1) Procedures, (3) Communications and (4) Management System.)12. Add a:prerequisite to the pump baseline .tests for the Decay Heat pumps to gather the following data at a point(s) near the maximum hydraulic horsepower:
Same requirements listing as CAF #11 13. Add a prerequisite to the pump baseline tests for the High Pressure lnjection pumps to gather the following data at a'point(s).near the maximum hydraulic horsepower:.
Same requirements listing as CAF #11 14. Add a.prereq uisite to the pump baseline tests -for the Component Cooling Water pumps to gather the following data at a point(s) near the maximum hydraulic horsepower:
Same requirements listing as CAF #11 15. Add a prerequisite4o the pump baseline tests for the Makeup pumps :to gather the following data at a point(s) near the maximum hydraulic horsepower:
Same requirements listing as CAF #11 16. Add a pierequisite'to the pump baseline tests for the Containment Spray pumps to gather the following data at a point(s) near the maximum hydraulic:
horsepower:
, Same requirements listing as CAF #11.: 17,.Design Electricalto provide Instrument Uncertainty values to Nuclear Design and Mechanical Design for calculation input for IST Acceptance Criteria for the following pumps, as required.Service Water Pumps Decay Heat Pumps (LPI)High Pressure"Injection Pumps Component Cooling Water Pumps Containment Spray Pumps The IST Engineer has determined that this action should be addressed prior to Mode 4.(This Corrective Action was generated by PES at the request of Nuclear Design, in order to allow Electrical Design to supply the necessary inputs for the calculations required for the above listed* pumps. This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4) Management System.) .18. Design Electrical to provide InstrumentUncertainty values to Nuclear Design and Mechanical Design.for calculation input for IST Acceptance Criteria for the AFW pumps,, as required.The IST Engineer has determined that this action should be addressed prior to Mode 3.: Page 9 of 59 qif- r'201' ~CORRECTIVE ACTION c ubr NOP-LP-2001-05
.'02-04514 (This Corrective Action was generated by PES at the request of Nuclear Design, in order to allow Electrical Design to supply the necessary inputs for the calculations required for the above listed pumps. This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4).Management System.)19. Design Electrical to provide Motor Frequency (RPM) values to Nuclear Design, Mechanical Design and Plant Engineering for calculation input.for IST Acceptance Criteria for the following.
pumps, as required, Service Water Pumps (Nuclear).
Decay Heat Pumps (LPi) (Nuclear)High Pressure Injection Pumps (Nuclear)Component Cooling Water Pumps (Nuclear)Containment Spray Pumps (Mechanical)
Makeup Pumps (Test and Performance)
The IST Engineer has determined that this action should be addressed prior to Mode 4.(This.Corrective Action was generated by PES at the request. of Nuclear Design, in order to allow Electrical Design to supply the necessary inputs for the calculations required for the above listed pumps. This CAF satisfiesBasic Causes (1) Procedures, (3) Communications and .(4) Management System.)..
: 20. Mechanical Design to provide IST Acceptance Criteria which defines the minimum design required system flow value(s), that will govern pump testing for the Containment Spray.Pumps to FES.The IST Engineer considers this actionto be required for Mode 4-., (This Corrective Action was generated .by PES at the request of Nuclear Design, in order to allow Mechanical Design to supply the necessary calculation required for the Containment Spray pumps.: This'CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4) Management System.)The Programmatic Issues that resulted in the Makeup Test Criteria not havi.ng Design Engineering input has been addressed by the changes made in communication between the IST Program .Owner and Design Engineering.
The corrective actions of 02-05887 Condition Report will.correct the procedure issues for these pumps, and the test procedure and acceptance criteria requirements of CR 02-04514 corrective actions.should prevent reoccurrence of this issue. No further actions are required for this Corrective Action.Alternate Corrective Action or Justification if Corrective Action not performed, Corrective Action Implementation.
Date: 3/7/2003 J Signature indicaztes Corrective complete:.Completed By: Date: 3/7/2003 J Signattre indicates verification for SCAQ CRs: Verified By: Date: J Enter Name and Sign: Implementing Organization Approval:
vR.....ZT
_ ..Date: 3/7/2003.Page 10 of 59 Site: G201 CO RRECTIVE ACTION :, ".CR Number:: i / , I 24 oQ V Comments: U E The human performance causal factor review that was performed in the cause analysis of this*A .R condition report satisfactorily addressed the progranmatic issues.of -thiscorrective action. This correctiVeaction has been verified as acceptable.
Concurrent verification performed by Y E Approval:
J Date:. 4/8/2003 Page 11of 59 Site: G201 CR Number: CORRECTIVE ACTION C ubr NOP-'LP-2001-05.
02-04514 CR Category:
Action Type: Schedule Type: CA Number: AA (") .(E) Refueling/Forced Outage 5 Corrective Action Type:. Cause Code: Resp Org: (PR Preventive Action. (806) Prog/lrocess weak PES R
 
== Description:==
 
PE/SYME ensure that the Basic Cause Analysis for CR 02-04514 addresses the cause for CR 02-G ' 05993 andthat Corrective Actions developed in CR 02-04514 address the common, generic issues for the two CRs.-N-A* T Assign to McGaha'0.R...........
........ .... ... .. ...7 7-CompletedBy, Orga-ization
[ Date: , , * , Phone: " Attaehment
" PE .1/2'3/2003 8326 .F-Yes viNo If. a Refueling:.Outage is required, Other Tracking # Corrective Action Due Date: A CC Enter the Refueling Outage number: .13RFO _ N/A 3/7/2003 E P T ro T.............
............
........ .-. ..........
... .... ... ............
... .......................
..... .. ......... T6 ,t 7 ------ .. .--. .-PT Approval:.(Enter; Name and Sign) m : Date: QJAL. Quality Organization Approval:
.." Date: 1/320 ITY -., 3/19/2003" 'Response:
C 0 Complketedas Revised/AlterIate Solution' Not Performed W. ForCR 02-05993, the HPI pump flow acceptance criteria used tomeet the new 0M-I 'Code.-P requirements doees not include instrument-errors.
Also,. the flow acceptance criterionis the L rT 'inimum value used in the LOCA analysis, which is non-conservative.
The programmatic issues E associatedwith acceptance criteria and instrument errors are addressed by the Correcctive Actions M associated~with this Condition'Report.
The problems noted.in CR 02-05993 falls under the Human E Performance-causal factor analysis that has been performedas part of CR 02-04514 Basic Cause N Analysis.T*I Four Basic Causes were identified for this Condition Report. Three of those causes were related to N CR 02-05993.1) Procedures
-Situation not covered because: (1) there was no design configuration hold point in 0 the IST pump test procedures for notification and request of design.engineering to review and R evaluate new pump curves; (2) design engineering was unaware of the minimum design flow, G requirement (including instrument error) being needed to evaluate pump curves effectively for IST acceptance criteria.3) Communications
-No Communication (rommunication system was not in place). (1) DEEC Design (DBE) was unaware of the requirement to properly identify instrument uncertainty and its impact on the minimum-design basis flow requirements and how the lack of this information could impact the IST Program and Pump Operability' (2) NED (DBE) was not required to include the instrument uncertainty and degraded voltage (Diesel Loading) into IST acceptance criteria.4) Management System.- Standards, Policies, or Administrative Controls (SPAC) not in place.. (.1)No Standards, Policies or Administrative Controls were in place in the IST Procedures.or Design Engineering Calculation procedures requiring verification of the current pump curves or updates of Page 12 of 59 Site: G201 CORRECTIVE ACTION' Rube::.NOP-LP-2001-05
....- ..: ..".i., " 02 0 .5 4 ":.existing curves based on re-baseline of pump curves. (2).No requirement was in.place for Design Engineering to review new pump curves and compare them to the accident analysis or other design limiting requirements.
The Corrective Actions listed below are shown as written with a brief explanation for each.CAF,, .6.: NED provide 1ST Acceptance Criteria whichdefines the minimum design required system flowv.value(s), that will govern pump testing to PE.The IST Pumps are:'Component Cooling Water Pumps Decay Heat Removal Pumps.High Pressure Injection Pumps Service Water Pumps, The IST and System Engineer(s) recommends that this is a Mode 4 Restart action.(This Corrective Action was.generated.
by PES and selected only the pumips listed at the request: of'.Nuclear Design. This CAF satisfies.
Basic Causes (1) Procedures; (3) Communications:and (4).,.,.Management System.)7. Revise.existing IST procedures to require- a Condition Report to be generated for NED:to:review all new .pump curves created as a result of a rebuilt, refurbished or replaced pump.within theIJST:-Program..
This Condition Report will be used by NED to evaluate the new pump curve based on Design Flow Requirements and Assumptions to maintain:
the System Design Basis .The IST-Engineer considers-this.
Corrective Action to not be required for Restart.'.(This Corrective Action was generated by PES for including new, lST accepta nce criteria in to procedures.
This CAF satisfies BasicCause (1) Procedures,)
.8. Revise existing ST procedures (pump testing) to include.the caiculations for Minirium*Design':" Flow Requirements as Technical References so that changes can be linked for updates.when.'
Calculations are. revised that effect IST Pump Testing Accebtance Criteria.The IST Engineer considers this Corrective Action to not be required for Restart": (This CorrectiveAction was generated by PES for including new IST acceptance criteria into procedures, This CAF. satisfies Basic Cause (1) Procedures.)
: 9. NED provide iST Acceptance Criteria which defines the minimum design required system flow value(s), that will govern pump testing to PE for the AFW Pumps.The IST and System Engineer(s) recommends that this is a Mode 3 Restart action,: (This Corrective Action was generated by PES and selected only the pumps listed at. the request of Nuclear Design.. This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4)Management System.)*.10. IST Engineer to create an administrative controlwithin the.IST Administrativeprocedure to govern the changes to pump curves and notification of Design Engineering for'support of evaluations and review to ensure the design basis is maintained.
The IST Engineer has indicated this action is required Post Restart. ..(This Corrective.Action was generated by PES for. creation of a mechanism.to evaluate new pump curve test.data.
This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4), Page 13 of 59 Page 13 of 59 Sitp C,201 CORRECTIVE ACTION : CR Number: NoP-LP-2001-05.
.:i:: :: ; : .02-04514 Management System.).11. Add a prerequisite to the pump baseline tests for the Service Water, pumps to gather the following data at a point(s) near the maximum:hydraulicc horsepower:
(1) Head (2) Flow (3) Perform energized motor testing per DB-PF-05064 to obtain Motor Voltage and Current data Note: The PdMA report is recommended for voltage andcurrent The IST Engineer recommends this'action'be addressed Post Restart.(This Corrective Action, as well as 12 through 16,was generated by PES.for:evaluating changes to pump flow characteristics that could affectocalculation inputs and'assumptions.*
These CAFs satisfy Basic Causes (1) Procedures, (3) Communications and (4) Management System.)12. Add a prerequisite to the pump baseline.tests for the Decay Heat pumps to gather the following data;at a.poiht(s) near the maximum hydraulic horsepower:.
Same requirements listing as CAF #11 13. Add a prerequisite to the pump baseline.tests for the: High Pressure Injection pumps to gather the following data at a point(s) near the maximum hydraulic horsepower:
Same requirements i'isting::as CAF #41 14. Add a prerequisite'to the:. pump baseline tests for:the Component Cooling Water pumps to.gather the following data at a point(s) near the maximum hydraulic horsepower:., Same requirements listing as CAF #11 -15. Add a prerequisite to thepump baseline tests for the Makeup pumps togather thc following data at a point(s) near the maximum hydraulic horsepower:..
Same requirements listing as CAF #11 1i6. Add a prerequisite to the pumpbaseline tests for the: Containment Spray pumps to gather the: following data at.a point(s) near the maximum hydraulic horsepower:
Same requirements listing as.CAF #11': 17. Design Electrical to provide Instrument Uncertainty values to Nuclear Design and Mechanical Design for calculation input for ISTAcceptance Criteria for the following pumps, as required.Service Water Pumps Decay Heat Pumps (LPI)s High Pressure Injection Pumps Component Cooling Water Pumps Containment Spray Pumps The IST Engineer has determined that this action should be addressed prior to Mode 4.(This Corrective Action was generated by PES at the.request of Nuclear Design, in order to allow Electrical Design to supply the necessary inputs for the calculations required for the above listed pumps. This CAF.satisfies Basic Causes (1) Procedures, (3) Communications and (4) Management System.)18. Design Electrical to provide Instrument Uncertainty values to-Nuclear Design and Mechanical Design for calculation input for IST Acceptance Criteria for the AFW pumps, as required.The IST Engineer.has determined that thisaction should be addressed prior to Mode 3.Page 14 of 59 Site: -G201 CORRECTIVE ACTION OR Numlber: NOP-LP-2001-05 " ,"02-04514 (This Corrective Action was generated by PES at the request of Nuclear Design, in order to allow Electrical Design to supply the necessary inputs for the calculations required for the above listed ,pumps. This CAF satisfies Basic Causes (1) Procedures, (3) Communications.
and (4) Management.
System.)19. Design Electrical to provide Motor Frequency (RPM) values to Nuclear Design, Mechanical Design and Plant Engineering for calculation input for IST Acceptance Criteria for the following pumps, as required.Service Water Pumps (Nuclear)Decay Heat Pumps (LPI) (Nuclear)High Pressure Injection Pumps (Nuclear)Component Cooling Water-Pumps (Nuclear)Containment Spray Pumps (Mechanical)
Makeup Pumps (Test and Performance)
The IST Engineer has determined that this action should be addressed prior to Mode 4..: (This Corrective Action was generated by PES at the request of. Nuclear Design, in order to allow Electrical Design to supply the necessary inputs for the calculations required for the above listed pumps. This CAF satisfies Basic Causes. (1). Piocedures, (3) Communications and (4) Management
".System..)
: 20. Mechanical Design to provide IST.Acceptance Criteria.which defines the minimum design required system flow value(s), that will.govern pump testing for the Containment Spray Pumps to: PES.:The 1ST Engineerconsiders this-.action, to berequired-for Mode 4.(This Corrective Action was generated by PES at therequest of Nuclear Design, in orderlto allow Mechanical Design to supply the necessary calculation required for the Containment Spray pumps., This CAF satisfies Basic Causes (1) Procedures, (3) Communications and (4).Management.
System.)The Programmatic Issues that resulted in the HPI Pump Test Criteria not having the required Design Engineering input has been addressed by the changes made in communication between the.IST Program Owner and Design Engineering.
The corrective actions of 02-05993 Condition Report.will correct the procedure issues for these pumps,. and the test procedure and acceptance criteria requirements of CR 02-04514 corrective actions should prevent reoccurrence of this issue. No further actions are required for this Corrective Action.Alternate Corrective Action or Justification if Corrective Action not performed:
Corrective Action implementation Date: 13/72003.J Signature indicates Corrective Action complete:-Completed.
By: Date: 3/7/2003 J Signature indicates verification for SCAQ.CRs: Verified By: Date: A Enter Name and Sign: Implementing Organization Approval:
Date: 3/7/2003 Page 15 of 59 Site: .G201 CORRECTIVE ACTION , ...Rume: NOP-LP-2001-05 02-041 a V Comments:.
u E The Human Performance causal factor review that was performed as part of the cause analysis A R satisfactorily addressed the prgramatic issues relative to acceptance criteria and instrumentaerror
.'documented by this corrective action. This action has been.verified an determine T I acceptable.
Conccurrentverification performed by'Bob Stanley.Y_. E..R Approval:
... Date: A/8/2003 Page 16of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 " 02-04514 CR Category:
Action Type: Schedule Type: CA Number:.AA ( ) (E ) Refueling/Forced Outage -6 Corrective Action Type: , Cause Code:. .Resp Org: (PR) Preventive Action (F04) Config/design changes *DES R Description,:
I NED provide IST Acceptance Criteria which defines the minimum design required System flow..G value(s), that will govern pump testing to PES.N The IST Pump are: A T Component Cooling Water Pumps O Decay Heat Removal Pumps High Pressure Injection Pumps R Service Water Pumps TheIST and System Engineer(s) recommends that this is a Mode 4.1Restart action.. __. _. *.Completed By: Organization:
oDate: T Phone: Attachiments-PE 1/30/2003 7637 ' Yes .No If a Refueling Outage is required, Other Tracking #". Corrective Action Due Date: ACC- Enter the Refueling Outage number: 13RFO -N/A '' " '.. '7. I ./ I.r23/2003-EPT Approval:- (Enter Name and Sign) 1Section:
Date: DES 3/7/2003 QUAL, QualityOrganization Approval:
Date: ITY 8/23/2003
.Completed as writen C) Revised/Alternate Solution C) Not performed.
M Please refer to the attached response.
The response Was attached for formattin'g considerations,.
P L E M E Alternate Corrective Action or Justification if Corrective Action not performed:
._," , N T, N G Corrective Action Implementation Date: 7/18!2003 G J Signature indicates Corrective Action complete: O " Completed By: ___ Date: 7/18/2003 R .. Signature indicates verification for SCAQ CRs: G Verified By: *
* _Date: J Enter Name and Sign: Implementing Organization Approval:
Date: 7/18/2003 O V Comments: u E This corrective action required Design Engineering to provide acceptance criteria to Plant A R Engineering for Inservice Testing (IST) for thecomponent cooling water (CCW), decay heat (DH), L I high pressure injection (HPI), and the service water (SW) pumps. CCW ,DH, HPI criteria is..T established in calculation C-NSA-016.04-001 Rev. 1, C-NSA-049.02-33 Rev. 0, C-NSA-052.01-003 Y E Rev.6 respectively.
The criteria for service water is based on a 5% degraded pump curve from DB-R PF-06704 using the P3-3 curve based onProto-PowerCalculation 03J-13 Rev. A.Page.17 of 59/.,..
Site: G201* CORRECTIVE ACTION CR.Number:
NOP-LP-2001-05 Q 02-045,14, Based on the review of the established acceptance criteria the [ST. pump testing for the systems , previously listed; this corrective action is resolved.
Concurred byW Approval:
Date: 9/4/2003 Page 18 of 59 Site.: G201 CORRECTIVE ACTION CR Number: NOP-LP;2001-05 02-04514 CR Category:,., Action Type: Schedule Type: CA Number: AA () (E) Refueling/Forced Outage 7 Corrective Action Type: Cause Code: Resp Org: (ES ) Evaluation Support (B06) Prog/process weak PES R
 
== Description:==
 
I Revise existing IST procedures to require a Condition Report to be generated for NED to review all SG new pump curves Created as a result of a rebuilt, refurbished or replaced pump within the IST Program. This:Condition Report will be used by NED to evaluate the new pump curve based on-N Design Flow Requirements and Assumptions to maintain the System Design Basis.A AT The IST Engineer considers this Corrective Action to not be required for Restart.0 R -,..., ........ ..... .....Completed By: J Organizaion:
Date.:- Phone: Attachments: " .." PE .1/30/2003 7637 Yes N .If a Refueling.Outage is required, Other Tracking # Corrective Action Due Date: ACC- Enter the Refueling Outage number:. N/A .N/A ..9/30/2004 EPT Approval: (Enter Name and Sign) Section: Date:...I ,: ..... ' " ' l PES .."i 2/1/2003 ..OUAL. Quality Organization Approval:
Date: S ITY" " .. .." -**3/16/200:4
-: I
* Response: ( ' Completed as written (h Revised/Alternate Solution Not Performed M iST Pump baseline testing procedures gather data to develop new pump performance curves.P following maintenance.
This corrective action requires alteration of existing.
IST pumpbaseline L testing procedures to include a step requiring a Condition Report to be initiated for Design.E Engineering toevaluate the effect of the rebuild on the design basis. IST Program pumps.include M .the Auxiliary Feedwater (AFW) Pumps, Service Water (SW) Pumps; Containment Spray (CS)E .Pumps, Component Cooling Water (CCW) Pumps, Decay Heat (DH) Pumps, Boric Acid (BA)N Pumps, Emergency Diesel Generator (EDG) Fuel Oil Transfer Pumps, Makeup (MU) Pumps and.T High Pressure Injection (HPI) Pumps..N AFW Pumps: DB-PF-03251, Revision 02 and DB-PF-03260, Revision 02 are currently effective G and have been altered in accordance with this corrective action.O, SW Pumps: DB-PF-03218, Revision 01; DB.PF-03224, Revision 02 and DB-PF-03233, Revision 02 R are currently effective and have been altered in accordance with this corrective action. DB-PF-G 03217, Revision 01; DB-PF-03223, Revision 01 and DB-PF-03232, Revision.01 are currently inactivated but have been altered in accordance with this corrective action.CS Pumps: There are no existing IST procedures that perform baseline testing of these pumps.However, DB-PF-03437 and DB-PF-03438 are in development.
CA 02-04514-29 will address incorporation of this corrective action into DB-PF-03437.;
CA 02-04514-30 will address incorporation of this corrective action into DB-PF-03438.
CCW Pumps: DB-PF-03572, Revision 02; DB-PF-03573, Revision 02 and DB-PF-03574, Revision 03 are currently effective, and have been altered in accordance with this corrective action.DH Pumps: DB-PF-03236, Revision 01 and DB-PF-03237, Revision 02 are currently effective and Page 19 of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05
.02-0.4514 have been altered in accordance with this corrective action.BA Pumps: DB-PF-03550 and DB-PF-03551 have not been altered in accordance with this corrective action. CA 02-04514-25 will address the alteration to DB-PF-03550.
CA 02-04514-26 will address the alteration to DB-PF-03551.
EDG Fuel Oil Transfer Pumps: There are no existing IST procedures that perform baseline testing of these pumps. Due to system design and construction, these pumps are tested per a Relief Request in the IST Program and no test procedures are needed to develop a pump performance.
curve for these pumps.MU Pumps: There are no existing IST procedures that perform baseline testing of these puinps.However, DB-PF-03472 and DB-PF-03477 are in development.
CA 02-04514-31 will address .incorporation of this corrective action into DB-PF-03472.
CA 02-04514-32 will address incorporation of this corrective action into DB-PF-03477.
HPI Pumps: DB-PF-03407, Revision:06 and DBPF-P03408, Revision 06 are currently.effective and have been altered in accordance with this corrective action. DB-PF-04207 (being replaced by DB-PF-03082) and DB-PF-04208 (being replaced by DB-PF-03083) have not been. altered. in accordance with this corrective action. CA 02-04514-27 will address the alteration to DB-PF-04207.(DB-PF-03082).
CA 02-04514-28 Will address the alteration to.DB-PF-04208 (DB-PF-03083).
:Alternate Corrective Action or Justification if Corrective Action not performed:
correctiveAction Implementation-Date:
'.8/16/2004 A Signature incdcates Corr, tive Action complete: Completed By: Date: 8/16/2004 BIgn y indicates verification for SCAQ CRs: Date : J Enter Name and Sign: 4 Implementing Organization Approval:
* .Date: 81712004 Q V Comments: U El The procedures listed in the implementation section of this condition report for the pumps of the A R following plant sytems: auxiiliary-feedwater (AFW), service water (SW), component cooling Water L I (CCW), decay heat (DH) and high pressure injection (HPI) pumps were verified to satisfactorily T Iaddress the concern of the conditiorn report. For the pumps of plant sytems containment spray (CS).y E make up (MU), boric acid (BA), and high pressure injection (HPI), the corrective actions P assignments have been established within this condition report to provide final resolution.
The response relating to the emergency diesel generator (EDG) fuel oil transfer pumps&#xfd;,documented by a relief request within the IST Program, is acceptable.
Based on the implementation response of this corrective action, it can be closed..Approval:
A.... .Date: 8,21/2004 Page 20 of 59 Site: G201 CORRECTIVE ACTION cR Number: NOP-LP-20014-5 02N04514 CR Category:
Action Type: Schedule.
Type: CA Number: AA E) Refueling/Forced Outage. 8 Corrective Action Type: Cause Code: Resp Org: (ES ) Evaluation Support (B06) Prog/process weak DBTS.. ....... ..... .R:
 
== Description:==
 
Revise existing IST procedures (pump testing) to include the calculation's for Minimum Design Flow G Requirements as Technical References so0thatchanges can be linked. for updates when Calculations are revised that effect IST Pump Testing Acceptance Criteria.N A The IST Engineer considers this Corrective.
Action to not be required for Restart.0 R Completed By: Organization Date: Phone: Attachments:
d 1 PE 1/30/2003.
763 Yes ~N If a Refueling-Outage is-required, Other Tracking # j Corrective Action Due Date: ACC- tnter the Refueling Outage number: NA N/A 7/2112006: ,EPT Approval:
,(Enter Name and Sign), " Section: Date: DBTS. 2/1/20.03QualityOrganization Approval:
Date: ITY ..
* 8/25/2003 Repose ..
ted as, wriften. U : R et w Revised/AIternate Soluotin , Not Perforrned M CR 02-04514 identified inadequate:dbommunicaiion between Plant and DesignrEngineering*
P personnel with respect to pump design basis requirements andtthe associated acceptancecriteria L being used in test procedures.
Numerous corrective.actions were, created to address this issue.E This.CorrectiVe Action was intended to create a link between the. pump testing procedures and the M calculations that define their acceptance criteria.N This corrective action as written. requires existing IST Pump testing procedures to be revised to.T include the calculations-for minimum design flow requirements as technical references so that I changes can .be linked for updates when calculations are revised. The affected IST Pump testing.N -procedures consist of the following:
.G Auxiliary Feedwater Pumps: DB-SP-03151, DB-SP-03160 O Boric Acid. Pumps: DB-SP-03450, DB-SP-03451 R. Component Cooling Water Pumps: DB-PF-03072, DB-PF-03073, DB-PF-03074; DB-PF-03075 G Containment Spray Pumps: DB-SP-03337, DB-SP-03338 Decay Heat/Low Pressure Injection Pumps: DB-SP-03136, DB-SP-03137 Emergency Diesel Generator Fuel Oil'Transfer Pumps: DB-PF-03201, DB-PF-03202 High Pressure Injection Pumps: DB-PF-03207, DB-PF-03208, DB-SP-03218, DB-SP-03219 Makeup Pumps: DB-PF-03372, DB-PF-03377, DB-SP-03371, DB-SP-03377 Service Water Pumps: DB-PF-03017, DB-PF-03023, DB-PF-03030, DB-PF-03117, DB-PF-03123, DB-PF-03130 Actions have been taken for the following pumps: Boric Acid Pumps: The acceptance criteria for these pumps are defined by Calculation 034.009.This calculation is listed .in the reference section and discussed in the purpose section of test Page 21 of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-04514 procedures DB-SP-03450, Revision 12 and DB-SP-03451, Revision 12. All actions complete for these pumps.Containment Spray Pumps: The acceptance criteria for these pumps are defined by Calculation C-ME-061.01-078.
This calculation is listed in the reference section of test procedures DB-SP-03337, Revision 11 and DB-SP-03338, Revision 14, All actions complete for these pumps.Emergency Diesel Generator Fuel Oil Transfer Pumps: The acceptance criteria for these pumps are defined by Calculation 016.05. This calculation is listed in the reference section and discussed in the purpose section of test procedures DB-PF-03201, Revisiori 04 and DB-PF-03202., Revision 03. All actions complete for these pumps: Makeup Pumps: The acceptance criteria for these pumps are defined by Calculation C-NSA-065.01-.
0.19. This calculation is listed in the reference section and discussed in the purpose section of test procedures:DB-PF'03372, Revision 07 and DB-PF-03377, Revision 09. Actions remain for test procedures DB-SP-03371 and DB-SP-03376.
For the remaining IST pumps, this corrective action (to link the pump test procedures to the calculation):
has been implemented as follows: At the time thiscorrective action was created,'differential pressure versus flow (DP vs Flow) acceptancecriteria'values or curves were contained
, within the specific individual surveillance testing procedures.
Part of the resolution tothe:overall, .issue identifiedby this CR was to relocate pump acceptance criteria curves into procedure'DB-PF-06704 to consolidate pump curve information into-a single document.
The acceptance criteria curves in DB-PF-06704 clearly indicate the acceptable regions of pump performance.and~contain
.reference to the calculation that was used. to generate the,.curve.
The specific surveillance tests have been altered to contain instruction to.obtain a copy of the appropriate curve from DB-PF-.06704 and plot the test data 'point to determine whether acceptanceecriteria-are met. Therefore:
the*linking of the test procedures to the appropriaie curve in DB-PF-06704 fulfills'the intent of this.,corrective action. This is incorporated into DB-PF-06704.
Revision 17.and the pump-specific:
procedures listed below'Auxiliary Feedwater Pumps: The acceptance criteria for these pumps are defined by.Calculation C-..NSA-050.03-028.
Test procedures DB-SP-03151, Revision 14 and DB-SP-03160, Revision 16.contain a prerequisite step to obtain the appropriate curve (CC 14.89b, CC 14.89c, CC 14.89e or CC 14.89f) from DB-PF-06704 and verifythe data point obtained during the. surveillance test is in the acceptable range of the curve. Each of these curves references Calculation C-NSA-050.03-028. All actions complete for these pumps.Compone'nt Cooling Water Pumps: The acceptance criteria for these pumps are defined by.Calculation C-NSA-016.04-001.
Test procedures DB.-PF-03072, Revision 09, DB-PF-03073,.
Revision 11, DB-PF-03074, Revision 09 and'DB-PF-.03075, Revision 03 contain a prerequisite step to obtain the appropriate curve (CC 14.20k, CC 14.201, CC 1.4.20m, CC 14.20n, CC14.20o, or CC 14.20p) from DB-PF-06704 and verify the data point obtained during the surveillance test is in the acceptablerange of the curve. Each of these curves references Calculation C-NSA-016.04-001.
All actions complete for these pumps.*Decay Heat / Low Pressure Injection Pumps: The acceptance criteria for these pumps are defined.by Calculation C-NSA-049.02-033.
Test procedures DB-SP-03136, Revision 13 and DB-SP-031 37, Revision 13 contain a prerequisite step.to obtain .the appropriate curve (CC 14.34f, CC 14.34g, CC 14.34i or CC 14.34j) from DB-PF-06704 and verify the data point obtained during the surveillance.
test is in' the acceptable range of the curve. Each of these curves references Calculation C-NSA-'049.02-033.
All actions complete for these pumps. * 'Page 22 of 59 2 Site: G201 CORRECTIVE-ACTION CR Numlber: NOP.-LP-2001-05 0 i, :1 2.054:*'High Pressure Injection Pumps: The acceptance criteria for these pumps are defined.by Calculation.:
C-NSA-052.01-003.
Test procedures DB-PF-03207, Revision 06, DBRPF-03208;:
Revision'05, DB-SP-03218, Revision 12 and DB-SP-03219, Revision 12.contain a prerequisite step to obtain*the appropriate curve (CC 14.92b, CC 14.92d, CC 14.92f or CC 14.92h).from.DB-PF-06704 and verify the data point obtained during the surveillance test is in the acceptable range of the curve. Each of these curves references Calculation C-NSA-052.01.003.
All actions coniplete for;these pumps..Service Water Pumps: The acceptance criteria for these pumps are defined by Calculation C-NSA-011.01-016.
Test procedures DB-PF-03017, Revision 12, .DBI-PF-03023, Revision 14,.DB-PF-03030, Revision 11, DB-PF-03117, Revision 07, DB-PF-03123, Revision 10 and DB-PF!703130,:
Revision 09 contain a prerequisite step to obtain the appropriate curve (CC 14.73h, CC 14.73i, CC 14.73j, CC 14.73k CC 14.731 or CC 14.73m),from DB-PF-06704 and verifytthe data point obtained: during the surveillance test is. in the acceptable range of the Curve. Each of these curves references Calculation C-NSA-011.01-014 and Operability Evaluation 2003-032 as.the source, documents, but these should be replaced with Calculation C-NSA-0"-1I!01-01 6 which isa recently developed calculation.
Note that the acceptancecriteria curves are unaffected, only the document, reference on each curve page.This corrective action is complete with the exception of:the following items which are being.,converted to Activity Tracki g Items in SAP., 1) Delete reference to Calculation C-NSA-011.01-7014.and Operability Evaluation 2003-0032 on: the Service Water Pump acceptance criteria Curve pages:in Attachm6nt.60 of DB-PF-06704, Pump Performance Curves and add reference to Calculation C-NSAl.01.01_016.&#xfd; (SAP Activity Tracking Item 45912, NotificatiOn 600312949)
--2) Add-an administrative Limit and Precaution to procedure
:DBTSP-03371' Q::Quarterly Ma keup Pump 1 Inservice Testing and Inspection stating that.Calculation C-NSA-065,0.1-019 provides the* minimum pump performance acceptance criteria for. MakeupPump
: 1. Also, add this calculation to the reference section of DB-SP-03371.
Note: This item shouild.:be coordinated with Corrective
: Action 03-03603-11, which will update DB-S P-.03371 as a result of CaLcuIlati,'on C-NSA-065.01.-019.(SAP Activity Tracking Item 45904, Notification 600312947)" 3) Add an administrative Limit and Precaution to procedure DB-SP-03376, Quarterly Makeup Pump 2 Inservice Testing and Inspection stating that Calculation C-NSA-065,01-019 provides the minimum pump performance acceptance criteria for Makeup Pump 2. Also,-add this calculation to the reference section of DB-SP-03376.
Note: This item should be Coordinated with Corrective Action 03-03603-12, which will update DB-SP-03376 as a result of Calculation C-NSA-065.01-019.(SAP Activity Tracking Item 45907, Notification 60031294,8)
Alternate Corrective Action or Justification if Corrective Action not performed:
Corrective Action Implementation Date: 7/10/2006 JSignature indicates Corrective Action complete: Completed By: Date: 7/10/2006_J Signature indicates verification for SCAQ CRs: Verified By:, ..Date: J Enter Name and Sign: Implementing Organization Approval:
Date: 7/10/2006 Page 23 of 59 Sitp (~2O1':,
* CORRECTIVE ACTION'c ubr NOP.ILP-20011-05 02-04514 V oCmments:"'
U E -This Enhancement Action was to revise the existing IST pump testing procedures to include a A R .referehce to0the calculation(s) for the minimum design flow requirements, in order to help ensure the FLI procedures are updated.when the calculations are revised. The action was not implemented as T I originally stated,: but since this was an enhancement action, the action could be revised or deleted E with (at least).th e supervisor's concurrence.
The assqciatedmanager approved the implementation R of this action, so the deviatidnis acceptable.
The Implementation Response.provides .a list of the affected IST pumps and the associated testing" procedures.
Reviewed ISTP3 (Third Ten Year Inservice Testing Progranm, Revision 5, 4/13/06) and ISTB2 (Pump and Valve Basis Documeht, Volume it, Pump Basis, Revision 3, 5/24106).
Verified:that all ofthe pumps.listed in the lST program were discussed in the Implementation Response.The Implementation Response discusses the actions taken for each of the pump testing procedures.
For the Boric Acid Pumps, Containment Spray Pumps, Emergency Diesel Generator Fuel Oil Transfer Pumpns, and. the Makeup Pumps, verified the procedures listed referenced the associated calcUlations.
One typographical error was noted in the Implementation Response:
The calculation-
'for the EDG Fuel Oil Transfer.Pumps was list~d as calculation "016.05.".
This should be calculation.
'`016;015," Verified the associated procedures referenced the correct calculation.
For the remaining IST pumps (AFW Pumps; CCWPumps, DH/LPI.Pumps, HPI Pumps,; and SW.Pumps), the Implemerntation Response indicates that, instead of (or in addition to)-including the.calculation-in the individual test procedure, the calculation is referenced on the pump performance curves in'DB-PF-06704'and the test procedure directs the user to obtaintheappropriate curve from, DB-PF-:b6704., Verified DB-PF706704, ,Pump Performance Curves, contains the.Pump Acceptance Criteria Curves'(Attachment 6),which references the associated calculations for those pumps as indicated in the:..,Implementation Response..
Verified the associated, test. procedures direct the user to obtain the:" appropriate curves from DB-PF-06704..
The Implementation Response indicates some actions still remain:'-revise the calculation reference on the. SW pump curves in DB-PF-06704 from C-NSA-011.0.1-014 to C-NSA-011.01-016.
*Notification 600312949 was initiated to perform this update. Since the implementation of this action,'the changes have been made. Verified the calculation reference was updated in Revision..18 to- DB-PF-06704.
include reference' to the calculation in Makeup Pump tests DB-SP-03371 and DB-SP-03377.
Verified Notifications 600312947 and 600312948 were generated to accomplish these procedure..
enhancements.
This enhancement action has been adequately implemented.
* Doug Andrews DBOV, '8/10/06 Approval:
Date: 8/10/2006 Page 24 of 59 Site: `G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-04514 CR Category:
Action Type: Schedule Type: CANurne..
: AA ( ) (E) Refueling/Forced Outage .9.Corrective Action Type: Cause Code: RespOrg: (PR) Preventive Action (F04) ConrIg/design changes DES R
 
== Description:==
*
* ."- '" .'". ., ,:" ..."[: .'I NED provide IST Acceptance Criteria which defines the minimum design required system flow, G value(s) that will govern pump testing to PE for the AFW Pumps.N A The IST and System Engineer(s) recommends that this is a Mode 3 Restart action., R Completed By: Organ zation: Date: Phone: Attachments:
PE *,2/42003 7637. [0e vN If a Refueling Outage is required, Other Tracking # Corrective Action Due Date-, ACC- Enter the Refueling Outage number:..
13RFO NA 5/9/2003 n &#xa5; ' *~~........
".. .. ... ' " " ..... : I ., 1:-, ' ../1 20 .- .. ...:................
.... .... ... .EPT Approval: (Enter Name and'S! m):J con: II) te DUAL. Ouality Organization Approval:
Dt:..I
-,Completed as wftten K" Revised/Alternate Solutiion ) Not.Performed M Calculation C-NSA&#xfd;050.03-028,revision 0 provides Inservice Test-acceptance.criteria P ' pumps. This calculation which was approved on 5/8/2003 provides the following criteria':
L E AFW. pumps must be capable of delivering 857 gpm.at a pump differential pressure of 1"160psid.'.
E Alternate Corrective Action or. Justification if Corrective Action not performed:, N T N' Corrective Action Implementation Date:. 5/8/2003 G .. ........j Signature indicates Corrective Action compiete:.
Completed By: ".Date: 5/8/2003 R S Signature indicates verifilcation for SCAQ CRs: Verified By: Date: J Enter Name and Sign: Implementing Organization Approval:
---- Date: 5/8/2003 Q V Comments: U E Calculation C-NSA-050.03-028 Rev. 0 was approved 5/8/2003 to provide inservice test acceptance A R criteria for the Auxilliary Feed Water (AFW) pumps. The acceptance criteria defined the minimum L F required design system flow values.I F Y E R ". .....- , ......Approval:
%.Date: 6/19/2003 Page 25 of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05
.02-04514 CR Category:I Action Type: Schedule Type: "CANumber:
AA I). (A) Owner Assigned/Controlled 10 Corrective Action Type: .Cause Code: Resp Org: PR ) Preventive Action ( F04 ) Contig/design changes ...PES 0 ..... .... .. .. ..................
..R
 
== Description:==
 
.IST Engineer to create an administrative control within the IST Administrative procedure to govern G the changes to pump curves and notification of Design Engineering for support of evaluations and review to ensure the design basis is maintained.
A The IST Engineer.has indicated this.action is required Post Restart..0 R Completed By: Organization:
Date: Phone: Attachments:
MOGAHA, M PE 2/4/2003 7637 1ryes v'No If a Refueling Outage. is required, Other.Tracking#
Corrective Action-Duo Date: AC-Enter the Refueling Outage number: 1RO/A8/3/2003 ACC 13RF ..... ...... .. ........ ...... .-EPT :Approval: (Enter Name and Sign) Section: Date:-QUAL" ouality Organization Approval:
Date: fCG HA M5/9"'P ./200.3//20
.".63.
* i e ~ iN I Response, -. .Completedas wdfflen 1.) Revised/ltlernate Solution (~Not Performed M
* B-PF-00 201 Revision 05was. altered under PAT 03-0276 and we nt : effective on i 4/17/03. This P alteration includes information regarding preser-vic~e/baseline testing of pumps. Added a* L equremnt to update pupcrepocedure DB-PF 06704.when developing a new pum E .performance Curve or superCeding~an existing curve. Also,a aded-the :requiremeit to initiate a CR M 'to Design Engineering, iOa the absence of clearly defined .acceptance criteria, to evaluate the test&#xa3; data for impact on-the design~b as is.,.N, Alternate Corrective Action .or Justification
&#xfd;if Corrective Action not performed.
.....'. ..' ..' -..." ...... ...." .... .. ... ... .. .. ... .. ...........
.. .. ...... ..... .." T N Corrective Action implementation Date: 4/7/00 o jSignature indic-,ites Coriective Action comphlete:
.Completed By: Dat:, 4/17/2003 G A Signature indicates verification for SCAQ CRs: Verified By: upDate: j Enter Name and Sign: Implementing Organization Approval:
Date:~ 4/18/2003 o v Comments: U E Revision 5 of procedure DB-PF-00201 satisfactorily addresses the procedural requirements of A R changes to pump curves"and notification to Design Engineering for Support of evaluations. .S tep L E 6.6.8 addresses the Pump Curve Book and step 8.8.2.1 addresses the notification of Design T AlEngineering required by this corrective action. Verification concurred by:.Y E R ........ ....... .. ......"_______-__
_ ..'*,oApproval Al D ate: " i/251 3 .Page 26 of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-04514 CR Category:
Action Type: Schedule Type: A Number: AA ( ) (A) Owner Assigned/Controlled
* .1 Corrective Action Type: i Cause Code:. Resp Org: PR) Preventive Action (B06) Prog/process weak PES R
 
== Description:==
 
Add a prerequiste to the pump baseline tests for the.Service Water pumps to gather the following' G data at a point(s) near the maximum hydraulic horsepower:
(1) Head N (2) Flow A (3) Perform energized motor testing per DB-PF-05064 to obtain Motor Voltage and Current data T 0 Note: The PdMA report is recommended for voltage and current R The IST Engineer recommends this action be addressed Post Restart* If a Refueling Outage is required, Other Tra~cking
#.. C'orrective:Act.,ion.
Due Date ACC- .Enter the Refueling Outage number: NM N/ /112004 EPT Approval: (Enter Name and Sign) &#xfd;econ:. 77Date: PE S: 2/25./2003 QUAL Quality Organization Approval:
Date: ITY " .1 1/20/2003 peCompletedas written C) Revise*Alt.rnateSolution Q.L otp erformed.M The following IST Service Water Pump baseline testing procedures have. been altered to. include.f steps requiring motor. data to berobtainedduring therbasekine test. Steps have been'A ritten.suech L
* Ethat this data. will be obtained along with differential .pressure at each flow point. This was verified to EPT Abe incorporated into the following:-'
M E DB-PF-a3117 Revision 03, which went, effective on 3/29/03.T DB-PF-03123 Revision 05,.which went effective on 4/28/03.N DB-PF-03 130 Revision 05, which went effective on 4128/03.G The following IST Service Water Pump baseline testing procedures hav e been altered to include o *steps conditionally requiring motor data to be obtained during the baseline test. When. testing is R being performed as a result of pump/motor maintenance or when directed by Plant Engineering, G gathering motor data is required.
When motor-data is required, steps have been. written such that this data will be obtained along with differential pressure at each flow point. This was verified to be incorporated into the following:
." DB-PF-03217 Revision 00, which went effective on 9/3/03.DB-PF-03223 Revision 00, which went effective on 9/3/03.DB-PF-03232 Revision 00, which went effective on .9/3/03.Alternate Corrective Action or Justification if Corrective Action not performed:.
.....g .... .... " .The folowingIST.Sevice Wter Pu Pbageie tesin of cdue 59ebe'leedt nld Site: (G201*CO RRECTIVE ACTIO N .-,. CR Nurlber: NO P-L F Action or Justification if Correctiv.e Action not perform ed i: .-.&#xfd;- 04 : , 5 :.Corrective Action Implementation Date: 9/9/2003 j Signature indicates Corrective Action complete: Completed By: .. Date: 9/9/2003 j Signature indicates verification for SCAQ CRs: Verified By: _______, __"_______,.
.:bDate: J Enter Name and Sign: Implementing Organization Approval:
Date: 9/16/2003
.Q U A L T Y E R F-E R Comments: Reviewed the referenced procedures:
Procedures DB-PF-03217.
DB-PF-03223, and DB-PF-03232 contain the steps:to collect motor.data in conjunctionwith head and flow data as described.
Procedures DB-PF-031 17., DBPF-03123, and DB-PF-03130 contain the. steps.t0o collect motor data in-conjunction with head and flow data as described.
These three procedures
:are currently inhactiv'e.however verified that they were changed as. noted,.App roval. .Dae.. 1 0.. .. .." A p p.r1o! 9v3 )l.: D a te '. ../2 /2 6 .." : , " : " :: ..: ..: : ; " ' .i : " "* -":L: " .'Page 28 of 59 Sitp: ~ l CORRECTIVE ACTION cR Number: NOP-LP-2001-.05.
02-04514 CR Category:
Action Type' i Schedule Type: CA Number: (A) Owner Assigned/Controlled
* 12 ,Corrective Action Type: Resp Org: (PR Preventive Action [(806) Prog/process weak ... 'PES R
 
== Description:==
 
Add a prerequiste to the pump baseline tests for the Decay Heat pumps to gather the following data G at a point(s) near the maximum hydraulic horsepower:..-
(1) Head.N (2) Flow.A (3) Perform energized motor testing per DB-PF-05064 to obtain Motor Voltage and Current data*T Note: The PdMA report is recommended for voltage and current R The IST Engineer recommends this'action be addressed Post Restart Completed By: Organization:
Date: Phone: Attachments:
PES. 2/24/2003
.. 7637 j YeS v'bNo If a Refueling Outage is required, OtherTracking
# Corrective Act ionDue .Date: ACC- Enter the Refueling Outage number: N/A " 3/31/2004 EPT Approval: (Enter Name and Sign) Section: bate:.*ES 2125..QUAL- Quality Organization Approval:
Date: I Response:
.i Compketedas written Revised/Alternate Solution `CD Not Performed M The following Decay Heat Pump baseline testing procedures have been altered to include steps P conditionally requiring motor data to be obtainedduring the baseline test.. .When testing is being., L performed as a result of pump/motor maintenance or when directed by Plant. Engineering, gathering
.E motor data is required.
When motor data is required, steps have been written such that this data M will be obtained along with differential pressure at each flow point. This was verified to be, E incorporated into the following:
.N T DB-PF-03236 Revision 01, which went effective on 2/18/04.N DB-PF-03237 Revision 02, which went effective on.2/10/04..
G Alternate Corrective Action or Justification if Corrective Action not performed:
0 R G Corrective.
Action Implementation Date: 3/1/2004 J Signature indicates Corrective Action complete: Completed By: Date: 3/1/2004 J Signature indicates verification for SCAQ CRs: Verified By: , Date: J Enter Name and Sign: Implementing Organization Approval:
V797==. Date: 3/.1/2004 I Page 29 of 59 sitc,- r9n1 p ~CORRECTIVE, ACTION, :L:i.,'' ;N0P-LP-2001-05.
., ., .., ., iO " 4 1 Comments: U- E Reviewed the referenced, procedures:,'AR ILF PrIcedures DB-PF-03236 and DB-PF-03237 were revised to reflect steps to collect motor data that T Will be obtainedin conjunction with head and flow data as described.
'Administrative contiols have Y E been enhanced.
to adequately address the identifed concern of the decay heat pump baseline test.:'.R- No further actions are required.'
Approval:
.Date: 511412004 Page 30 of 59 Site: G201 CORRECTIVE ACTION NOP7LP-201-05::
02-0414.CR Category:
Action .Type: Schedule Type: ". CA Number: A) Owner Assigned/Controlled 13: Corrective Action Type: .Cause Code: RespOrg: O (PR) Preventive Action,. (06) Prog/prcess weak .DBTS o .........* *'............
* ....... ..SDescriptionV....
Add a prerequiste to the*pump baseline tests for the High Pressure Injection pumps to gather the G following data at a point(s) near the maximum hydraulic horsepower:
(1) Head.N (2) Flow A (3) Perform energized motor testing per DB-PF-05064 to obtain Motor Voltage and Currentcdata 0 :.Note: The PdMA report is recommended for voltage and current' The IST Engineer recommends this action be addressed Post Restart C m ltdB:Organization:
Date: Phone:','
Attachments::.
,,'I PES 2/24/2003 7637 YesNo If a Refueling Outage is required; Other -Tracking
# Corrective Action Due Dte: ACC. Enter the Refueling Outage number: N/A [ 7/26/2006,.:ERT :Approvali (Enter Name and Sign) ., .**Section:
Date: 4,T 2/25/2003 QUAL Quality Organization Approval:.
Date: ITY3/620 I Response:
.. Completed as writen ,. Re'ysed/Alternate Solution , Not'Performed
.The original HPI Pump baseline tests are DB-PF-04207 and DB-PF-04208-.
These procedures S.have been superseded by new procedures:DB-PF-03082 and DB-PF-03083:
DB-PF-03082 and DB-..L PF-03083 include'prerequisites and instructions for gathering the;.following data at maximum.flow.
E (1) Head, (2) F ow and(3) Motor data voltage and currenrt data.: Th is is incorporated into DB-PF-M 03082, Revision.00 and DB-PF-03083, Revision 00 whichboth went effective on 2/13/06.'E N , Additional HPI Pump baseline tests DB-PF-03407, Revision 07 and DB-PF-03408, Revision 04 also T IconItain similar prerequisites and instructions for gathering the following data at maximum fioW. .(1). .I 'Head, (2) Flow and(3) Motor data voltage and current data. Both revisions of these procedures N went effective on 6/2/05., G Alternate Corrective Action or Justification if Corrective Action not performed:
0 , G Corrective Action Implementation Date: '. 2/1512006 J Signature indicates Corrective Action complete: Completed By: __ , Date: 2/15/2006 j Signature indicates verification for SCAQ CRs*Verified By: .Date: " ..'_J Enter Name and Sign: Implementing Organization Approval:. " Date: .2/16/2006, Page9 31 of 59 Site: G201 CORRECTV ACTIONhe NOP-LP-2001-05 a v .Comments:
-.' ' ...U E Four procedures were reveiwed and found to contain appropriate steps for, data gathering-as, R required.T -',I,2/2416 Y E Approval:
-Date: 24,24/2006 Page 32 of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-20011-05
.."2.045,4 CR Category:.
Action Type. Schedule Type:. CA Number;AA ( A ),Owner Assigned/Controlled 14 Corrective Action Type: -' Cause Code: Resp rg: (PR) Preventive Action B(06) Prog/proce ssweak, .PES R
 
== Description:==
 
Add a prerequiste to the pump baseline.t6sts for the Component.Cooling water pumps to gather the G., following data at a point(s) near the maximum hydraulic horsepower:
(1) Head N (2) Flow A (3) Perform energized motor testing per DB-PF-05064 to obtain Motor Voltage and Current data T&#xfd;-0 Note:The PdMA report is recommended for voltage and Current........
R The lST Engineer recommends .thiis ,action be.addressed.PostJRestart Completed By: Organizatioae:
Phone: Attachments:
P: : 2/24/2003**.
7637 *: ., ' Yes 1 No : If aRefueling Outage is required, Other Tracking:#
Corrective Action Due.Date: AC-' Enter the Refueling Outage number: N/A N/A .8/23/2003 EPT App:roval: (Enter Name and Sign) : .Section::
.Date:: PES, 2/25/2003
: QUAL Quality Organization Approval:
: Date: TY ....612/2003 I Response:
t% C~ompteted as wi*ten (Re~vis&.d1Altrna te Solution-Not Performed M The following IST Component Cooling Water'Pumlp baseline testing procedures have been:created P. and include steps requiring motor data toebe obtained during the baseline test. Steps.have been L written such that this datawill be obtained along with differential pressure at.each flow point.E ..: MA DB-PF-03572 Revision 00 was created under PAT 01-0426, whichtwent effective on 5/22/03.E N DB-PF-03573 Revision,00 was created under PAT 01-0488, which went effective on 5/22/03."T I DB-PF-03574 Revision 00 was created under PAT 01-0489, which went effective on 5/22/03.N Alternate Corrective Action or Justification if Corrective Action not performed:
0 R S .Corrective Actiorn Implementation Date: 5/22/2003"G ~ ~ ~~~ ., .- ,,- *. , .... ..... ............
:. .. .......j Signature indicates Corrective Action complete:
... Dae 520 Completed By: Date: 5/22_2003_j Signature indicates verification for SCAQ CRs:.Verified By: "___'_"_Date:.j Ente~r Namne arid Sign: Implementing Organization Approval:.
-Date: 5/2312003 Page 33 of&#xfd;59 Site: G201 CORRECTIVE ACTION CR Number;NOP-LP-2001i-05 02-04514 o V: Comments: U E Component Coolong Water (CCW) pump baseline test procedures, DB-PF-03572 Rev. 00 Steps A R 4&#xfd;21 and 4.22, DB-PF-03573 Rev. 00 Steps 4.21 and 4.22 and DB-PF-03574 Rev. 00 Steps 4.2.24 L I and 4.2.25, were revised to satisfactorily address the corrective action. The corrective action is I F T I considered closed. Concurred by -Engineering NQA.Y E Approval:
U -' Date: 6/12/2003 Page 34 of 59 Site: G201 CORRECTIVE ACTION 'CR.Number NOP-LP-2001-05 02-04514 CR Category:
Action Type: Schedule Type:., CA Number: A I ( .(A).Owner Assin dC nr0ld " ". 5i Corrective Action Type: Cause Code: Resp Org: (PR) Pieventive Action ) rocess weak' * .'Description:
Add a prerequiste to the, pump baseline tests for the Makeup pumps to gather the following data at a G point(s) near the maximum hydraulic horsepower:'
t (1) Head N (2) Flow A (3) Perform energized motor testing per DB-PF-05064 to obtain Motor Voltage-and Current data T .Note:hePdM o Note: ThePdMA report is recommended for voltage and Current R The !ST Engineer recommends this action be addressed.Post Restart Completed By: " Organization:
Date: Phone: Attachments:
PES 2/24/2003 7637 .Li Yes v'.No: If a Refueling Outage Is requiredi Other Tracking # Corrective Action.Due Date: ACC- Enter the Refueling Outage number N N/A .. 3020. '......... ..... ....... ---- ........ ... .. .. ... .... .... 6. .... .0 ...... ... .......EPT Approval: (Enter Name and Sign) Section: Date: QuAL Quality Organization Approval:
Date: ITY 3/16/2004
, I Response: Completed aswrlten.
.. " Revised/Alternate solution Q'C) Not Performed M This corrective action as written requires a prerequisite tbecreated to the pump baseline tests:for P the Makeup Pumps to gather the following data at a:point(s) near the'rmaxim um hydraulic*L..horsepower:
(1). Head, (2) Flow, (3).Perform energized motor testing per DB-PF-05064 to obtain, E Motor Voltage and Current data. Note: The PdMA report is recommended for voltage and current.M E This corrective action is being completed as written, with. minorchanges in how the actions were N implemented that do not change the scope or intent as allowed by NOP-LP-2001, Revision 10 Note, T .preceding substep 1 of Step 4.13.4. The scope and intent ofthis corrective action is to gather I motor data during pump baseline testing, following pump maintenance that could affect pump N performance, near the point of maximum hydraulic horsepower.
G.Baseline test procedures have been developed for, each Makeup Pump: 0 R DB-PF-03472 for Makeup Pump 1 G DB-PF-03477.
for Makeup Pump. 2 Each of these procedures contains a prerequisite Step 3..4 to verify an Order exists if motor data is to be gathered per DB-PF-05064, Step 3.1 4 is conditional since it is possible that this procedure-may be performed for reasons other than pump maintenance, which would not require gathering motor data.. For example, during 13RFO, baseline tests were performed for the Service Water Pumps, which gathered motor data. Afterwards, baseline tests were performed again to gather pump hydrualic data using different flow instruments.
The motor data had already been gathered and was not needed during the second baseline tests. Motor data will be gathered during baseline testing following pump maintenance that could affect the pump performance as recommended in the Post Maintenance Testing Manual, Revision 26. The baseline test procedure, and PdMA motor Page 35 of 59 Site: G201 CORRECTIVEACTION.
cR Number: NO P-L"P-2001-05
'.02-04514 testing will be specifiedas post-maintenance testing .in the associated Order performing the maintenance.
When specified, motordata is gathered per.DB-PF-05064 at each point where pump hydraulic (head and flow') data are gathered including a point near the maximum hydraulic , "horsepower.
.: .: This fulfills the intent of this corrective:action, to gather energized motor data (along with pump hydraulic data. near a point of maximum-hhdraulic horsepower) during pump baseline testing as a result of pump maintenance that may affect the hydraulic performance of the pump.This is incorporated into DB-PF-03472, Revision.00 and.DB-PF-03477, Revision 00, which went effective on 5/10/05.-":Alternate Corrective Action orJustification if .Corrective Action not performed:
...... ..... ... .. ... .... ... .. .. .......... .... ........ ..........Corrective Action Implementation Date: 5/10/2005 J Signature indicatesCorreciive Compileted By: __._._, Date:, 5/10/2005* ' Sigrnatureindicates verification for SCAC CRs: Verified By: D ......... .... D.te...Enter, Name~ and Sign:-Implementinrg Organization Approval:
*.. " EM Date: 5/10/2005 Q V Comments:*u, E This Preventive Action was to "add a prerequiste to the pump baseline tests for the Makeup pumps'.AR tlo-gather-the following data'at a point(s)*near the maximum hydraulic.horsepower:
I.". (1) Head.T .(2).Flow.yv E (3) Perform energized motor testing per DB-PF-05064 to obtain Motor Voltage and Current data." R .Note: The PdMA report is recommended for voltage and Current." The Implementation of this corrective action states that it is being completed as written, with minor, changes in how the actions were.implemented that do notchange the scope or intent as allowed-by..
NOP-LP-2001, Revision 1.0 Note, preceding substep 1 of Step 4.13.4. The scope and intent of this corrective.
action is to gather motor data during pump baseline.testing, following pump maintenance
'.*that could affect pump performance, near the point of maximum hydraulic horsepower.
The Implementation further states that baseline test procedures DB-PF-03472 (Makeup Pump #1) and DB-PF-03477 (Makeup Pump #2) have been developed, effective 5/10/05, which contain a prerequisite Step 3.1.4 to verify an Order exists if motor data is to be gathered per DB-PF-05064.
That step is conditional since it is possible that this procedure may be performed for reasons other than pump maintenance, which would not require gathering motor data.The Implementation response indicates motor data will be gathered during baseline, testing following pump maintenance that could affect the pump performance as recommended in the Post Maintenance Testing Manual, Revision 26. The baseline test procedure, and PdMA motor testing will be specified as post-maintenance testing in the associated Order performing the maintenance'..When specified, motor data is gathered per DB-PF-05064 at each point where pump hydraulic (head and flow) data are gathered, including a point near the maximum hydraulic horsepower.
Verified the Post Maintenance Test Manual (PMTM) includes the two new baseline test procedures.
The PMTM testing specified for pump assembly/repair/replacement includes an "Other" test per Note 14, which states: Page 36 of 59 Site: G201 CORRECTIVE AC IO C... .....NOP-LP-2001-05
". ..02-0145.14
"'Pump baseline testing is required when changing impeller or wear rings, adjusting the governor, or other maintenance the [sic] could affect the hydraulic performance or rotating speed of the pump,.DB-PF-05064 should be performed when baseline testing electrically-driven pumps." Reviewed new procedures DB-PF-03472, Makeup Pump 1 Baseline Test, and DB-PF-03477, Makeup Pump 2 Baseline Test. Verified the'tests were effective 5/10/05. *Verified each, has a -conditional step 3.1.4 stating that IF PdMA motor data is to be taken for the pump, THEN verify an Order has been generated to perfor~m the applicable sections of DB-PF-!05064, Electrical Machine' Testing Using PdMAMotor.Tester, concurrently with this test AND record the Order Number. .The tests also have a step,4.3 which states IF PdMA motor data .is to'takenfor the Makeup Pump, THEN verify personnel are stationed to obtain data in accordance with DB-PF-05064.
Verified the baseline test procedures'take data (pump head and pump..flow) at50 gpm, 100.gpm :* 150 gpm, 200 gpm, 250 gpm, 285 gpm, and. 325gp'm. At each of these readings, a. step is included.to'take motor data (if motor data is to be taken). These readings include data near the maximum%hydraulic horsepower (per curves in DB-PF-06704, Pump Performance.
Curves).*Concur that the implementation response fulfills the intent of this Corrective Action to gather energized motor data, along with pump hydraulic data, near a point. ofmaximum hydraulic:..
* horsepower during pump baseline testing as;a result of pump' maintenance that may affect the hydraulic performance of:the pump.This corrective action' has been adequately implemented..
DBOV, 5/23/05 Approval:
Date: .5/23/2005 Page 37 of 59 Site: G201 CORRECTIVE ACTIONU CR Number: NOP-LP-2001-05 02-04514" ' ' ' Scedule-Type:
L .' "-: .- ANu br CR Category:
Action Type .CA Number: AA ( ) " ' "(A) Owner ,,ssignedfControlled 16 Corrective Action Type: Cause.Code:
RespOrg: (PR ) Preventive Action B ' 06) Prog/process weak, .DBTS 0 ..... .. ...................
....... .... ... .j..,........
...............
: " :.............
..." R
 
== Description:==
.
Add a prerequiste to the pump baseline tests for the Containment Spray pumpst gather the G following data at a point(s) near the maximum hydrauliC.
horsepower.:.." (1) Head N (2) Flow A (3) Perform energized motor testing per DB-PF-&#xfd;05064 to obtain MotorVoltage and Current data T Noe Th ed. fo volag andL qurrent''.' " .Note: The PdMA report is recommend.fi .rrent:.R The IST Engineer recommends this action be addressed:
Post Restart:......... .. .. .. .... .... ..." i ......Completed By:' rgnization-Date- Phone: Attahments:.
PES: 2124/ 200 3 73F..YeS No If,,aRefueling Outage is required, .OtherTracking#
Correctve Action Due-Date: ACC Enter the RefUeling Outage number: N/A_ 1- /26/20 106 EPT Name and Sign) ., .." Section: Date: DBTS 2125/2003 QUAL 'Qua&#xfd;ity Orgnization Approval:
Date: I.TY I .3!16/2004.
, M., L E M E'N T N G 0 R G Response:
.(..Comp4etedaswdtten Revised/Alternate.Solu tion NotPerformted DB-PF 03437, Containment Spray Pump :! Baseline,3Test, Rev.0 and DB-PF-03438 Containment Spray Pump 2 Baseline Test,:Rev 0, obtain the following data at max flow: (1) Head (2) Flow (3) Motor Voltage and Current-Both Rev 0 procedures were made effective 2/2106.::Alternate Corrective Action or.Justification if Corrective Action not performed:
...*J Signature indicates Corredtive Action complete: Completed By: ective Action Implementation Date: 3/1/2006 Date: 2121/2006 j Signaturke indicates verification for SCAQ CRs: Verified By: Date: ,J Enter Name and Sign: Implementing Organization Approval: Date: 3/1/2006 Page 38 ot 59 Site: G201 CORRECTIVEACTI.ON , ... ....ber NOP-LP-2001-05 0-41 a v Comments: U. E Verified DB-PF-03437, Containment Spray Pump 1 Baseline Test,..Rev
: 0. and DB-PF-03438, A R Containment Spray Pump 2 Baseline .2, Rev 0, were made effective,2/21/06.
L I I F T I Verfied steps were added.toeach procedure to gather head (pump differential pressu.re), flow, and Y FE motor data (per'DB-PF-05064) at various flow rates, including maximumjflow (approximately 1700 R gpm). Prerequisite (pre-test brief) indicates this data Will be taken.This corrective action was adequately implemented.,'
.. ..... ..... ........ ... .. .... .........DBOV, 3/21/06Date:
3 .12 ......A
..........
..............
....... :................
..........
: .....ate.. ..-:-- : .3/21/2006
:-Page 39 of 59 Site: G201 CORRECTIVE ACTION .Nme NOP-LP-2001-05
.02-04514 CR Category:
ActionType:
Schedule Type: CA Number: AA LL? I (E )Refueling/Forced*Outage 17 Corrective Action Type: Cause Code: Resp Org:" (PR ) Preventive Action (F04) Configldesign changes DES, R
 
== Description:==
 
I Design Electrical to provide Instrument Uncertainty values to Nuclear Design and Mechanical Design.G for calculation input for IST Acceptance Criteria for the following pumps, as required.N Service Water Pumps A Decay Heat Pumps (LPI)T High Pressure Injection Pumps.0 Component Cooling Water Pumps R Containment Spray Pumps The IST. Engineer has determined that this action should be addressed prior to Mode 4'..-Completed.By:
Organization:
Date:: Phone: Attachments:"..
PES W//2003 7637 H es % No If a Refueling Outage Is required, Other Tracking # Corrective Action Due Date:, ACC- Enter the Refueling Outage number: 13RFO N/A 5/29/2003
" EP Approval: (Enter Na-me and Sign) Section: Date:* ." " ' " " ' ' ' " .. .'" ' ....""- DES": M / i:..:: 3820(03 ., .QUAL, Quality Organization Approval:.ITY ,. .' .. .* ".' ,: ..." ...6t16f2003.
-. " Ii :.Response:.
CompIeted as written
* Revised/Alternate Solution .Not'Performed M. The following calculations were completed for each of the systems:: L 'SW -C-ICE-01 1.01-002, Revision.0, 5/6/03.E DH -C-IC'E-049.02-003, Revision 0, dated 5/9/03 M HPI- C-ICE-052.01-001,.Revision 0, dated 4/30/03.E CCW-C-ICE-016.03-002, Revision 0, dated 4/30/03:N CS -C-ICE-052.01-001, Revision 0, dated 5/29/03 T Alternate Corrective Action or Justification if Corrective Action not performed:
.... .. .. .... ... ... .......... ..............N o Corrective Action Implementation Date: 5/29/2003 R J. Signature indicates Corrective Action complete: G Completed By: Date: 5/29/2003 J Signature indicates verification for SCAQ CRs: Verified By: Date:.J Enter Nme and Sign: Implementing Organization Approval: .Date: 59/2003 Page 40 of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-04514.a v Comments: u E Uncertainty calculations to resolve this corrective action have been issued as follows: A R 1. Service Water: C-ICE-01 1.01 -002 Rev. 0 2. Decay Heat: C-ICE-049.02-003 Rev. 0 I F T .3. High Pressure Injection:
C-ICE-052.01-001 Rev. 0 Y E 4. Component Cooling Water: C-ICE-016.03-002 Rev, 0 R 5. Containment Spray: C-ICE-061-001-002 Rev.0 The implementation section of the condition'report listed the containment spray uncertainty calculation as C.-ICE-052.01-001 Rev. 0. This calculation number was misreferenced.
The correct calculation number was verified to be C-ICE-061.01-002 Rev. 0. Concurred by .NQA Eng._____Apoa D-- 6/16/ 20 3 ..Approval: -b "" :ate: '.6116/2003.
Page 41 of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-04514.CRCategory:
Action Type: ,Schedule Type: CA Number: AA ( ) ___ (E) Refueling/Forced Outage .8 Corrective Action Type: Cause Code: Res p Org"* ' Res&#xfd;p O0g 9 (PR) Preventive Action .F04 Config/design changes' DES, R
 
== Description:==
 
Design Electrical to provide Instrument Uncertainty values to Nuclear Design and Mechanical Design G for calculation input for IST Acceptance Criteria.for the AFW pumps, as required.N" A The IST Engineer has determined that this action should.be addressed prior to Mode 3.*T 0 R.O .. .*" , ...........
.. .... ..........
Completed By: Organization:
.Date:. Phone: " Attachments.
IPES 3/7/2003 I. 7637 .[ Yes No If a Refueling Outage is required, Other Track.ng # Corrective Action Due Date: ACC- Enter the Refueling Outage number: :13RFO .N/A 5/7/2003 SApproval: (Enter Name and Sign) ;: Section:&#xfd; Date:'DES -3/8/2003 QUAL. Quality Organization Approval:
Date: ITY *6/12/2003
* Response:
-Completed as writr.en "Revised/Alternate Solution Not Performed ,M The- following calculation was completed for AFW instrument Uncertainty:
P .L C-ICE-050.03-002, Revision 0, issued 517/03. Copy supplied to ADUland MDu.E M E Alternate Corrective Action or Justification if Corrective Action not performed:
N T .N Corrective Action Implementation Date: 5/7/2003 J Signature indicates Corrective Action complete: O 'Completed By: __.__*_._Date:
517/2003 R j Signature indicates verification for SCAD CRs:*G VerifiedBy:
Date: J Enter Name and Sign-Implementing Organization Approval:
-Date: 5/7/2003 Q v Comments: u E Calculation C-ICE-050.03-002 Revision 0 issued 05/07/03 satisfactorily addresses the Auxilliary A R Feed Water (AFW) instrument uncertainty.:
This corrective action is acceptable and is considered L I closed. Corrective action response concurred by I F Engineering NQA.T I Y E Approval:'
.'. Date: 6/12/2003 Page 42.of 59 G201 Site: G201'CORRECTIVE ACTION CR Number: NOP-LP-2006-05' 02-04514.CR Category:
Acti6n Type: Schedule Type: .CA Number: AA .() .-(E) Refueling/Forced Outage I 19'Corrective Action Type: ... Cause Code: , "' ..Resp Org: (PR) Preventive Action. (:F04) Config/design changes -DES"'R
 
== Description:==
 
Design Electrical to provide Motor Frequency (RPM) values to Nuclear Design, Mechanical Design G and Plant Engineering for calculation input for IST Acceptance Criteria for thefollowing pumps, as required.N A Service Water.Pumps (Nuclear), T Decay Heat Pumps (LPI) (Nuclear).
0o High Pressure Injection.
Pumps (Nuclear)R Component Cooling Water Pumps. (Nuclear).
Containment Spray Pumps (Mechanical)
Makeup Pumps (Test and Performance)
The IST Engineerhas determined that this action should be addressed prior to Mode 4.* I--Comprleed By: .. .'' Olanization:'i Date: " i Phone .&#xfd;ttachments:-
PES. 3f7/2003 7637 ._ Yes L_ No*If a Refueling Outage is reurd ".. .Other tracking # 9 Corrective Action Due Date: , ACC- Enter the Refueling Outagelnumber:.
1 3RFO N/A V54/2003 EPT Approval: (Enter Name and Sign) Section- Date.DES 3/8/2003 QUAL Quality Organization Approval:
..Date: IT -.6/12/2003 Response: (I. Completed as writen Revised/Alternate Solution. ( Not Performed M The motor frequency information Was evaluated in CR 03-00866.
Theresult of this evaluation is'. that the frequency limitations are 59.5 Hz & 60.5 Hz.L E M E Alternate Corrective Action or Justification if Corrective Action not performed:
N T N Corrective Action Implementation Date: 5/1/2003.G ._ __ _S signature indicates Corrective Action complete:.Completed By: .Date: 5/1/2003.R J Signature indicates verification for SCAQ CRs: G Verified By: .Date: J Enter Name and Sign: Implementing Organization Approval:
Date: 5/1/2003 Q V Comments:.
U E Condition Report CR.03-00866 documents the evalution of the motor frequency.
The cause A R analysis of condition report CR'03-00866 appropriately describes the motor frequency in the range L I of 59.5 to 60.5 Hz. This corrective action is considered closed. Concurred by Eng.I F T I NQA.Y E R .............
.........Approval:
gDate: 6/12/2003 Page 43 of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-04514 CR Category:
Action Type: 1 Schedule Type: CA Number: AA ( ) (E) Refueling/Forced Outage 20 Corrective Action Type: Cause Code: Resp Org: (PR) Preventive Action (F04 ) Config/design changes * ...... ..DES.R
 
== Description:==
, 1 MechanicalDesign to provide IST Acceptance Criteria which defines the' minimum design required G system flow value(s), that will govern pump testing for the Containment Spray Pumps to PES.N The IST Engineer considers this action to be required for Mode 4.A T 0 R Competed By. .. .Organization:
Date: Phone: Attachments:
1PES 3/72003 7637 !'Yes : No6 If a Refueling Outage is required, Other Tracking # Corrective Action DueDate: ACC, Enter the Refueling Outage number: 13RF N/A .5/2/2003.EPT Approval: (Enter Name and Sign) Section: Date: SDES 3/9/2003 QUAL Quality Organization Approval: , Date" ITY ." .6/16/2003.I ... .Response
-. .... ..C-om&#xfd; pletedas-w ten- ....... Revised/Alternate Solution ..Not Performed.
M Bechtel Calculation37.
11, dated 3/5/86, calculated the proper setting for Containment Spray Auto P Control Valves CS1 530 and CS1531 to adjust the Containment Spray System resistance to provide L. 1300 gpmsystem design flow with a Containment Spray (CS) Pump_ performance.degradation of -E 10%. Calculation 37.11 calculated a setting of 60%6open for CS1530/1531at a required head'acroSs.
M the CS pump of 369.7 ft to provide a flow of 1300 gpm to the spray header. A TDH of'369.7'ft is E *equivalent to 160.1 psid across the CS pump. Calculation 37.11 Was prepared for CS Pump P56-2 N .and its associated piping which Supply flow to the highest elevation containment spray. ring header, T therefore the calculation is conservative for CS Pump. P56-1 and its associated piping, which is 20'-, I 2" below the upper spray ring header. The certified pump curves for P56-i and P56-2 show P56-1 N to be the slightly stronger pump, as well.G In order to establish minimum CS pump IST differential pressure acceptance criteria, it is necessary O to combine, information from calculation 37.11 for system resistance at a 60% open setting on R valves CS1530/1531, information from calculation C-ME-61.01-078 for instrument error, and G information from calculation 37.08 for CS system conditions during recirculation phase.The 0.5% accuracy of the test instruments specified to be used in DB-SP-03337 and DB-SP-03338 are well within the 2% required instrument accuracy specified in the 1995 ASME OM Code.Including an allowance for instrument accuracy in the acceptance criteria calculation is not required by the ASME OM Code, but has been included for the CS Pumps as additional margin.Using the CS system resistance curve for a setting of 60% on CS153011531, a flow Of 1325 gpm ,(the maximum flow. allowed during performance of the IST), and an allowance of 1.7% uncertainty from .C-ME-61.01-078, a minimum requireddifferential pressure across theCS pump of 165.8 psid is required.
The present CS pump' minimum differential pressure acceptance criteria of 166 psid contained in IST Procedures DB-SP-03337 and 03338 is therefore acceptable.
Page 44 of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-04514 Another Condition Report; CR 02-09150,,has been Written to address the fact that the CS pump IST acceptance criteria are not current and Corrective Action #1 .ofCR 02-09150 requires preparation of a calculation to establish minimum CS pump differential pressur e at 10%. pump degradation, including'instrument uncertainties.
Alternate'Corrective Action or Justification if Corrective Action: not performed:
.. .............
i .... ..... .... .Corrective Action Implementation.
Date: 4/10/2003 j Signature indicates Corrective Actibn cbriplete:
Completed By:. ....:. Date: 4/10/2003 J Signature indicates verification for SCAQ CRs: Verified By: .Date:..... J ..ne ~ a e .n. i n .... ........::. ....... .!.. .... ... ...... ....... ...., jEnter Name and Sign: Implementing Organization Approval:
Date:_, 5/2/2003 Q" V Comments:
i u E The implementing organization provided informationfor the lST program based on Bechtel A R' 'Calculations 37.11, 37.08 and Davis-Bessgecalculation These calculations were LI reviewed and determined to b1e acceptable.
I F TI Y E Since.the-time of issuance of this condition report a new calculation has been developed.
This R calculation reflects different results than of the ones cited above. :Condition report.03-03751 was issued to-document this issue... QA Will request a review of this condition repbrt to ensure that the input for the lST requirerrents:are addi~essed
.This corrective action is closed. The issue identified as a result of the new. calculation will be tracked by CR 03-037511.
... ..................
.. ....... .........
... ....... ........ ....-..., -Approval:
."Date: 6/1612003 Page 45 of 59' Sit~ (2fl1 CORRECTIVE ACTION CR Number: NOP-LP-20M1-05 02-04514 CRCategory:
Action Type: Schedule Type: CA Number:.AA ' )_"_ j(A) Owner Assigned/Controlled 21 66Cor'r'ective Action Type: Cause Code: Resp Org: ( ES ) Evaluation Support N NA) Not a Deficiency PES R
 
== Description:==
 
The MRB reviewed this CR on 3/8/03 to approve the transition from a Basic Cause to an Apparent G Cause to support the requirements of NOP-LP-2001R4.
Because the MRB approved a new Category/Evaluation method for this CR after 311/03, this CR is no longer exempt from the N requirements of NOP-LP-2001 R4.A T 0 R...... ... ........Completed By: Organization:
Date: Phone: Attachments."P1 3/13/2003 I 8284- ' -Yes- No If a Refueling, Outage is required, Other Tracking # Corrective Action Due Date: Acc Refueling e number:
* A .N/A 3/2112003.........
...........
..............
fg ~ i : " -.......................................
..................................
e.t.o ...... ... .a e EPT 7Approval: (Enter Name and Sign). Section Date:-*.PES ' " 3/18/2003 QUAL Quality Organization Approval:
Date: tTY ". .*. .4/4/2003 S-'Response:
..... K ,Completed as written .Revised/Alternate Solution .. :Not Perfrormed M The Cause Analysis contains the attrubutes of an ApparentCause Analysis as specified by Revision..P 4obf:NOP-LP-2001.
L'E M E -Alternate Corrective Action or Justification if Corrective Action not performed:
N T N Corrective Action Implementation Date: 3/20/2003 G ......... ..... ..........
.... ....J Signature indicates Corrective Action complete: o comnpleted By: _____&#xfd; ____ Date: 3/20/2003 R j Signature indicates verification for SCAQ CRs: G Verified By: -------- Date: J Enter Name and Sign: Implementing Organization Approval:
9 Date: 3120/2003 Q V Comments: U E The Cause Analysis of CR 02-04514 meets the requiremen-ts prescribed by.Attachment 6 of NOP-A R LP-2001 Rev. 4 for an Apparent Cause.L I I' .F T I Y E Approval:
Date: 4/4/2003 Page 46 of 59 Site: G201 CORRECTIVE ACTION: CRNumber: NOP-LP-2001-05
.02-04514 CR Category:
Action Type: Schedule Type: CA Number: AA ( ) (A) Owner Assign*ed/ControlIed 22.'....Corrective Action Type: Cause Code:,. Resp0rg: o CA) Corrective Action, (804) Design analysis P" " R
 
== Description:==
 
Address the QA rejection concerns of CR 02-04514 in CR 03-03603.N A T 0 R Completed By: Organization:
Date n Attachments PE-S 6/27/2003 7756 Y N, If a Refueling Outage is: required, other Tracking ~ orrective Action Due Date: AC- Enter the Refueling Outage number: _N/A:NA63020
:_- ,, 7.-EPT Approval: (Ent er Namne and-Sign)
.eton: Date:&#xfd;-PES 6/27/2003 QUAL Quality O rganization Approval:
..Date::*1Y ....: ....7/2/2003 I Response:
Completed as ywftfn Performed M The QA rejection concerns of CR -04514 have be en ad dressed in CR 03-03603.P E M E Alternate Corrective Action or Justification if Corrective Action not performed:
.... .N T SCorrective Action Implementation Date: .6/30/2003 j Signature indicates Corrective Action complete: SCompleted By: -Date: 6/30/2003 R j Signature indicates verification for SCAQ CRs: 6 Verified By: _________Date:
jEnter Name and Sign: Implementing Organization Approval:
.Date: 6/30/2003 Q. v Comments: U E Verified that the resolution for Condition Report CR-02-04514 have been captured in CR 03-03603.A P This action satisfactorily resolves this corrective action. This corrective aciton is closed. Concurred IL Iby C:&#xfd;: -NQA Engineering.
TI Y E Approval:
Date: '7/2/2003 Page 47 of 59 Site: G201 CORRECTIVE ACTION TR Number: NOP-LP-20011-05.
02-04514 CR Categ e: Schedule Type: CA Number: AA () (A) Owner Assigned/Controlled 23 Corrective-Action Type: Cause Code: Resp Org:;0 (ES) Evaluation Support... (NA) Not a Deficiency
.PES R
 
== Description:==
 
Please address the comments from the 8/26/03 CARB.G 1-Contact.the Engineering Training Review Committee to review the ISP job and determine what training and qualifications are appropriate.
N 2-Perform an Experience Review.A T 0 R-...... ...; .. -----" '- 'CompletedBy.1 Organization Date: Phone: Attachments:
P1 8/28/2003 8590 Yes No If a Refueling Outage is required, Other Tracking # Corrective Action Due Date: ACC- Enter the Refueling Outage number: N/A ' _ _ 10/8/2003 EPT -~Section:
Dt: Aproval: (Enter Name and Sign)P PES 8128/2003 OUALI Quality.Organion Approval:
Date: ITY 9/24/2003:.--. .written .(j Revised/Alternate Solution -... Not Performed.M Experience review completed 09/16/03.P. Contacted training section (T&#xfd; Simonetti/T.
Tackett).
Training section will conduct a job and task L analysis involving the supervisor and appropriate program owne&#xfd;(s)-
CA 24 initiated:for completion E of job and task analysis.M EL Alternate Corrective Action or Justification if Corrective Action not performed:
N -: -....--:: -- -- -: --: ........ .: -.....~ ~~~... ... ... -..........
..........
.... .................
.. ...........
...... .. ....... .--- ....... .........
----------------
...............
..- .. .... .----: ..N T N Corrective Action Implementation Date: 9/18/2003 G 1J Signature indicates Corrective Action complete: o Completed By: Date.. 9/18!2003 R J Signature indicates verification for SCAQ CRs: G Verified By: .Date:.J Enter Name and Sign: Implementing Organization Approval:
Date: 9/18/2003 Q V Comments: u E The implementing organization information indicated that the Experience Review and an A R Engineering Training Committee review of the IST position be rolled into a new corrective action, CA L I '24. An Experience Review was performed on 9/16/03. The Experience Review, included in the I F T. Generic Implications of CR 02-04514, was reviewed and determined to be acceptable.
As a result y E of the Generic Implications review, condition report CR 03-07765 was initiated to expand the original R issue identified by CR 02-04514 to other functional areas other than iST. Concurred by![ lL .Ndclear Quality Assurance.
Page 48 of 59 Site: G201 CORRECTIVE ACINCR N~umber:.'.
* -02i-0451.1' Approval: "Dale 9/24/2003", CR Category:
Action Type: Schedule Type: CA Number:'AA ( ) (A Owner Assigned/Controlled 24 Corrective Action Type: Cause Code: Resp Org: CA ) Corrective Action (F04 ) Config/design changes TRAN.0 ..,L : ..J ......'...,. .. -..L.;.:.. .. ..Description:
Conduct a job and task analysis of the IST Pump engineers position to identify required qualifications G and training.N A-T 0 R N .. ."/ .... .. ......* If a Refueling Outageis requ ired, Oth er Tra Icking # Corrective Action Due Date':..ACC_1 Enter the Refueling Outag e number: N/ A N/ 41620 N/ 4/161200 EPT Approval: (Enter Name and Sign) ion: Date: TRA.N 9/117/2003&#xfd; QUAL- uality Organization Approval.
Date: ITY 3/16/2004.
.*.Response:
'C" Completed 0s writen. Q'D R~evised/Alte~rnateS~oluition i Not Performed M .Conducted tabletop anailysis of the position, then developed and implemente .d Re vision 0 ofJ JFG'P ESI-1 35, ASME Section XI lnsrviceTest (ST) Pump and Valve Program- Owner. This JG..was..L- completed on 3/18/2004.
E .[.M Reviewed by TGS E Alternate Corrective Action or Justification if Corrective Action not performed:
N ----.. ... ...T N G Corrective Action Implementation Date 3/29/2004 Signature indicates Corrective Action complete: 0 Completed By: M____________
Date: 3/24/2004 R _J Signature indicates verification for SCAQ CRs: G Verified By: Date: T Enter Name and Sign: Implementing Organization Approval:
Date:. 3/29/2004 O v Comments: U E Reviuewed the task analysis paperwork and Revision 0 of JFG ESIme-1e35, Inservice Test (IST), Pump A R and Valve Program Owner. Verified the JFG was approved forFuse on 3/17/04.L I I F Y E NQA,'5/24/04 R ................. .op e e ..... 3/ 8 2 0 4 ... .... .... .... ". ..... ... ....... .. .. .:... ... ......M__.Reiewd
_b TGS____ ......7~Apprnvil-M_7:77 flatp* R/11/2004 Page 49 ot 59 Site: G201 CORRECTIVE ACTIONC NOP-LP-2001-05 02-04514 CR Category:.
Action Type: Z. chedule Type: : '. ; CA Number: AAAO (.A-).wnerAssigned/Controlled " CA.N,. -25 Corrective Action Type: Cause Code: Resp Org: (ES) Evaluation Support "'(NA) Not a DefiCiency
-.DBPT 0 .,. ....R
 
== Description:==
:
Alter DB-PF-03550 to require.a generatedfor DesignEngineering to review G the pump curve information gathered during this test, and to, evaluate for any: impact on Design.Flow 1 Requirements and Assumptions to maintain the System Design Basis.A This corrective action has been created from CA 02704514707 to split out the remaining work and T create.an individual corrective action for each procedure that still requires alteration,:
0 R..... ....... .7 7 Completed By: O.rganization:
Date: .Phone Attachments: "PES .8/16/2004
.7756, " Yes "[e, No If a Refueling Outage is required, .Other Tracking#
CorrectiveActibnbue.Date:
ACC- Enter~the Refueling Outage number: ./A 'B-PF-03550, .30/2004 EPT Approval: (Enter Name and Sign) section: Date:,:."":BPT O 8/-1712004 QUAL. Quality Organization Approval:
Date: ITY 1 /201/2004 4- --- -os ---.
........--.
Revised/Alternate SoIutioO.-..-..Not Performed M DB-PF-03550 has been altered to require a Condition Report tobe generated for Design P Engineering to review the pump curve information.
gathered during this test, and to evaluate for any L impact upon the Design Basis. Thistis inicorporated in Revision 01, which went effective on 9/27/04.E M E Alternate Corrective Action or justification if.Corrective Action not performed:
.. .... ................. ..... ...... .. ...... ....... ........ ..........
.----N I N Corrective Action Implementation Date: 9/2712004 G J Signature indicates Corrective Action conlplete:.
o 'Completed By: ."" Date: 9/27/2004 R U Signature indicates verification for SCAQ CRs-Verified By: Date: ' J Enter Name and Sign: Implementing Organization Approval:
.. Date: 9/27/2004 Q V Comments: U E Verified rev 1 to procedure effective 9/27/04-and added appropriate steps.AbR L I -10/20/04 T I .Y E R ......... .... ... ...............
Approval:
Date: 10/21/2004 Page 50 of.59 Site: G201 CORRECTIVE ACTION CR NUmber: NOP-LP-2001-05 0204514 .CR Category:
Action Type: Schedule Type:.. CA Number:" AA ( ) (A) Owner.Assign~ed/COntrolled.
26: Corrective Action Type: Cause Code: ...RespOrg: (ES ) Evaluation Support ( NA) Not a Deficiency
: DBPT R
 
== Description:==
 
Alter DB-PF-03551 to require a Condition Report to be generated for Design Engineering to review G the pump curve information gathered dulring this test, and to evaluate for any impact on Design: Fow Requirements and Assumptions to maintain the System Design Basis.N A This corrective action has been created from CA 02-04514-07 to split out the remaining work arid..'T create an individual corrective -action for each procedure that still requires alteration., 0 R Completed By: Oganization:
.Date: Phone: i Attachments:
PES. 8/16/12004 7756 __-Yes NO If a Refueling Outage is required, -Other Tracking # Corrective Action Due Date: A Enter the Refueling Outagd N.A A 1 DB-PF-03551 9130/2004
.EPT Approval: (Enter Name and Sign) Section I Date:* :3" *. ..' ." " " " .. '. DBPT .": "':, 81f17/2004:.',:':
QUAL, Quality Organization Approval:
Date .I Response:
.----. Compleed-as wpilten ..Re~visedf/Aternate Solutio, .-. .Not Performed:.
M DB-PF-03551 has been altered to require atConrdition Report to*tbe generated for Design :.P Engineering to review the pump curve information, gathered during.this test, and:to evaluate for any L. impact upon the Design Basis. This is incorporated in Revisioh 02, which wenteffective on 9127104: E M E Alternate CorrectiveAction or Justification if.Corrective:Action not perforrned:.:
N T N ' Corrective Action Implementation Date: 9/27/2004 G Signature indicates Corrective Action cornplete:
__Completed By: * ---Date: 9/27/2004 o1 --..... ..,.p e e y _ ......................
..... ............
-... ... .:. .........
:............: ... ... : ._ _:.:D ._ .{ .............
:.......R J Signature indicates verification for SCAQ CRs:....rified By: _. ............
.... ..._*_._._ D ate. " ..._j Enter Name an-. Sign: Implementing Organization Approval:
Il Date: 9/27/2004 Q v ! Comments: u E Verified Revision 2 to procedure effective 9/27/04 and added appropriate steps.A R LI 11-1 0/20/04 TI R .. .......... ...... ........ ..... .......Approval:
W__- _ " .,te&#xfd; 10/21/2004.'
Page 51:of 59 Site: G201 CORRECTIVE ACTION .CR Number: NOP-LP-2001-05
.02-04514 CR Category:
Action Type: I Schedule Type: CA Number: AA (() (A Owner Assigned/Controlled 27, Corrective Action Type: Cause Code: .." ". Resp Org: (ES) EvaluationSupport (NA) Not a Deficiency " .DBTS R
 
== Description:==
 
Alter DB-PF-04207 to require a Condition Report to be generated for Desigh Engineering to G 'the pump curve information gathered during this test, and to evaluate for any irhpact on Design Flow..Requirements and Assumptions to maintain the System Design Basis:*N A Note that the current plans. are to replace existing procedure DB-PF-04207.with DB-PF-03082 .which.:., T is in development.
0 R o .This corrective action has been created from..CA 02-04514-07 to split out the remaining work and R create an individual corrective,action for each procedure' that still requires alteration..
Completed By:: Organization:
6Date: Phone: Attachmrbents::
ES 8/16120M4 7756 .Yes. L.If a Refueling Outage is required, Other Tracking # Corrective Action Due Date: ACC-
* Enter the Refueling Outage number: * .A .* DB-PF-03082 1/312005.:
ET Approvat: (Enter Name and Sign) Seto Date, osr .8/7/004 QUAL" Q'uality Organization Approval:
..Date .ITY ,5/23t.2005
.'Resporse..
..Completedas wrften (. Revised/Alternate Solution ....QNot Performned M This corrective action is an.enhancement action that has been converted to SAP Activity Tracking .P Item 17186. This corrective action meetsthe requirements as.an activity tracking item per* L 'Attachment 1 of NOBP-LP-2019.
'M E Alternate Corrective Action or Justification if Corrective Action not performed:
.N T CN " corrective Action Implementation Date: .12/2/2005" ,G , Signature indicates Corrective Action complete:.-
* Completed By: Date: 12/2/2005 R ~ ~ ~~~~ ...........
ia~o ..............
O: Q *. , .........-
~i :..........
R _j Signature indicates verification, for SCAQ GRs: G Verified By: Date:.Giid y .... .......................
..............-
............
..,.......
............
................
...... .....................................
...............
................
'....... ..... .;.'. ................ ) .a~. ..... .............
..L'.............
'..... ...._j Enter Name and Sir n Implementing Organization Approval:
--* Date: 1217/2005 Q V Comments:.U E Verified notification 600266401 generated from AlF 17186 to perform this action.A R SI * -12/20/05 T. I R' ~~ ~~~~~ -- -- ..........
.............
Approval:
Date: 12/20/2005 page 52 of 59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-04514 CR Category; Action Type; Schedule Type' .CA Number: AA .4. ) (A) Owner Assigned/Controlled
.28 Corrective Action Type: Cause Code: .Resp Org: (ES ) Evaluation Support (NA) Not a Deficiency DBTS R
 
== Description:==
.
Alter DB-PF-04208 to require a Condition Report to be generated for Design Engineering to review G the pump curve information gathered duringthis test, and to evaluate for any impact on:Design Flow.I .Requirements and Assumptions to maintain the System Design Basis.N A Note that the current plans are to replace existing procedure DB-PF-04208&with DB-PF-03083, which T is in development.
0 This corrective action has been created from CA 02-04514-07 to split out the remaining work and SR create an individual corrective action for each procedure that still requires alteration., " ~~~PES 0 /1,104 ' 7756 ...', iYes: [i;No ."If a efueling Outage is required, -----Other Tracking CorrectiveAction DueDate: ACC- Enter the Refueling Outage number: N/A DB.PF-03083 12/31 /2005 EPT Approval: (Enter Name and Sign) .Seciion:-
Date:...7DBTS8/2004".
QUAL. Quality Organization Approval:
Date: ITY ' ., 5/23/2005 Response-
.. .CoMplked~as~writen
,, Revised/Alternate Solution -. ,,-Not Performed M This corrective action is an enhancement action that has been converted to SAP Activity Tracking P Item 17188. This corrective action meets the requirements as an activity tracking itemper L Attachment 1 of NOBP-LP-2019.
E Alternate Corrective Action or Justification if Corrective Action not performed:
... .... ...... ... ... .. .....- -- --N T N Corrective Action Implementation Date: 12/2/2005 G .j Signature indicates Corrective Action complete:-Completed By: ..... ... Date: 12/212005 R d Signature indicates verification for SCAQ CRs: G Verified By: Date:.......... ...... ..... ..... ...... ....................
..... ................................
...J r, t e N a m e a n d S g n :. ...........
...............
................
.....................
...... ....................................................................................................................................
JEnter Name and Sign: Implementing Organization Approval:
o Date: 12/7/2005 Q V Comments: U E Verified notification 600266402 generated from AIF 17188 to perform this action.A R LI -12/20/05 T I YE E Approval:
Date: 12/20/2005 Page 53 of,59 Site: G201 CORRECTIVE ACTION CR Number: NOP-LP-2001-05 02-04514 CR Category:
Action Type: i Schedule Type: -CA Number: AA ( ) (A) Owner Assigned/Controlled 29 Corrective Action Type: Cause Code: Resp Org: (ES) Evaluation Support (NA): Not a Deficiency.
DBTS R
 
== Description:==
 
DB-PF-03437 is a baseline testing procedure for Containment Spray Pump 1 that is currently in G development.
This procedure needs to contain a step that requires a Condition Report to be generated for Design Engineering to review the pump curve information gathered during this test,.N and to evaluate for any impact on Design Flow Requirements and Assumptions to maintain the A System Design Basis.T This corrective action has been created from CA 02-04514-07 to split out the remaining work and create an individual corrective action for each procedure that still requires alteration.
R"~~~~. ......................
' "' .Completed By: Organization:
ate: .Phone:' ..Attachments:,.
PES 8/16/2004., 7756.. " Yes I.No, If a Refueling Outage is required, Other Tracking # Corrective Action Due Date:""AcC Enter the Refueling Outage number: N/A DB-PF-(34.7 11/30/2005
" EPT Approval: (Enter.Name and Sign) Section-, Date:.DBTS .8/17/.2004 QUAL, Quality Organization Approval:
Date: ITY. ." .5/23/2005
.-. Response:-.
0 Completed as written .0 RevisedtAlternate.Solution.
-.Not Performed
.. .'M This corrective action is an enhancement action that has been converted to SAP Activity Tracking P Item 16390 in accordance with NOBP-LP-2019.
E M E Alternate Corrective Action or Justification if Corrective Action not performed:
N T Corrective Action Implementation Date: 11/28/2005 G ..........
j Signature indicates Corrective Action complete: O .Completed By: .Date: 11/28/2005 R j Signature indicates verification for SCAQ CRs: .D G Verified By: * "__-Date:_
_J Enter Name and Sign:..Implementing Organization Approval:
.-Date: 11/28/2005 Q v Comments: U E Verified notification 600264896 generated from AIF J16390 to perform this action., A R L I-~~12/20/05.
.IF.T I-Y E Approval:
Date: 12/20/2005 Page 514 of 59 Site: G201 CORRECTIVE ACTION CR Number: tNOP-LP-2001-05 02-04514 CR Category:
Action Type: Schedule Type: CA Number: AA .(A) Owner Assigned/Controlled 30 Corrective Action Type:, Cause Code: Resp Org: (ES) Evaluation Support (NA) Not a Deficiency I *DBTS R
 
== Description:==
 
DB-PF-03438 is a baseline testing procedure for Containment Spray Pump.2 thai is currently in-G 'development.
This procedure needs to contain a step that requires a Condition Report to be I generated for Design Engineering to review the pump curve information gathered during this test, N and to evaluate for any impact on Design Flow Requirements and Assumptions to maintain the A System Design Basis.T This corrective action has been created from CA 02-04514-07 to split out the remaining work and R create an individual corrective action for each procedure that still requires alteration.
EP ApCompleted By e a Sig ione: D Attachments:
PEST 8116/2004
.7756 /i. 11Yes _N'".If a"Refueling Outage is required, ':. .""Other..Track ing # .Corrective Action Due Date: AC'C "Ent~er the.Refueling Outage " : ) "I :'BP-03438
* '.110205"'"'
.... A- pprov al:(Ente r,Na me and S ig n) &sect;... ec n: -6 ! :* :' '. .... ....: .. ,.* DBTS, ' " 8/17/2004 QUAL, Quality Organization Approval:
Date: ITY.. * ", 5/23/2005-- Response:-....
Comrpetedas ,atten ) RevisediAlternate Solution .Not Performed M This corrective action is an enhancement action that has:been converted to SAP Activity Tracking P 'Item 16394 in accordance with NOBP-LP-2019.
.L E M E Alternate Corrective Action.or Justification if Corrective Action not performed:
E :.........
..........
.................
................
........................
I................
...........
.. ............
".........
... ' .....:I I'IIII ....................
.. ... .... ... ......................
... ........................
... ...............
.... .........................................
...................................'II : ........ .I ........ ..N T N N Corrective Action Implementation Date: 11/28/2005 G -_ _J Signature indicates Corrective Action complete: 'Completed*By:
..Date: 11/28/2005 R _J Signature indicates verification for SCAQ CRs: G .Verified' By: Date: j Enter Name and Sign: Implementing Organization Approval: Date: 11128/2005 Q v Comments: u E Verified notification 600264895 generated from AlF 16394 to perform this action.A R L I I F 12200 T I Y E R % ; ; ;; i ..............
...... .........
.........
... ..........................
.............
... .... ....... ......... .: ~ .............
.. ... .-.-----I : ..........
....... ................
:....... ...... .;i ; i ;' .........
.. ... ...Approval:
.Date: 12/20/2005 Page 55 of 59 Site: G201 CORRECTIVE ACTION CR Number.NOP.LP-2001-05
--02-04514, CR Category:
Action Type: ScheduleType:. , .CA.Number:
AA .. --. : (A.) Owner Assigned/Controlled
.31 Corrective Action Type:, .Cause Code: -* .. ..Resp' rg: ('ES) Evaluation Support' l (NA) Not a Deficiency
' ." :: " ". DBTS .0 .........................
.........
.... .... ...L,...R
 
== Description:==
 
6 DB-PF-03472 is a baseline testing procedure for Makeup Pump 1 that is currently in development,.
G This procedure needs to contain a step that requires a Condition Report to be generated for Design-I Engineering to review the pump curve information gathered during this test," and to evaluate for any*N. 'impact on Design 'Flow*Requirements and Assumptions to maintain the System Design Basis.'A " " .""".T This corrective action has been created from CA 02-04514-07 to.split out the remaining work and create an individual corrective action for each prodedure that still requires alteration.
R" ' "Or aniz tio -'' I ,ate: -" Completed By: Organization:
Date: -".Phone:
1 Attachments:
PES -/: 811612004.
7756 ..A'.yes, 1 No"" .. 'f a Refueling.
Outage is required, * '*'." : .- iOttler:Tracking
# ' " .Co)rrective Action:DuJeD a't~ : :.ACC- :Enter the Refueling Outage number: NiA DB-PF-03472*
6/3012005 EPT 'Approval: (Enter Name and ,Sin) .' ' Section: Daei: QUAL&#xfd; Quality Organization Approval:
ate:.Response:-.
* Completed aswriten .Revised/AlternateSolution
.Not: Performed Baseline test procedure DB-PF-03472 has.been created for Makeup Pump 1. Step..468 of.this,..procedure requires a Cohdition.Report to be generated for:Design Engineeringto review/evaluate L the mpact of motor data.and pumphydrau li.data on-the Design Bas is: This is incorporated into E Revisioni 00, which went effective on 5/10/05.,'
M, E Alternate Corrective Action or Justification if Corrective Actiontnot performed:6&
.. ................
N T N.Corrective Action'lmplementati6n Date: 5/10/2005 G J Signature indicates Corrective Action complete:.o .Completed By: ' ' Date: 5/10/2005 R J Signature indicates veification for SCAQ CRs: .*G Verified By: , .* .Date:.JEnter Name and Sign: Implementing Organization Approval:
' ' Date: 5/10/2005 Q V Comments: U E This enhancement action is to include a step in baseline test procedure DB-PF-03472 (for.Makeup.:
A R Pump #1) that requires a Condition Report to be generated for Design Engineering to review the L I., pump curve information gathered during this test, and to evaluate for any'impact on Design Flow T I Requirements and Assumptions'to maintain the System' Design Basis. TheImplementation
.-Y E Response is that the procedure has been created (effective.
5/10/05) and Step-4.68 requires'a' R Condition Report to be generated for Design. Engineeringjto review/evaluate the impact of. motor data and pump hydraulic data on the Design.Basis.
'Page 56 of 59 Site: G201 i i ~C.ORR.ECT1VE ACTION.''i:i; i'IC.ubr NO Pi-LP-2001-05
: 'i: : " ::,.I '* Al ,::...:i:.
". i :.:i:.:: :' .2 0 5,.
* 1* Reviewed DB PF 03472, Makeup Pump 1 Baseline Test.
theprocedure was effective 5/10/05..Verified Step 4.68 states: "Ensure, Engineering has performed the.following:
-Baselinetest data has been evaluated and associated documrientation is attached.,-Pump hydraulic'data have been forwvard [sic] to'Design Engineering.
-IF PdMAmotor data was.taken,THEN motor data have been forward [sic] to Design Engineering.
.-Initiated a Condition R~eport for'Design Engineerig
:to reiew/eva uate:the impact*of the following on the'Design Basis:-Motor data-Pump hydraulic data This corrective:action has been adequately implemented.
DBOV, 5/23/05...
.. ....r&deg;........rov.l:
..... ..Date. 5123/2005 Page 57 of 59
&#xa3;ifo"
* CR Number: NOP-LP-2001-0.502451 CR Category.
Action Type: Schedule Type: CA Number: AA () (A) OwnerAssigned/Controlled 32 Corrective Action Type: Cause Code: -Repg: (ES )Evaluation Support (NA).Nota Deficiency DBTS R
 
== Description:==
 
DB-PF-03477 is a baseline testing procedure for Makeup Pump 2 that is currently in development.
G This procedure needs to contain a step that requires a Condition Report to be generated for Design Engineering to.review the pump curve, information gathered during this test, and to evaluate for any.impact on Design Flow Requirements and Assumptions to maintain the System Design.Basis.
a'A.This corrective action hasbeen created from CA 02-04514-07-to.split out the remaining.work and:create an individual corrective action for each procedure that still requires alteration.
R ,~. ....'.. .... ..,'Completed By: Organization:
I Date: Pone: [.Attaclhments PES 8/16/20 .7756 Yes No" a 1 1Refueling Outage is required, .. ther Tracking # Corrective Action Due Date:,.ACC. Enterthe, Refueling outage number: N/A DB-PF-03477 6/30/2005 EPT .Approval:' (E-nter Name and Sign) Section: Date: ' DBTS .8/17/2004 QUAL ,QualityOr'ganization Approval:
Date: t 5/23/,20 05 Response:
Completed as wriften .Revised/AIternate Solution fN-) ot Pefforme :.!M Baseline test procedureDB-PF-03477 has been created f6r.Makeup Pump 2. Step 4.68,of this P procedure requires a Condition Report to be generated for. Design Engineering.
to review/evaluate L the impactof motor data and pump hydraulic data on the Design Basis. This is incorporated into" E Revisioni.00, which went effective0on 5/10/05. .%E ..Alternate Corrective Action or Justification if Corrective Action not performed:
...., ...... ... .... ..:. .. .... ..... .. ... .. ....... ............%N: N G ' .* Corrective Action Implementation Date: 5/10/2005 I Signature indicates Corrective Action con ' .te, 0. Completed By: -Date: 5/1012005 R .J' Signature indicates verification for SCAQ CRs*6 Verified By: Date: J Enter Name and Sign: Implementing Organization Approval:
b Date: 5/10/2005' Q V Comments:.u E This enhancement action is.to include a step in baseline test procedure DB-PF-03477 (for Makeup;A R Pump #2) that requires.
a Condition Report to be generated f6r Design Engineering to review the I F pump curve information gathered during this test, and to evaluate.for any impact on Design Flow T It Requirements.and Assumptions to maintain the'System Design Basis. The Implementation Y E Response is that the procedure has been created (effective 5/10/05) and Step 4.68 requires a R Conldition Report to be generated for Design Engineering to review/evaluate the impact of.motor data and pump hydraulic data on the Design Basis..Page 58 of 59 Site: G201 CORRECTIVE ACTI O U. ;: .: T': CIR Numbr NOP-LP-2001-05 " ., (: ; 2054.ReviewedDB-PF-03477, MakeupPump 2 Baseline Test. Verified the procedure was'effetive
,5/10/05.
Verified Step 4.68 states: "Ensure Engineering has performed.the following:
-Baselinetest data has been evaluated and associated documentation is attached.-
.Pump hydraulic data have been forward [sic] to Design Engineering.
:-IF PdMA motordataWas taken, THEN motoi&#xfd; data have been forward [sic] to.Design Engineering.
.- Initiated a Condition Report forDesign Enginee~ihg to review/evaluate the impact of the following on the DesignmBasis:
-Motor data-Pump hydraulic data'CR #____ _This~corrective action has been adequately implemented.
D BOV, 5/23/0 Aprv. Date:....3/2.0 0 Approval:..
.. ......: ... Date: .5/23/20.05&
Page 59 of 59 Per ISTS 3.0.1 Bases, SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits.ISTS SR 3.5.2.4 states "Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head." The frequency is stated as "In accordance with the Inservice Testing Program." CTS surveillance requirement 4.5.2.h is equivalent to ISTS SR 3.5.2.4 (shown in the ITS Conversion as SR 3.5.2.2).ISTS SR 3.5.2.4 Bases state "Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code (Ref. 6 -listed as ASME Code for Operation and Maintenance of Nuclear Power Plants.).
This type of testing may be accomplished by measuring the pump's developed head at only one point of the pump's characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant accident analysis.
SRs are specified in the Inservice Testing Program of the ASME Code. The ASME Code provides the activities and Frequencies necessary to satisfy the requirements." Verification against pump baseline is the code requirement for trending to detect degradation over time. Verification of performance at the test flow being greater than or equal to the performance assumed in the plant accident analysis ensures operation within safety limits.For Item 1.b.(i): Per CR 02-04514 corrective actions, Design Engineering calculations were performed to establish the 1ST surveillance procedure Design Basis Acceptance Criteria.
Included in the calculated acceptance criteria are adjustments for instrument inaccuracy, and for EDG frequency degradation.
Therefore, meeting the surveillance procedure acceptance criteria assures that the acceptance criteria described in the ISTS SR 3.5.2.4 Bases are met.For Item l.b.(ii):
See above discussion for ISTS SR 3.0.1 Bases and.ISTS SR 3.5.2.4 Bases.
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Bases JFD. Number: None None L.2 None Ref: Attachment 1, Volume 10, Rev. 0, Page 30 of 98 (also 43 and 55 of 98)---- REQUEST: 1.Clarify the Justifications for Deviation (JFDs) associated with the following deviations from STS SR 3.3.2.3-associated BASES: STS SR scope reduction by addition of the following BASES phrases: "...by venting the ECCS pump casings and discharge piping high points" (taken from CTS SR 4.5.2.b), and".... and the fact that some venting points are not accessible during normal operation," 2. Explain why the proposed revisions to the BASES do not alter ITS SR 3.3.2.3 and thus make the modified BASES part of the TS, contrary to &sect; 50.36 (a).Comment ----BASIS FOR REQUEST: 10 CFR 50.36 (a) states that a summary statement of the bases or reasons for (technical) specifications...
shall also be included-in the application, but shall not become part of the technical specifications.
The proposed deviation from STS SR 3.3.2.3 BASES appears to limit the SR scope. It does not simply provide a summary statement of the bases or reasons for the TS SR.BJFD 2 states: "Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description." BJFD 3 states "Changes are made to reflect those changes made to the Specification" BJFD's 2 and 3 were provided to explain why the ISTS bases were modified.---- REGULATORY REQUIREMENTS:
10 CFR 50.36 (a) A summary statement of the bases or reasons for (technical) http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 8, specifications...
shall also be included in the'application, but shall not become part of the technical specifications.
Issue ]01/31/2008 Clo~s Dae &sect;i [ 06/19/2008 Logged in User: Anonymous'Responses Licensee Response by Jerry Davis-Besse believes that the reviewer incorrectly stated STS Jones on 02/13/2008 3.3.2.3. The page numbers and the section reviewed would be associated with ITS SR 3.5.2.3. Question 1: CTS 4.5.2.b (Volume 10, Page 30) requires verifying the ECCS piping is full of water every REFUELING INTERVAL (i.e., 24 months) "by venting the ECCS pump casings and discharge piping high points." This requirement was moved to the ITS SR 3.5.2.3 Bases (Page 55) as described in Discussion of Change LA02 (Pages 34 and 35).Furthermore, the Frequency for this SR in the ISTS is 31 days.Since the Davis-Besse current licensing basis Frequency for this SR is 24 months (which is maintained in the ITS), an appropriate reason for this less frequent interval "and the fact that some venting points are not accessible during normal operation" has been included in the Bases. Both of these changes are justified by Bases Justification for Deviation (JFD) 2 (Page 59). This JFD includes the reasons of "system description" and "licensing basis description," which Davis-Besse believes covers the addition of this information to the Bases. Question 2: ISTS SR 3.5.2.3 only requires the ECCS piping to be filled with water. It does not explain how to verify the piping is filled with water. The Davis-Besse current licensing basis explanation "by venting..." provides a manner in which to perform this SR. Thus, this change cannot alter the ISTS SR requirement.
The other statement being added is simply a reason for why the 24 month Frequency is acceptable.
Thus it cannot alter the ISTS SR requirement.
Furthermore, the same SR in both the BWR/4 and BWR/6 ISTS, NUREG-1433 and NUREG-1434, includes the statement that the SR can be performed by venting at the high points (Page B 3.5.1-8 for both__NUREGs).
NRC Response by Ross Telson on 02/15/2008 Mr. Jones is correct. The reviewer incorrectly stated STS 3.3.2.3.The intended reference was STS/ITS SR 3.5.2.3. and associated BASIS. Mr. Jones also stated that the ITS basis explanation of how to satisfy the ITS SR cannot alter it. 10 CFR 50.36(a) addresses TS Bases, in part, as follows: "A'summary statement of the bases or reasons for such specifications
... shall also be included in the application, but shall not become part of the technical specifications." Where the ITS requirement is explicit, an unintended contradiction in the Basis cannot alter it. However, when the ITS requirement has had details relocated to the Bases, those details may be used by operators or staff (in a similar manner to any other current licensing basis document) to interpret the http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/lfddcea1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 3 of 8 requirement.
The ITS basis is part of the current licensing basis.Staff may deny a TS amendment if its Bases contain information that contradicts or unacceptably interprets the ITS requirement.
This, in part, is why the Bases Control Program prohibits licensees from changing the Bases in a manner that alters the TS -without NRC approval.
Thus, review of ITS Bases is an integral part of ITS conversion.
Mr. Jones also stated that the current licensing basis Frequency for this SR is 24 months (as opposed to the STS bracketed frequency of 31 days) and that this is an appropriate reason for this less frequent interval.
Based on the information obtained in the ITS conversion application, the Q&A process, and readily-available sources (e.g. applicable licensee event reports, inspection findings, other operating experience, ...) the ITSB reviewer remains unable to arrive at a determination that the proposed ITS SR 3.5.2.3. and BASES deviations from the NUREG are acceptable.
Regarding the proposed deviations, the reviewer questions whether: (a) they are in the spirit of the applicable references listed below, (b) they represent plant-specific provisions that affect ITS completeness, (b) changes are needed to make ITS SR 3.5.2.3. and BASES acceptable from the standpoint of adequate protection or compliance with NRC regulations, (c) they comply with 10 CFR 50.36 in assuring the necessary quality of systems and components will be maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met (d) they are consistent with several key elements of the current licensing basis (e.g. NRC Administrative Letter 98-10: "Dispositioning Of Technical Specifications That Are Insufficient To Assure Plant Safety" provides guidance for compliance with current licensing basis elements such as 10 CFR 50.36 and 10 CFR 50 Appendix B Criterion XVI.) If, for instance, a licensee determines that administrative controls are necessary to survey the ECCS for gas accumulation more often, in different locations, or using different methods than specified in the CTS (to assure the necessary quality of systems and components will be maintained, that facility operation will be within safety limits, or that the limiting conditions for operation will be met), that determination could signify a condition adverse to quality -one that may warrant a TS amendment in accordance with AL 98-10.Based on Mr. Jones's response and the above information, the reviewer has requested the engagement of NRR staff with the functional authority to assess and to concur with the adequacy of the proposed deviations from STS SR 3.5.2.3. and BASES.Reactor Systems Branch (SRXB) staff have knowledge and experience in dealing with ECCS gas management and should be able to assist in resolving this matter. The reviewer will also seek insights from Regional and Resident staff. ---Applicable References NEI 96-06 -Improved TS Conversion Guidance, 2.7 Deviations from the Applicable ISTS: "...a high threshold should be satisfied for deviating from the ISTS..." 58 FR 29132 (pp 39132-39139)
Final Policy on &sect; 50.36 Technical Specifications, http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 4 of 8 IV. The Commission Policy: "...it is the Commission intent that the wording and Bases of the improved STS be used in the Technical Specification related submittal to the extent practicable." Licensee Response by Jerry Jones on 03/07/2008 Davis-Besse is providing additional information to explain why our Current Technical Specification is adequate in only requiring the verification to be performed every 24 months. Davis-Besse also notes that the most recently approved ITS conversion, Beaver Valley Units 2 and 3 (a plant that is part of the FENOC organization), did not require this SR to even be included in the ITS. USAR Section 6.3.3.2.5 states: "During normal Station operation, the ECCS lines will be maintained full by the static head created by the relative elevations of the BWST (bottom at elevation 585') and the emergency sump valves (elevation 506'-8".) During the postulatedaccident, the minimum water level in the BWST before transfer to the emergency sump is above the tank discharge line. Since the tank discharge line is never drained during the postulated accidents, water hammer due to line filling will not occur. The highest point in any discharge piping (593"-03/4" for the Auxiliary Spray Line) is well below the operating level in the BWST, thus providing a significant positive head on the system. A small amount of gas or vapor could be trapped at the closed HPI discharge valves (at the containment vessel) or closed LPI check valves (at the reactor coolant piping). This small volume of voiding could not cause a water hammer. Manual venting capability is provided at the ECCS pump casings and discharge high points. Even in the event that the piping downstream of the motor-operated HPI discharge throttle valves is completely void of liquid water, an analysis has been performed to verify that no unacceptable forces on the lines will occur in the event of an HPI system actuation." License Amendment 25 (ADAMS Accession Number ML021160263) revised CTS 4.5.2.b.The previous Surveillance (prior to. Amendment
: 25) required verification that piping is full of water by venting at the high points once every 31 days. License Amendment 25 revised the SR to state: "At least once per 18 months, or prior to operation after ECCS piping has been drained by verifying that the ECCS piping is full of water byventing the ECCS pump casings and discharge piping high points." The NRC Safety Evaluation provided insight into why it was determined that the change was acceptable, and stated: "To vent the piping at the high points requires entry into the containment and results in a certain amount of radiation exposure to personnel involved.
For the low pressure safety injection (LPSI)lines, if a leak should develop on the discharge side of the pumps, the pathway from the Borated Water Storage Tank (BWST) is open (through check valves and locked open valves) and available to replace any leakage out of the system. As a result, the LPSI lines will always be full." The Safety Evaluation also discussed the High Pressure Injection System and that the likelihood of air pockets forming in the piping is remote. It further stated that: "On http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e....
7/17/2008 NRC ITS Tracking Page 5 of 8 the basis of the above, we find the proposed change would not decrease the margin of safety and therefore is acceptable." Additional insight can be found in License Amendment 208 (ADAMS Accession Number ML021210110).
The NRC Safety Evaluation states: "As stated in the Safety Evaluation Report (SER) related to operation of the Davis-Besse Nuclear Power Station, Unit 1, NUREG-0136, issued in December 1976, manual vents are provided at the ECCS pump casings and discharge piping high points. Additionally, the SER states the plant TS will require the ECCS system piping to be verified as full by observation prior to startup and venting be a periodic SR. Supplement No. 1 to the SER issued April 1977, specifically states that an SR in the TS verify that the ECCS is water solid to minimize the potential for water hammer." License Amendment 214 (ADAMS Accession Number ML021220337) further revised the 18 month frequency of SR 4.5.2.b to each Refueling Interval (i.e., 24 months). In the NRC Safety Evaluation for this frequency change, the NRC-again reiterated that the proposed frequency for the verification of piping filled surveillance was acceptable.
Therefore, based on the above NRC Safety Evaluations, it is clear to Davis-Besse that the NRC has previously reviewed and approved the Surveillance Requirement to verify the ECCS piping is filled with water, and that it is acceptable, to maintain the Surveillance at the current 24 month Frequency.
Furthermore, both License Amendments 208 and 214 were approved after the issuance of NUREG-1430 (which was originally issued in 1992). Thus, the information that NUREG-1430 required this specific SR at a 31 day Frequency was readily available to the NRC staff.Licensee Response by Bryan Kays on 03/13/2008 As a followup to question 2 in the NRC reviewer's original question, the Davis-Besse response on 2/13/2008 stated that the same SR in both the.BWR/4 and BWR/6 ISTS, NUREG-1433 and NUREG-1434, includes the statement that the SR can be performed by venting at the high points (Page B 3.5.1-8 for both NUREGs). Davis-Besse has noted that the words in these NUREGs state that "one acceptable method of ensuring the lines are filled..." in lieu of the manner in which Davis-Besse included this information.
For consistency with the other NUREGs, Davis-Besse will modify the Bases of SR 3.5.2.3 to state the Davis-Besse acceptable method of performing this SR in a manner similar to that found in the other NUREGs. A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Ross Telson on 03/21/2008 A conference call was held at 3:15 PM on Thursday, March 20, 2008 between Alan McAllister, Davis-Besse staff reviewer (standing in for Bill Bentley, FENOC Davis-Besse ITS Conversion Project Manager) and the following NRR' staff: Russell Gibbs -Chief, Plant Licensing Branch 111-2 Thomas Wengert -Project Manager, Davis Besse (D-B) Robert Elliott -Chief, Technical Specifications Branch (ITSB) Ross Telson -ITSB Lead Reviewer for D-B ITS 3.5, Emergency Core Cooling Warren Lyon -Generic http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 6 of 8 Letter 2008-01 Lead-& Senior Reactor Systems Engineer, Reactor Systems Branch (SRXB) The purpose ofthe call and the thrust of the message was that, following discussions between NRR staff of the Reactor Systems Branch, the Technical Specifications Branch, and PlantLicensing Branch 111-2, that Davis-Besse should be advised of the following:
: 1. The licensee-proposed Improved Technical Specification Surveillance Requirement (ITS SR)3.5.2.3, (which demonstrates that the emergency core cooling system (ECCS) is full of water), and Bases, while largely (but not entirely) consistent with the Current Technical Specification (CTS), deviates from the Standard Technical Specification (STS)SR in (1) frequency, (2) scope, and (3) prescriptiveness, in a direction that is contrary to the direction staff anticipate that the NRC and industry will move in connection with actions taken to address Generic Letter (GL) 2008-01, "Managing Gas Accumulation In EmergencyCore Cooling, Decay Heat Removal, And Containment Spray Systems." 2. The proposed deviation from the STS SR 3.5.2.3 and Bases would warrant detailed SRXB review. Subjecting the licensee to such a review at this time would be premature, given the pending nature of GL-2008-01.
STS &Bases adoption, without deviation, would not warrant a detailed SRXB review, a commitment of staff resources that were not budgeted for the ITS conversion effort. Pursuing the STS deviation at this time would also unnecessarily burden the licensee and would likely delay- the ITS conversion.
: 3. The STS SR and Bases are more performance-based and less prescriptive than the CTS or the licensee-proposed ITS and Bases and, although more frequent performance (every 31 days vs. every 24 months) of the SR would be required, the STS model provides increased latitude regarding the licensee's implementation methodology.
This should permit the licensee to interpret and implement the STS in a manner that is both (1) consistent With the STS SR and Bases and (2) with the CLB, without (3) unnecessary burden. In addition to the above, staff suggested that the licensee reconsider adoption of the STS SR and Bases because: 4. The major objective of converting from plant-specific CTS to the ITS is to achieve as much consistency in the license requirements as possible, to the extent that the plant-specific design basis can conform with the related typical plant design reflected in the improved STS, 5. Significant discussion and disposition of philosophical and technical issues between the industry and the NRC occurred during development of the STS. A high threshold should be satisfied for deviating from the STS in the ITS. This high threshold is used to preserve the standardization of the use and meaning of the requirements for the industry and the NRC. Language and format preferences, unless justified on a plant-specific basis, should be avoided, and 6. STS SR 3.5.2.3 and Bases represent the current. staff model for meeting CLB requirements to maintain ECCS full of water and has been broadly adopted by the industry.Licensee Response by Bryan As part of the NRC/Industry ISTS development, each owners http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 7 of 8 Kays on 04/02/2008 group submitted proposed ISTS to the NRC. The currently issued NUREGs (NUREGs-1430 through -1434) are based on these original owners group submittals.
The B&W Owners Group submitted BAE-2076, "B&W Owners Group Revised Standard Technical Specification," on May 7, 1989. This draft did not include a Surveillance requiring verification that the ECCS piping was filled with water. The specific Surveillance in NUREG-1430, ISTS SR 3.5.2.3 (Volume 10, Page 43) was added during the negotiation phase of the ISTS development.
It was added as a bracketed Surveillance, though, which means that the requirement is plant-specific.
Brackets were added to various items in the NUREGs when the requirement was not common to all plants. A review of the three B&W plants that have adopted the ISTS shows that two of the plants, Crystal River 3 and ANO-1, do not have this Surveillance in the NRC-approved ITS. The third plant, Oconee Units 1, 2, and 3, includes only a requirement to vent the HPI and LPI pump casings every 31 days. Both ANO and Oconee explained in their ITS submittal that they were not adopting the ISTS SR as written for the same reason that Davis-Besse is not adopting the ISTS SR -it is not possible to vent all points in the piping without requiring entry into the containment, which results in a certain amount of radiation exposure to personnel involved.The applicable portions of the three B&W plants ITS submittals, as well as the original B&W Owners Group submittal is provided as an attachment to this response.
Furthermore, the last two NRC-approved Westinghouse ITS submittals, DC Cook Units 1 and 2 (SER dated 6/1/05, ML050620034) and Beaver Valley Units 1 and 2 (SER dated 7/10/06;, ML061940177), and the last NRC-approved CE ITS submittal, Calvert Cliffs (SER dated 5/4/98, MLO 10520026), do not include the ISTS SR. Therefore, Davis-Besse does not believe that the NRC reviewer's comment that the ISTS SR has been broadly adopted by the industry is correct. As shown above, all three B&W plants that have adopted the ISTS did not adopt the SR and the last three PWRs that have adopted the ISTS did not adopt the SR. In addition, Davis-Besse has decided to change the SR to be consistent with the current wording in the CTS. The SR will now read "Verify ECCS piping is full of water by venting the ECCS pump casings and discharge piping high points" and the Frequency will be maintained at 24 months (i.e., the REFUELING INTERVAL).
Thus, the details on how to ensure the piping is filled with water, which was moved to the Bases as described in Discussion of Change LA02, will be maintained in the SR, consistent with CTS current licensing basis. A draft markup regarding this change is also attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
This change also supersedes the third response of Davis-Besse (dated 3/13/2008) and the draft markup attached to the response.Licensee Response by Jerry Jones on 04/23/2008 The following is in addition to the 4/2/2008 response.
Davis-Besse is aware of the issues related to Generic Letter 2008-01 and is http:l//www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1fddcealOd3bdbb585256e...
-7/17/2008 NRC ITS Tracking Page 8 of 8 sensitive to the NRC's concerns.
Davis-Besse, in conjunction with.the other plants in the FENOC organization, is currently gathering the data requested by the Generic Letter. However, the Generic Letter does not specifically require any changes to the current Technical Specifications of any plant. The Generic Letter requests that each plant evaluate its ECCS, DHR, and Containment Spray System licensing basis, design, testing, and corrective actions to ensure that gas accumulation is maintained less than the amount that challenges OPERABILITY of these systems, and that appropriate action is taken when conditions adverse to quality are identified.
Furthermore, it requests that the plants notify the NRC of the results of the above required actions, and if any modifications to the plant, programmatic, procedure, or licensing basis is required, then the schedule for the corrective actions that have yet to be completed within 9 months of the date of the Generic Letter (approximately mid-September).
Thus, Davis-Besse will be evaluating the current requirements related to maintaining the piping filled with water, including the current Frequency and the location for the venting. In addition, the Generic Letter states that the NRC has yet to determine the proper actions to take concerning the issue. Therefore, Davis-Besse believes that the proper approach to take regarding this issue is to allow the Generic Letter requirements to control any potential Technical Specification changes. When the Davis-Besse evaluation is complete, Davis-Besse would take the appropriate actions required by the Generic Letter, which could include a Technical Specification change. Furthermore, when the NRC receives input from the utilities and makes a decision on any further necessary changes, it would probably be promulgated to the utilities via generic correspondence.
Thus, at that time Davis-Bessewould make any required changes to the Technical Specifications.
Since the proposed ITS SR is consistent with the current Technical Specifications, as described in the previous response, Davis-Besse believes that the ITS submittal, with respect to this SR, can be approved as is and any changes will be made as dictated by the Generic Letter 2008-01 outcome.NRC Response by Ross Telson INo further questions at this time. (Carl Schulten)on 06/19/2008 1 Date Created: 01/31/2008 12:41 PM by Ross Telson Last Modified:
06/19/2008 09:15 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
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ITS 3.5 Ross Telson None 30 Information ITS.Number:
OSI.; DOC.Number:
Bases.. JFD .Nu.m ber;None None L.2 None Comment /This comment reopens thread 200801311241 which was mistakenly closed 6/19/2008 (Carl Schulten).
The last licensee response was 04/23/2008.
IssueD!ate J 06/23/2008 Close Date [07/02/2008 Logged in User: Anonymous'Responses Licensee Response by Jerry During a recent phone conversation with the NRC, the NRC Jones on 07/02/2008 reviewer requested changes to the draft markup provided in the 4/2/08 Davis-Besse response to RAI 200801311241.
A draft markup regarding this change, is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
This draft markup supersedes the 4/2/08 markup for 200801311241.
Licensee Response by Jerry During a recent phone conversation with the NRC, the NRC Jones on 07/02/2008 reviewer requested changes to the draft markup provided in the previous response.
A draft markup regarding this change is attached and supersedes the previous draft markup.Licensee Response by Jerry During a recent phone conversation with the NRC, the NRC Jones on 07/02/2008 reviewer requested changes to the draft markup provided in the previous response.
A draft markup regarding this change is attached and supersedes the previous draft markup.http://www.excelservices.com/exceldbs/itstrack.
davisbesse.nsf/1 fddcea 1Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 2 Date Created: 06/23/2008 09:12 AM by Ross Telson.Last Modified:
07/02/2008 02:15 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
7/17/2008 Section 3.7 RAIs NRC ITS Tracking Page I of 2 I Return.tQo View Menu Print DocuLm!ent RlI Screening Required:
Yes This Document will be approved by: Tim Kobetz This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC ITS TRACKING NU J~V? I? wor ID1200711051331 Conference Call Requested?
No L lIn Scope ITS.-Section:
TB POC:. JD..Number.:
PageNumber(s):
ITS 3.7 Bill Cartwright Bill Cartwright None 365 Information ITS.N.umnber.:
O.S-:. DO..C Numfl.ber.:
Bases JFD Nu.mllber.:.
3.7.16 None None None Related to #200710090904 Please provide justification for the spent fuel loading patterns referenced in TS 3.7.16 to be located in the TS basis.10CFR50.36(d)(4) requires design features to be included in TS such as"geometric arrangements" that "are not included in categories described in (d)Comment (1, 2 or 3)." Thus the geometric arrangement of fuel must be included in TS.The geometric arrangements were referenced in the LCO section (c2, now d2), under TS 3.7.16.The actual loading patterns in your submittal are described in the bases rather than in the TS, or a document recognized as acceptable for relocating TS requirements (eg TRM, COLR).Issue.Date 11/05/2007 Close Date 12/27/2007 Logged in User: Anonymous-Responses Licensee Response by Jerry Jones on 11/30/2007 The allowance to include the spent fuel loading patterns referenced in ITS 3.7.16 (Volume 12, Page 369) in the Technical Specification Bases in lieu of including the patterns in the Technical Specifications was previously approved by the NRC, as documented in Amendment 247, dated October 19, 2001. The CTS Figure that is provided in the ITS submittal (Figure 3.9-1, Page 365) was included in the issued Technical Specification Amendment 247 as CTS Figure 3.9-3. The Figure stated that the http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
5/30/2008 NRC ITS Tracking Page 2 of 2 loading pattern considerations are described in the Bases.Furthermore, the Bases were also issued as part of the Amendment 247, and stated that the loading pattern restrictions are maintained in fuel handling administrative procedures.
As previously stated in the Davis-Besse response to question 200710090904, the second sentence of the Note to ITS Figure 3.7.16-1 will be modified to state "The approved loading patterns applicable to Category "A,""B," and "C" assemblies are specified in the Bases." These words provide a more positive requirement related to the loading patterns.
However, since the NRC has already approved the allowance to include the spent fuel loading patterns in the Bases, and changes to the Bases will be controlled by the Bases Control Program in ITS 5.5.13, it is the Davis-Besse position that referencing the loading patterns in the Bases is acceptable.
Therefore, no changes are required to the ITS submittal.
NRC Response by Bill Response to this question and 200710090904 are acceptable.
This Cartwright on 12/27/2007 item is closed " Date Created: 11/05/2007 01:31 PM by Bill Cartwright Last Modified:
12/27/2007 07:50 AM http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/1 fddceal Od3bdbb585256e...
5/30/2008 NRC ITS Tracking Page I 0f 2 Return to ViewMenju Print Document RAI Screening Required:
Yes This Document will be approved by: Carl Schulten; Gerald Waig This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC Reviewer I D1200804100917 Conference Call Requested?
No Caeor ln Scope ITS..S.ec.tionW:
TB POC: JFD Nunibe r:. PageNumber(s):.
ITS 3.7 Bill Cartwright 1 455 Information ITS Numnber: OSR: D OC Nunbher: Bas.esJFD Numbnh-er1:
3.7.4 None None None The Davis Besse UFSAR 15.2-34 (Loss of feedwater accident)
#7 states that "If electric power is not available or if the SFRCS signal can not be cleared, cooldown is accomplished by manually operating the atmospheric vent valves Comment with the feedwater being supplied by the auxiliary feed water system." Please reconcile the Davis Besse UFSAR with the statements made in JFD #1 proposing to not adopt ITS 3.7.4 such as "Davis Besse does not credit the AVVs in the accident analysis." ssueDate 04/10/2008 Clos~e.Date 04/28/2008 Logged in User: Anonymous-'Responses Licensee Response by Jerry Jones on 04/23/2008 ISTS 3.7.4 Justification for Deviation (JFD) 1 (Volume 12 Page 455) states that the AVVs are not credited in the accident when a loss of power has occurred.
UFSAR Section 15.2.8.2.3, Loss of Normal Feedwater, Case 2 describes the loss of feedwater event when a loss of power has occurred.
This is on UFSAR pages 15.2-37 and 38. Items 4 and 8 of Case 2 have been changed by UFSAR Change Notice (UCN) 06-594, which was approved on 1/17/07.The UCN changed the UFSAR to clarify that thermal equilibrium is established and the decay heat is removed by steam flow through the steam generator main steam safety valves, not through the atmospheric vent valves. The main steam safety valves remain in the Davis-Besse Technical Specifications (ITS 3.7.1). A copy of http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/lfddcea 1Od3bdbb585256e...
5/30/2008 NRC ITS Tracking Page 2 of 2 UCN is attached.response is acceptable.
This item is closed Date Created: 04/10/2008 09:17 AM by Bill Cartwright Last Modified:
04/28/2008 05:33 PM http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/lfddcealOd3bdbb585256e...
5/30/2008 CHANGE NOTICE FORM e q Page 01 NO- 06-594 NOP-LP-4008-01 Rev. 00 SECTIONAA1ITATION&#xfd; Y.r- BV1 Q- BV2 0 DB n PY ED UFSAR 0 Technical Specification Bases El TRM-ist Affected Sections, Tables, Figures and Applicable Pages Section Pages Section Pages 15.2.8.2.3 15.2-36, 15.2-37, 15.2-38 Initiating Document(s):
Corrective Action 05-05216-01 Brief Change
 
== Description:==
 
To enhance clarity, Section 15.2.8.2.3, Loss of Normal Feedwater, is revised to clarify that thermal equilibrium is established by steam flow through the Steam Generator main steam safety valves, not through the atmospheric vent valves. This correction is done in Cases I and 2 of this section.Indicate documents included in package: Included Marked-up Pages 0 10CFR50.59 Regulatory Applicability Determination 0 Change Notice Administrative Change Evaluation Form El Change Notice Review Form(s) 0 Not required Included Not required El 0 El 10CFR50.59 Screen 0 10CFR50.59 Evaluation El Other: __El 0 0 (El Preparer: Dennis BI Seviewer: Manager: Date: NDJ -Date Date ewer: Section Owner Reviewer: Section Owner Reviewer: Change Notice Approval (Fie rticensing):
.Date: Effective Date: 0l L26JO7 1/17/07" ...... ' SECTION3 ORPORATIO,'' , ae Document Revision 'Verified By Date: Training Complete El YES 0l N/A D-B ;Case 1 Based :- these assumptions the rupture of a feedwater line between the first feedwi. r line upstream check valve and the steam generator, with offsite power &#xfd;jailable, results in the following sequence of events: I. The rupture of a feedwater line causes an immediate two phase blowdown of the affected generator.
Low pressure in the affected feedwater line (differential pressure sensor) at I second after 114 rupture initiates closure of the main steam line isolation valves (5 seconds closing time assumed) thus isolating the unaffected steam generator on the steam side and initiating closure of the feedwater valves on the feedwater side.2. The termination of all feedwater to the affected steam generator will result in a reduction in its heat removal capability.
The heat removal capability of the affected steam generator rapidly diminishes as the steam generator blows dry as evidenced by the rapid loss of inventory, mixture height and steam generator pressure (Figure 15.2.8-2).
The unaffected steam generator maintains steam pressure which increases following isolation resulting in steam relief through the main steam line safety valves (Figure 15.2.8-3).
15 3. The eventual result is an increase in the Reactor Coolant System temperature and pressure which continues until the reactor trips on high reactor system pressure at 5.2 seconds (including
.6 sec.delay) in turn, initiating trip of the turbine generator.
: 4. Low steam line pressure (600 psia at 6.9 sec) in the affected steam generator initiates realignment of the auxiliary feedwater piping and pumps servicing the affected generator to take steam from the unaffected steam generator and start feeding the unaffected steam generator.
A conservative delay of 40 seconds was assumed.5. The feedwater line rupture control logic assures isolation of the unaffected steam generator to assure an adequate inventory and steam pressure to run the remaining available auxiliary feedwater pump turbine. Auxiliary feedwater is thus available to the unaffected steam generator for long-term core decay heat removal.6. At 35 seconds thermal equilibrium is re-established.
i.e. the heat removal rate steam flow through the t mos 9her c vent valve , is equal to the heat input (core decay heat), as evidenced y te average moderator temperature.(Figure 15.2.8-4).
EVISF 15.2-36 3REV 14 7191 D-B 7. Decay heat removal and cooldown of the Reactor Coolant System is then provided by #team relief to the atmosphere through the atmosphere vent valves with auxiliary feedwater being supplied to the unaffected steam generator.
A complete loss of feedwater reduces the heat removal capability of both steam generators.
This results in an increase in the reactor system pressure which continues until the reactor trips. As the above sequence of events states, the accident under consideration also results in a reduction in the secondary system heat removal capability.
The heat removal capability of the affected generator is completely lost as the generator blows dry.The unaffected steam generator always maintains some heat removal capability.
The result is a more rapid increase in reactor system temperature and pressure (Figures 15.2.8-4 and 15.2.8-5) than shown in Figure 15.2.8-1, thus the reactor trips sooner. Curves of reactor power and thermal power are presented In Figure 15.2.8-5 for the feedvater line break with offsite power available.
A curve of hot channel minimum DNBR versus time is presented in Figure 15.2.8-4.A plot of steam generator mass and energy release rate out the break as a function of time is presented in Figure 15.2.8-6.
Since the thermal power is less than 1121 throughout the transient, the reactor system pressure is less than core design limits and minimum hot channel DNBR is well above 1.3: there is no danger of core damage.Case 2 The case of s feedwater line rupture concurrent with a loss of offeite power at the time of rupture has been examined and was found to result in the following sequence of events.1. Loss of offsite power causes an immediate loss of all 4 reactor coolant pumps.2. The reactor trips on loss of power to the control rod drives.3. The rupture of the feedwater line causes an immediate two phase blowdown of the affected generator.
Low pressure in the affected feedwater line (differential pressure sensor) at 1 second after '114 rupture initiates closure of the main steam line isolation valves (5 second closing time assumed) thus isolating the unaffected steam generator on the steam side and initiates closure of the feedwater valves on the feedwater side.4. The affected steam generator will continuously depressurize while the unaf ected steam generator pressure increases to the tm spheric u setpoint to remove decay heat. Turbine bypass va ve re Veicpability is lost due to the loss of power to the condenser circulating pumps.15.2-37 REV 14 7/91 D-B 5. The blowdovn of the affected steam generator viii cause a reduction in the Reactor Coolant System temperature and pressure until most of its inventory is depleted.
Vith the affected steam generator and no feedvater flow assumed to the unaffected steam generator, the Reactor Coolant System temperature will increase as shown in Figure 15.2.8-7.6. Heatup of the Reactor Coolant System will continue until auxiliary feedwater is initiated.
Lov steam line pressure (600 psia at 6.9 sec) in the affected steam generator initiates realignment of the auxiliary feedvater piping and pumps servicing the affected generator to take steam from the unaffected steam generator and start feeding auxiliary feedaater to the unaffected steam generator.
A conservative delay of 40 seconds was assumed.7. The feedvater line rupture control logic assures isolation pof the unaffected steam generator to assure an adequate inventory nd steam pressure to run the remaining available auxiliary feedvater pump turbine. Auxiliary feedvater is thus available to the unaffected steam generator for long-term core decay heat removal.B. After initiation of auxiliary feedvater, thermal equilibrium is re-PEVISE established, i.e. the heat removal rate steam flov through the 9 r c ump is equal to the heat Input core ecay t). Dey eat r va and cooldown of the vy," %l espkm reactor system is provided by steam relief through the safety valves S-Rety with auxiliary feedvater being supplied to the unaffected steam generator.
The results during the critical period for this event are less severe than, but similar to the results for the station blackout, Subsection 15.2.9. The blovdown of the affected steam generator provides more cooling than the station blackout prior to reactor coolant heatup and initiation of auxiliary feedrater.
Plots of total reactor power, thermal power and RC System pressure are given in Figure 15.2.8-8.
Plots of Reactor Coolant System flow, minimum hot channel DWNt, and average 'moderator temperature are presented in Figure 15.2.8-7.
Steam pressure in the affected and unaffected generators are presented as a function of time in Figure 15.2.8-9.
Steam generator blovdown mass and energy release rates are given in Figure 15.2.8-10.
Thus the concurrent loss of offsite power with the feedvater line rupture does not result in any core damage or otherwise adversely affect the Reactor Coolant System.Case 3 A feadvater line break with a loss of offsite pover .at reactor trip was analyzed and found to be identical to the analysis vith offsite pover available (Case 1) until reactor trip. After trip, the results are similar to the loss of offxite power at the time of rupture. In the analysis for loss of offsite pomer at trip, there is less inventory in the affected steam generator at trip which results in less reactor system cooling. Auxiliary feedvater is initiated in the same manner as Cases 1 and 2 to provide for long-term decay 15.2-38 REV 1 7/83 NRC ITS Tracking Page I of 3 Reunto View Men~u Print Documen~t RAI Screening Required:
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NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes N~Dd 004 "Wn~n ID 1200804031123 Conference Call Requested?
No I CategoiyJ In Scope ITS .S.ection:
TB.P.OC,:'
JFD Number: PageN umber(s): ITS 3.7 Bill Cartwright Bill Cartwright 9 118 Information ITS Number: OR" DOC Number: BasesJFD Nlunmber.:
3.7.5 None None None Please provide justification why the design of the'Motor Driven Feed Pump (MDFP) train meets the design requirements assumed by the TSTF-412 model safety analysis.
TSTF-412 was used to justify some of the changes to ITS 3.7.5.The CLIIP TSTF-412 (Federal Register Vol 72, No. 136, page 39089) was referenced in the amendment as justification to modify the ITS.Per the Basis, the MLDFP is a non-safety related pump that is manually actuated if necessary (Basis insert #7, Attachment 1, Volume 12, p 137 of 461 in Comment the amendment package, and JFD #9). Basis insert #5, p 131 of 461 states that"The MDFP train is not credited in any accident analysis;
..." The model Safety Evaluation for TSTF-412 credits the automatic operation of AFW/EFW system for the technical specification changes proposed under that TSTF. See section 3.0 of the model safety evaluation, first sentenance, and other statements in Section 3.0 requiring AFW/EFW pumps to meet accident analysis assumptions (inferring automatic system operation).
issueDatei 04/03/2008 Close Date 05/28/2008 Logged in User: Anonymous 71Responses Licensee Response by Jerry Jones on 04/04/2008 The basic purpose of TSTF-412 was to add a new Condition to allow a motor driven AFW/EFW pump to be inoperable concurrent with the turbine driven pump being inoperable due to the reason in Condition A (i.e., one of the two steam supplies being inoperable).
Prior to TSTF-412, Condition A already http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
5/30/2008 NRC ITS Tracking , Page 2 of 3 allowed a turbine driven AFW/EFW pump to be inoperable due to an inoperable steam supply. All of the TSTF-412 changes, with the exception of new Condition C, are made either for clarification or are made to be consistent with the new Condition C. Furthermore, TSTF-412 was originally started as a Westinghouse Owners Group (WOG) change, and in the normal WOG design, there is one turbine driven AFW/EFW train and two motor driven AFW/EFW trains. The B&W design includes two turbine driven EFW trains and one non-safety related motor driven EFW train. This is clearly stated in the Background section of the ISTS Bases for ISTS 3.7.3 (Volume 12, Page 128). Davis-Besse adopted ISTS 3.7.5 ACTION A (Page 118), as modified by TSTF-412.
Since the allowance was previously in the ISTS prior to TSTF-412, and TSTF-412 only clarified that the turbine driven EFW train is inoperable when one of the two steam supplies is inoperable, Davis-Besse believes that this part of TSTF-412 is acceptable.
Davis-Besse also adopted the new Condition C (Page 120) allowed by TSTF-412.
For the normal WOG design, when in Condition C, the plant still has a fully OPERABLE automatic motor driven AFW/EFW train and an additional automatic turbine driven AFW/EFW train that still has one of the two steam supplies OPERABLE and is not inoperable for any other reason (i.e., it is inoperable only because of the inoperable steam supply). For Davis-Besse, when in Condition C, the plant still has a fully OPERABLE automatic turbine driven AFW train and an additional automatic turbine driven AFW train that still has one of the two steam supplies OPERABLE and is not inoperable for any other reason. In other words, the remaining AFW components for Davis-Besse are equivalent to that assumed in the TSTF. Furthermore, the Davis-Besse Current Technical Specifications (CTS) already allows this condition (an inoperable turbine driven AFW train concurrent with an inoperable motor driven feedwater pump) to exist for 72 hours. CTS 3.7.1.2 Action a (Page 97) allows a turbine driven AFW train to be inoperable for 72 hours and CTS 3.7.1.7 Action (Page 100) allows the motor driven feedwater pump to be inoperable for 72 hours. Nothing in the two CTS Actions precludes these two Actions from occurring simulataneously.
Thus, the proposed 48 hour time in ACTION C is more restrictive than the CTS, as described in Discussion of Change M02. Therefore, Davis-Besse believes that the addition of proposed ITS 3.7.5 ACTION C is acceptable and consistentwith the allowances in TSTF-412.NRC Response by Bill A review of the designs associated with 5 simular B&W units Cartwright on 04/10/2008 indicates that their EFW system designs include a combination of both motor and turbine driven Emergency Feedwater Pumps, all capable of meeting accident requirements.
This is not the stated design of Davis Besse. Thus the justification (model safety evaluation) does not appear to be applicable to Davis Besse. This is consistent with the wording of TSTF-412 that was written for 1OCFR50.36 criteria 3 considerations.
If the licensee wishes to pursue the TS changes in TSTF-412 then they will require http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
5/30/2008 NRC ITS Tracking Page 3 of 3 technical staff review as the proposed changes in the amendment deviate from the model application stated in the CLIIP.Licensee Response by Bill This item was discussed on the 4/9/07 phone call. At that time, Bentley on 04/14/2008 Davis-Besse stated that we wanted the change and understood that Tech Branch assistance would be required.
Since all of the changes are significantly more restrictive when compared to our CTS, we do not believe the additional assistance will be extensive, nor do we believe that it will likely cause any delay of the overall schedule.NRC Response by Bill Review by the Technical Branch indicates that no additional Cartwright on 05/28/2008 iinformation is needed from the licensee.
This item is closed.Date Created: 04/03/2008 11:23 AM by Bill Cartwright Last Modified:
05/28/2008 07:49 AM http://Www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea1Od3bdbb585256e...
5/30/2008 NRC ITS Tracking Page I of 2 Return to View Menu Print Document RAI Screening Required:
Yes This Document will be approved by: Carl Schulten; Gerald Waig This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC ITS TRACKING NUllCP piia I-A*A ID200804101104 Conference Call Requested?
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P.age..Number(s):
ITS 3.7 Bill .Cartwright None Inforiation ITS Number: OS: DOC Number: Bases JFD Number:-3.7.5 None None None Please justify that design of Davis Besse supports incorporating the Motor Driven Feed Pump (MDFP)in ITS 3.7.5, as one of three EFW pumps.A review of the designs associated
'with 5 simular B&W units indicates that their EFW system designs include a combination of both motor and turbine driven Emergency Feedwater Pumps, all capable of meeting accident Comment requirements.
This is consistent with the ITS 3.7.5 that was written for 10CFR50.36 criteria 3 considerations.
The surveilance activities and the basis were both written to support criteria 3.In the Davis Besse submittal, the MDFP is described as a manually operated pump that is "not credited in any accident analysis." Thus it does not appear that Davis Besse has the diverse design assumed by the ITS.IssueDate][
04/10/2008 Close Date 104/30/2008 Logged in User: Anonymous'Responses NRC Response by Bill Cartwright on 04/17/2008 A clarification to the initial question:
During the initial development of the ITS the NRC staff reviewed the industry reports recommending which Limiting Conditions for Operation (LCOs) should be retained in the ITS (Thomas Murley letter to Walter Wilgus, 05/09/1988).
the staffs' review concluded that one (Remote Shutdown Instrumentation) of the 75 proposed LCOs in the B&W ITS was required solely on risk (criterion 4). The http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/1 fddcealOd3bdbb585256e...
5/30/2008 NRC ITS Tracking Page 2 of 2 licensee proposes to incorporate the MDFP into TS 3.7.5 based solely on criterion 4 (due to the specific design.of Davis Besse).Please provide a summary level discussion of the criterion 4 justification used to incorpate the MDFP into TS 3.7.5 (deviating from the ITS). A satisfactory response to this item will resolve the initial question.Licensee Response by Jerry Jones on 04/25/2008 In the staff s review of the B&W report, documented in the NRC letter from Thomas Murley to Walter Wilgus, dated 5/9/88, the entire Auxiliary Feedwater System, which includes both motor driven and steam driven pumps, Was evaluated as one system and determined to meet Criterion
: 3. However, at davis-Besse, the Motor Driven Feedwater Pump (MDFP) is not an automatic system and is not assumed to function in any accident analysis.Therefore, since Davis-Besse believed that the MDFP needed to remain in the ITS, it was classified as meeting Criterion
: 4. Based on the NRC reviewer's question, Davis-Besse personnel have reviewed this issue (including a review of the Davis-Besse PRA)and confirmed that the MDFP is a system which operating experience or probabilistic safety assessment has shown to be significant to public health and safety. Thus, it meets Criterion 4 of 10 CFR 50.36 (d)(2)(ii).
However, Davis-Besse notes that, consistent with many other ITS conversions, one Davis-Besse Current Technical Specification (CTS) that meets a Criterion has been allowed to be removed from the CTS (e.g., CTS 3/4.9.3, Decay Time -Volume 14, Pages 128 through 131) using an LA-type Discussion of Change (DOC). Thus, if the NRC desires, Davis-Besse can relocate the MDFP requirements to the Technical Requirements Manual using an LA-type DOC.NRC Response by Bill Cartwright on 04/30/2008 As the licensee believes that the motor driven pump should remain in Technical Specifications based upon criterion 4, this item is closed with no further action required.Date Created: 04/10/2008 11:04 AM by Bill Cartwright Last Modified:
04/30/2008 02:31 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e...
5/30/2008 NRC ITS Tracking Page I of I Return to View Menu]J Print Dcmn RAI Screening Required:
Yes This Document will be approved by: Carl Schulten; Gerald Waig This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC Reviewer ID 1200804091719 Conference Call Requested?
No Categoty 1 In Scope ITS Section:, TB.POC: JFD Number.:.
Page.Number(s);.
ITS 3.7 Bill Cartwright None 150 Information ITS-Numuber:
OSI: DOC..Number.:
BasesJFD Number: 3.7.6 None A.4 None A04 references License Amendment Request No 05-0007 dated April 12, 2007 as the justification for this change as administrative.
Comrmen.t In light of subsequent submittals regarding the measurement uncertainty recapture (MUR) request, please validate that this reference is still valid for this DOC and the related change.IIssue Da 04/09/2008 Close D~ate[ 04/14/2008 Logged in User: Anonymous'Responses Licensee Response by Bill DOC A04 is still valid. See the response to question Bentley on 04/11/2008 200803061234.
The only section that requires change as a result of changes to the MUR License Amendment is Section 3.3.1.NRC Response by Bill Thank you. This item is closed Cartwright on 04/14/2008 1 Date Created: 04/09/2008 05:19 PM by Bill Cartwright Last Modified:
04/14/2008 07:24 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea1Od3bdbb585256e...
5/30/2008 NRC ITS Tracking Page I of 2[ R..eturn to View Men.u Document RAI Screening Required:
Yes This Document will be approved by: Carl Schulten; Gerald Waig This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC ITS TRACKING NRC Reviewer HD] 200804091725 Conference Call Requested?
No Category In Scope ITS Section: TB *POC: JFD Number: Page Number(s):
ITS '3.7 Bill Cartwright None .425 Information ITS Number: OS: D.OC N.unmber:.
Bas.es.JFD..Nunber:.
None None R.1 None The relocation justification for CTS 3/4.7.2 references B&W Technical Report 47-1170689 in bullet #4 as part of the justification for this change.Comment Has this report been reviewed and approved by the NRC?If so, when was it approved?Alternately, please provide the referenced document used to justify this change.Issue Date I04/09/2008 Close Date [04/28/2008 Logged in User: Anonymous'Responses Licensee Response by Jerry The NRC has previously reviewed and approved this B&W Jones on 04/23/2008 document.
The review and approval is documented in an NRC letter dated May 9, 1988. This is also described in Volume 1, Page 3, the Application of Selection Criteria to the Davis-Besse Nuclear Power Station Technical Specifications.
A copy of the relevant parts of the NRC letter (those applicable to a B&W plant) is attached.NRC Response by Bill ][Thank you for the clarification, this item is closed Cartwright on 04/28/2008
____________________________
Date Created: 04/09/2008 05:25 PM by Bill Cartwriglht http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal0d3bdbb585256e...
5/30/2008 NRC ITS Tracking P Page 2 of 2 Last Modified:
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5/30/2008 "IsEXT- 88-04897 UNITED STATES NUCLEAR REGULATORY COMMISSION WASNINGoTON.
D.C. 2oSSS IL IL bORsUM" NPD UCENSiNa 9 MAY 9 1gg1 Hr. Walter S. Wilgus, Chairman The B&W Owners Group Suite 525 1700 Rockvllle Pike Rockville, Maryland 20852 MAY 10 19R8 3oM,2a3021=o
 
==Dear Mr. Wilgus:==
This letter is in response to your report identifying which Standard Technical Specification (STS) requirements you believe should be retained In the new STS and which can be relocated to other licensee-controlled documents.
The enclosure to this letter documents the NRC staff's conclusions as to which current STS requirements must be retained in the new STS. These conclusions are based on the Commission'.sInterim Policy Statement on Technical Specifica-tion Improvements and on severel interpretations of how to apply the screening criteria contained In that PolicyStatement.
The NRC staff considered comments-made by industry at a March 29, 19U meeting between HRC, HUHARC, and each Owners Group in making these Interpretations.
Based on our review, we have concluded that a significant reduction can be made in the number of Limiting Conditions for Operation (and associated Surveillance Requirements) that must be included in the STS. Our goal is to assure that the new STS contain only requirements that are consistent with 10 CFR 50.36 and have a sound safety basis.The development of the new STS based on the staff's conclusions will result in more efficient use of NRC and industry resources.
Safety improvements are expected through more operator-oriented Technical Specifications,'improved Technical Specification Bases, a reduction in action statement-induced plant transients, and a reduction in testing at power.As you are aware, the NRC staff and industry also have underway a parallel program of specific line item improvements to both the scope and substance of the existing Technical Specifications.
The need for many of these types of improvements was identified In the report (NUREG-1024) of a major staff task group established in 1983 to study surveillance requirements in Technical Specifications and develop alternative approaches to provide better assurance that surveillance testing does not adversely impact safety. The NRC will continue to actively Identify and pursue the development of specific line item Improvements to Technical Specifications and will make these Improvements Immediately available to licensees without waiting for the new STS. We encour-age each of the Owners Groups to continue to work with the NRC staff on these types of parallel improverants to existing Technical Specifications.
V ivZr V
, Y 3 -i&#xfd;.3 Mr. W. S. Wilgus I-2-We are confident that the enclosed staff report provides an adequate basis for the Owners Groups to proceed with the development of complete new STS in accordance with the Commission's Interim Policy Statement.
We will continue to interact with the IUMARC Technical Specification Working Group and each of the individual vendor Owners Groups as r-!ded to keep this, important program moving forward.Sincerely,.Or, i;.. d ltY" Thomas E. Kurley, Director Office of Nuclear Reactor Regulation
 
==Enclosure:==
 
As stated cc see next page b Or DISTRIBUTION:
DOEA R/F OTSB Kembers POR Central Files Murley/Sniezek ,TTKartin CERossi EJButcher AThadani LShao SAVarga DCrutchfleld JGPartlow JPStohr JWRoe FJMiraglia BABoger GCLainas FSchroeder JRichardson (W.S.WILGUS/LTR/SPLIT REPORT)CONCURRENCE:
*(see previous concurrence)-TSB:DOEA:hNRR
*TSB:NRR *C:TSB:NRR
*D:DOEA:NRR KDesai:pvic DCFischer EJButcher CERossi 4/18/88 04/19/88 04/20/88 04/22/88*D:DREP:NRR
#ADT:NRR " " JRStohr TiMartin EMrley.04/28/88 05/05/88 q5 /88*D:DEST:NRR AThadani 04/26/88*D:DEST:NRR LShao 04/26/18 C Hr. W. S. Wilgus cc w/encl:-3-Mr. Robert Gill B&W Owners Group P. 0. Box 33189 422 South Church Street Charlotte, North Carolina 28242 Mr. R. E. Bradley BWR Owners Group c/o Georgia Power Nuclear Operations Department 14th Floor 333 Piedmont Avenue Atlanta, Georgia 30308 Mr. Edward Lozito Westinghouse Owners Group c/o Virginia Power P. 0. Box 26666 Richmond, Virginia 23261 Mr. Joseph B. George Westinghouse Owners Group Texas Utilities 400 North Olive Dallas. Texas 75201 Mr. Stewart Webster CE Owners Group 1000 Prospect Hill Road Winstor. Connecticut 06095-0500 Mr. R. A. Bernier CE Owners Group c/o Arizona Nuclear Power Project P. 0. Box 52034 H.S. 7048 Phoenix, Arizona 85072 Mr. Thomas Tipton NUMARC 1776 Eye Street, N.W.Suite 300 Washington, D. C. 20006-2496 vA';90 NRC STAFF REVIEW OF NUCLEAR STEAM SUPPLY SYSTEM VENDOR OWNERS GROUPS'APPLICATION OF, THE COMM ISSION'S INTfRIM POLICY STATEMENT CRITERIA TO STANDARD TECHNICAL SPECIFICATIONS
: 1. INTRODUCTION On February 6, 1987, the Commission issued its Interim Policy Statement on Technical Specification Improvements (52 FR 3788). The Policy Statement encourages the Industry to develop now Standard Technical Specifications (STS)to be used as guides for licensees in preparing improved Technical Specifications (TS) for their facilities.
The Interim Policy Statement contains criteria (including a discussion of each) for determining which regulatory requirements and operating restrictions should be retained in the new STS and ultimately in plant TS. It also identifies four additional systems that are to be retained on the basis of operating experience and probabilistic risk assessments (PRA).Finally, the Policy Statement Indicates that risk evaluations are an appropriate tool for defining requirements'that should be retained in the STS/TS where Including such requirements is consistent with the purpose of TS (as stated in the Policy Statement).
Requirements that are not retained in the new STS would generally not be retained in Individual plant TS. Current TS requirements not retained in the STS will be relocated to other licensee-controlled documents.
One of the first steps in the program to implement the Commission's Interim Policy Statement is to determine which Limiting Conditions for Operation (LCOs)contained in the existing STS should be retained inthe new STS. An early decision on this issue will facilitate efforts to make the other improvements (described in the Policy Statement)*to the text and Bases of those requirements that must be retained in the new STS.Each Nuclear Steam Supply System (NSSS) vendor Owners Group has submitted a report to the NRC for review that identifies which STS LCOs the group believes should be retained In the new STS and which can be relocated to other licensee-controlled documents.
These four NSSS vendor submittals are as follows: (1) Letter dated October 15, 1987, R. L. 6ill, B&W Owners Group, to Dr. T. E. Murley, NRC,
 
==Subject:==
"B&I Owners Group Technical Specification Committee Application of Selection Criteria to the 5&W Standard Technical Specifications."* TOTAL PAGE.0l 1  (2) Letter dated November 12, 1987, R. A. Newton, Westinghouse Owners Group, to NRC Document Control Desk,
 
==Subject:==
OWestinghouse Owners Group MERITS Program Phase I1. Task 5, Criteria Application Topical Report.0 (3) Letter dated December 11, 1987, J. K. Gasper, Combustion Engineering Owners Group, to Dr. T. E. Hurley, NRC
 
==Subject:==
"CEN-355, CE Owners Group Restructured Standard Technical Specifications
-Volume 1 (Criteria Application)." (4) Letter dated November 12, 1987, R. F. Janecek, BWR Owners Group, to R. E. Starostecki, NRC,
 
==Subject:==
"BWR Owners Group Technical Specification screening Criteria Appllcbtion and Risk Assessment.*
These submittals provide the rationale for why each STS requirement (e.g.Limiting Condition for Operation) should be retained in the new STS or why it can be relocated to a licensee-controlled document.
They also describe how each Owners Group used risk insights in determining the appropriate content of the new STS.2. STAFF REVIEW The NRC staff focused its review on those requirements identified by the Owners Groups as candidates for relocation.
The staff evaluated each of these requirements to determine whether it agreed with the Owners Groups' conclusions.
During the NRC Staff's review, several issues were raised concerning the proper interpretation or application of the criteria in the Commission's Interim Policy Statement.
The NRC Staff has considered these issues and concluded the following:
(1) Criterion 1 should be interpreted to include only Instrumentation used to detect actual leaks and not more broadly to include instrumentation used 1 to detect precursors to an actual breech of the reactor coolant pressure boundary or instrumentation to identify the source of actual leakage (e.g., loose parts monitor, seismic instrumentation, valve position indicators).
(2) The "initial conditions" captured under Criterion 2 should not be limited to only "process variables" assumed In safety analyses.
They should also include certain active design features (e.g., high pressure/low pressure system valves and interlocks) and operating restrictions (e.g., pressure-temperature operating limit curves), needed to preclude unanalyzed accidents.
In this context. "actiye design features" include only design features under the control of operhtions personnel (i.e., licensed operators and personnel who perform control-junctions at the direction of licensed opera-tors). This position is consistent with the conclusions reached by the Staff during the trial application of the criteria to the Wolf Creek and Limerick Technical Specifications.
(3) The 'initial conditions" of design-basis accidents (DBA) and transients, as used in Criterion 2, should not be limited to only those directly *monitored and controlled" from the control room. Initial conditions should also in-clude other features/characteristics that are specifically assumed in DBA and transient analyses even if they can not be directly observed in the control room. For example, initial conditions (e.g., moderator temperature coefficient and hot channel factors) that are periodically monitored by other than licensed operators (e.g., core engineers, instrumentation and control technicians) to provide licensed operators with the information required to take those actions necessary to assure that the plant is being operated within the bcunds of design and analysis assumptions, meet Criterion 2 and should be retained in Technical Specifications.
Initial conditions do not, however, include things that are purely design requirements.
(4) The phrase "primary success path," used in Criterion 3, should be interpreted to include only the primary equipment (including redundant trains/components)
&#xb6;.,%entS I~d Primary success path does not include Owe 'r TtTPttjOy\
used to ,revEent aTnblyzed P. 02-40 accidents or transients or to improve reliability of themitigation function (e.g., rod withdrawal block which is backup to the average power range monitor high flux trip in the startup mode, safety valves which are backup to low temperature over pressure relief valves during cold .shutdown).
(5) Post-Accident Monitoring Instrumentation that satisfies the definition of Type A variables in Regulatory Guide 1.97, "Instrumentation for Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accldent,u meets Criterion 3 and should be retained in Technical Specifications.
Type A variables provide primary information (i.e., Information that. i essential for the direct accomplishment of the specified manual actions (incboding long-term recovery actions) for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for OBAs or transients).
Type A variables do not include those variables associated with contingency actions that may also be identified in written procedures to compensate for failures of primary equipment.
Because only Type A variables meet Criterion 3, the STS should contain a narrative statement that indicates that individual plant Technical Specifications should contain a list of Post-Accident Instrumentation that includes Type A variables.
Other Post-Accident Instrumentation (i.e., non-Type A Category 1) Is discussed on page 6.(6) The HAC's design basis for licensing a plant is the plant's Final Safety Analysis Report (FSAR) as qualified by the analysis performed by the staff and documented in the staff's safety evaluation report (SER). Because the staff's review and resulting SER are based on the acceptance criteria in the NRC's Standard Review Plan (NUREG-0800, SRP), the dose limits used in licensing a particular plant may be 'some small fraction of those specified in the Commission's regulations In 7itle 10 of the Code of Federal Regulations Part 100 (10 CFR 100). Accordingly, the SRP limits should be used to define the equipment in the primary success path for mitigating accidents and transients when developing the new STS. These types of conservatisms are required to compensate for uncertainties in analysis techniques and  provide reasonable assurance that the absolute numerical limits of the regulations will be satisfied.
On a plant-specific basis, systems and equipment that are identified in the NRC staff SER and assumed by the staff to function are considered part of the licensing basis for the plant and are captured by Criterion 3 (e.g., radiation monitoring Instrumentation that initiates an isolation function, penetration room exhaust air cleanup system).(7) DBA and transients, as used in Criteria 2 and 3, should be interpreted to include any design-basis event described in the FSAR (i.e., not just those events described in Chapters 6 and 15 of the FSAR). For example, there may be requirements for some plants which should be retained in Technical Specifications because of the risks associated with some site-specific characteristic (e.g., although not normally required, a Technical Specifi-cation on the chlorine detection system might be appropriate where a sig-nificant chlorine hazard exists in the site vicinity; similarly, a Tech-nical Specification on flood protection might be appropriate where a plant is particularly vulnerable to flooding and is designed with special flood protection features).
Criteria 2 and 3 should not be interpreted to in-clude purely generic design requirements applicable to all plants (e.g., the requirements of General Design Criterion 19 in Appendix A to 10 CFR Part 50 for control room design).The NRC staff has used the Commission's Interim Policy Statement and the conclusions described above to define the appropriate content of the new STS.The staff plans to factor these conclusions into' the Final Policy Statement on Technical Specification Improvements that will be proposed to the Commission.
The staff. reviewed the methodology and results provided by each Owners Group to verify that none of the requirements proposed for relocation contains constraints of prime importance in limiting the likelihood or severity of accident sequences that are coumonly found to dominate risk. For the purpose  of this application of the guidance in the Commission Policy Statement, the staff agrees with the Owners Groups' conclusions except in two areas. First, the staff finds that the Remote Shutdown Instrumentation meets the Policy State-ment criteria for inclusion in Technical Specifications based on risk; and second, the staff is unable to confirm the Owners Groups' conclusion that Category 1 Post-Accident Monitoring Instrumentation is not of prime importance in limiting risk. Recent PRAs have shown the risk significance of operator re-covery actions which would require a knowledge of Category I variables.
Furthermore, recent severe accident studies have shown significant potential for risk reduction from accident management.
The Owners Groups' should develop further risk-based justification tn support of relocating any or all Category 1 variables from the Standard Technieal Specifications.
As stated in the Commission's Interim Policy Statement, licensees should also use plant-specific PRAs or risk surveys as they prepare license amendments to adopt the revised STS to their plant. Where PRAs or surveys are available, licensees should use them to strengthen the Bases as well as to screen those Technical Specifications to be relocated.
Where such plant-specific risk surveys are not available, licensees should use the literature available on risk insights and PRAs. Licensees need not complete a plant-specific PRA before they can adopt the new STS. The NRC staff will also use risk insights and PRAs in evaluating the plant-specific submittals.
: 3. RESULTS OF THE STAFF'S REVIEW Appendices A through D present the detailed results of the staff's review of the Babcock and Wilcox, Westinghouse, Combustion Engineering, and-General Electric application of the selection criteria to the existing STS. Each Appendix con-sists of two tables. Table 1 identifies those LCUs that must be retained in the new STS. Table 2 lists those LCOs that may be wholly or partially relocated to licensee-controlled documents (or be reformatted as a surveillance requirement for another LCO). Where the staff placed specific conditions on relocation of particular LCOs the staff has so noted In the Tables. As a part of the  plant specific implementation of the new STS, the staff plans to review the location of. and controls over, relocated requirements.
In as much as practi-cable, the Owners Groups should propose standard locations for, and controls over, relocated requirements.
For each LCO listed in Table 1, the criterion (criteria) that required that the LCO be retained in Technical Specifications is identified.
If an LCO was retained in Technical Specifications solely on the basis of risk, "Risk" appears In the criteria column. Where an Owners Group determined that an LCO had to stay in 7echnical Specificatins (because of either a particular criterion or risk) and the Staff agreed that the LCO should be retained in Technical Specif-Ications, the staff did not, In giberal, verify the Owners Group's basis for retention.
However, in several instances the Owners Groups cited risk consider-ations alone as the basis for retaining Technical Specifications and the staff disagreed with the Owners Groups. In these instances, the staff's basis for retention appears in the criteria column of Table 1.Any LCO not specifically Identified in Table 1 or Table 2 (e.g., an LCO unique to an STS not addressed in the Owners Groups submittals such as the BWR5 STS)should be retained In the STS until the Owners Group proposes and the staff makes a specific determination that it can be relocated to a licensee-controlled document.Notwithstanding the results of this review, the staff will give further consideration for relocation of additional LCOs as the staff and industry proceed with the development of the new SIS.4. CONCLUSION The results of the effort of the Owners Groups and of the NRC staff to apply the Policy Statement selection criteria to the existing STS are an important step toward ensuring that the new STS contain only those requirements that are consistent with 10 CFR 50.36 and have a sound safety basis. As shown in the  following tables, application of the criteria contained in the Commission's Interim Policy Statement resulted in a significant reduction In the number of LCOs to be included in the new STS. The development of the new STS based on the staff's conclusions will result in more efficient use of NRC and industry resources.
Safety improvements are expected through more operator-oriented Technical Specifications, improved Technical Specification Bases, a reduction in action statement-Induced plant transients, and a reduction in testing at power.-.BABCOCK&wilcox LCOs Total Number Retained Relocated Percent Relocated WESTINGHOUSE 165 92 137 75 62 COMBUSTION ENGINEERING 159 87 72 45%GENERAL ELECTRIC B 4/BWR6 124/144 81/86 ,43/58 73 45%44%35%/40%We' are confident that the staff's conclusions will provide an adequate basis for the Owners Groups to proceed with the development of complete new STS in acccrdance with the Commission's Interim Policy Statement.
APPENDIX.
A RESULTS OF'THE NRC STAFF REVIEW BABCOCK &. WILCOX OWNERS GROUP'S SUBMITTAL RE RETENTION AND RELOCATION J)F SPECIFIC TECHNICAL SPECIFICATIONS APPENDIX A TABLE I LCOs TO BE RETAINED IN BABCOCK & WILCOX STANDARD TECHNICAL SPECIFICATIONS LCO 3.1 3.1.1.1 3.1.1.2 3.1.1.3 3.1.3.1 3.1.3.2 3.1.3.6 3,1.3.7 3.1.3.9 3.2 3.2.1 3.2.2 3.2.3 3.2.4 3.2.5 3.3 CRITERIA REACTIVITY CONTROL SYSTEM Shutdown Margin (Note 1)Moderator Temperature Coefficient Minimum Temperature for Criticality Group Height -Safety and Regulating Rod Groups Group Height -Axial Power Shaping Rod Group Safety Rod Insertion Limit Regulating Rod Insertion Limits Xenon Reactivity
: 2 2 2 2 2 2 &3 2 2 POWER DISTRIBUTION LIMITS Axial Power Imbalance Nuclear Heat Flux Hot Channel Factor Nuclear Enthalpy Rise Hot Channel Factor Quadrant Power Tilt DhNB Parameters 2 2 2 2 2 INSTRUMENTATION 3.3.1 3.3.2 3.3.3.1 3.3.3.5 3.3.3.6 3.4 Reactor Protection System Instrumentation (Note 2)Engineered Safety Feature Actuation System Instrumentation (Note 2)Radiation Monitoring Instrumentation (Notes 2 A 3)Remote Shutdown Instrumentation (Notes 2 & 4)Accident Monitoring Instrumentation REACTOR COOLANT SYSTEM 3 3 3 Risk 3 3.4.1.1 3.4.1.2 3.4.1.3 3.4.1.4 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7.1 Startup and Power Operation Hot Standby Hot Shutdown Cold Shutdown Safety Valve -Operating Pressurizer Relief Valve Steam Generators
-Water Level Leakage Detection System 3 3 3 Policy Statement (DHR)3 2&3 3 2 1 A-1 B&W-TABLE I (Continued)
LCO 3.4.7.2 3.4.9 3.4.10.1 3.4.10.3 3.5 3.5.1 3.5.2 3.5.3 3.5.4 3.6 3.6.1.1 3.6.1.3 3.6.1.5 3.6.1.6 3.6.1.8 3.6.2.1 3.6.2.2 3.6.2.3 3.6.3 3.6.4 3.6.5.1 3.6.5.2 3.6.6 3.7 3.7.1.1 3.7.1.2 3.7.1.3 3.7.1.4 3.7.1.5 3.7.3 3.7.4 3.7.5 3.7.6 3.7.7 3.7.8 Operational Leakage Specific Activity Reactor Coolant System Pressure/Temperature Limits Overpressure Protection System EMERGENCY CORE COOLING SYSTEM (ECCS)Core Flooding Tanks ECCS Subsystems
-Tavg 305)F ECCS Subsystems
-Tvg(305)&deg;F Borated Water Storage Tank CONTAINMENT SYSTEMS Containment Integrity Containment Air Locks Internal Pressure Air lemperature Containment Ventilation System Containment Spray System Spray Additive System Containment Cooling System Iodine Cleanup System Containment Isolation Valves Hydrogen Analyzers Electric Hydrogen Recombiners (Note 5)Penetration Room Exhaust Air Cleanup System PLANT SYSTEMS Safety Valves Auxiliary Feedwater System Condensate Storage Tank Activity Main Steam Line Isolation Valves Component Cooling Water System Service Water System Ultimate Heat Sink Flood Protection (optional)
Control Room Emergency Air Cleanup System ECCS Pump Room Exhaust Air Cleanup System (optional)
CRITERIA 2 2 2 2 &3 3 3 2 &3 3 3 2 2 3 3 2 &3 3 3 3 3 3 3 3 3 2&3 2 3 3 3 3 3 3 3 A-2 LCO 3.8 B&W-TABLE 1 (Continued)
ELECTRICAL POWdER SYSTEMS CRITERIA 3.8.1.1 3.8.1.2 3.8.2.1 3.8.2.2 3.8.2.3 3.8.2.4 A. C.A. C.A.C.A.C.D.C.D. C.Sources -Operating Sources -Shutdown Distribution
-Operating Distribution
-Shutdown Distribution
-Operating Distribution
-Shutdown 3 Policy Statement 3 Policy .Statement 3 Policy Statement (DHR)(OHR)(DHR)3.9 REFUELING OPERATIONS 3.9.1 3.9.2 3.9.3 3.9.4 3.9.8.1 3.9.8.2 3.9.9 3.9.10 3.9.11 3.9.12 Boron Concentration Instrumentation Decay Time .v Containment Buitding Penetration Residual Heat RemovAl and Coolant Circulation
-All Water Levels Policy Residual Heat Removal and Coolant Circulation
-Low Water Levels Policy Containment Purge and Exhaust Isolation System Water Level -Reactor Vessel Water Level -Storage Pool Storage Pool Air Cleanup System 2 3 2 3 Statement (DHR)Statement (DHR)3 2 2 2 Notes: 1. Required for Modes 3 through 5. May be relocated for Modes 1 and 2.2. The LCO for this system should be retained in STS. The Policy Statement criterie should not be used as the basis for relocating specific trip functions, channels, or instruments within these LCOs.3. The staff is pursuing alternative approaches which would allow relocation of some of these LCOs on a schedule consistent with the schedule for development of the new STS. The staff Is also initiating rulemaking to delete the requirement that RETS be included in Technical Specifications.
: 4. Because fires (either inside or outside the control room) can be a significant contributor to the core melt frequency and because the uncertainties with fire initiation frequency can be significant, the staff believes that this LCO should be retrained in the STS at this time. The staff will consider relocation of Remote Shutdown Instrumentation on a plant-specific basis.S. This LCO will be'considered for relocation to a licensee-controlled document on a plant-specific basis.A-3 LCO 3.1 3.1.2.1 3.1.2.2 3.1.2.3 3.1.2.4 3.1.2.5 3.1.2.6 3.1.2.7 3.1.2.8 3.1.2.9 3.1.3.3 3.1.3.4 3.1.3.5 3.1.3.8 3.3 3.3.3.2 3.3.3.3 3.3.3.4 3.3.3.7 3.3.3.8 3.3.3.9 3.3.3.10 3.3.4 3.4 3.4.2 3.4.6 3.4.8 3.4.10.2 3.4.11 3.4.12 3.6 3.6.1.2 3.6.1.7 3.7 3.7.2 3.7.9 3.7.10 TABLE 2 (Note 1)BABCOCK & WILCOX STANDARD TECHNICAL SPECIFICATION LCOs WHICH MAY BE RELOCATED REACTIVITY CONTROL SYSTEMS Flow Paths -Shutdown Flow Paths -Operating Makeup Pump -Shutdown Makeup Pump -Operating Decay Heat Removal Pump -Shutdown Boric Acid Pumps -Shutdown Boric Acid Pumps -Operating Borated Water. ource -Shutdown Borated Water Sburce -Operating Position Indicatiof Channels -Operating (Note 2)Position Indication-Channels Shutdown (Note 2)Rod Drop Time (Note 2)Rod Program INSTRUMENTATION Incore Detectors Seismic Instrumentation Meteorological Instrumentation Chlorine Detection System Fire Detection Radioactive Liquid Effluent Monitor (Note 3)Radioactive Gaseous Effluent Monitor (Note 3)Turbine Overspeed Protection REACTOR COOLANT SYSTEM Safety Valves -Shutdown Steam Generators Tube Surveillance (Note 4)Chemistry Pressurizer Temperatures Structural Integrity ASME Code (Note 4)RCS Vents CONTAINMENT SYSTEMS Containnent Leakage (Note 5)Containment Structural Integrity (Note 2)PLANT SYSTEMS Steam Generator Pressure/Temperature Limits Snubbers Sealed Source Contamination
(
B&W-TABLE 2 (Continued)
LCO 3.7.11.1 Fire Suppression Water System 3.7.11.2 Spray and/or Sprinkler Systems 3.7.11.3 CO System 3.7.11.4 Haion System 3.7.11.5 Fire Hose Stations 3.7.11.6 Yard Fire Hydrants and Hydrant Hose Houses 3.7.12 Fire Barrier Penetrations 3.7.13 Area Temperature Monitoring 3.9 REFUELING OPERATIONS
 
====3.9.5 Communications====
3.9.6 Fuel Handling Bridge 3.9.7 Crane Travel ;pent Fuel Storage Pool Building 3.10 SPECIAL TEST EXCEPTIONS 3.10.1 Shutdown Margin (Note 6)3.10.2 Group Height Insertion Limits and Power Distribution Limits (Note 6)3.10.3 Physics Tests (Note 6)3.10.4 Reactor Coolant Loops (Note 6)3.11 RADIOACTIVE EFFLUENTS (Note 3)3.11.1.1 Concentration 3.11.1.2 Dose 3.11.1.3 Liquid Radwaste Treatment System 3.11.1.4 Liquid Holdup Tanks 3.11.2.1 Dose 3.11.2.2 Dose -Noble Gases 3.11.2.3 Dose -Iodine -131, Tritium and Radionuclides In Particulate Form 3.11.2.4 Gaseous Radwaste Treatment Systems 3.11.2.5 Explosive Gas Mixture 3.11.2.6 Gas Storage Tanks 3.11.3 Solid Radioactive Waste 3.11.4 Total Dose 3.12 RADIOACTIVE ENVIRONMENIAL MONITORING (Note 3)3.12.1 Monitoring Program 3.12.2 Land Use Census 3.12.3 Interlaboratory Comparison Program A-5 B&W-TABLE 2 (Continued)
Notes: 1. Specifications listed in this table may be relocated contingent upon NRC staff approval of the location of and controls over relocated requirements.
: 2. This LCO may be removed from the STS. However, if the associated Surveillance Requirement(s) is necessary to meet the OPERABILITY requirements for a retained LCO, the Surveillance Requirement(s) should be relocated to the retained LCO.3. The staff is pursuing alternative approaches which would allow relocation of some of these LCUs on a schedule consistent with the schedule for develop-ment of the new STS. The staff is also initiating rulemaking to delete the requirement that RETS be included in Technical Specifications.
: 4. This LCO may-be relocated.opjt of Technical Specifications.
However, the associated Surveillance Reqbirement(s) must be relocated to Technical Specification Section 4.0, Survpillance Requirements.
: 5. This LCO may be relocated.
However, Pa, La, Ld, and Lt must be either retained in TS or in the Bases of the appropriate Containment LCO.6. Special Test Exceptions may be included with corresponding LCOs.
Section 3.8 RAIs NRC ITS Tracking Page I of 2 Return to View Menul "Pint Documenti RAI Screening Required:
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NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC Ilovpviwpr ID[ 200710091446 Conference Call Requested?
No Category In Scope ITS Section: TB POC; JFD Number: Page.Number(s).:
ITS 3.8 Robert Clark None 7 Information ITS Number: OSI: DOC.Number:
Bases JFD Number: 3.8.1 None M.3 None NRC Author Robert Clark Regulatory Bases: 10CFR50.36(c)(3), GDC 18 Comment Required Action: ITS SR 3.8.1.3 and SR 3.8.1.13 specified that the EDG be loaded between 90 and 100% of its continuous rating. Provide confirmation that this load range will bound the worst case DBA load.Issue :Date 10/09/2007 Close :D: at e] 11/16/2007 Logged in User: Jerry Jones-Responses Licensee Response by Jerry Jones on 10/18/2007 ITS SRs 3.8.1.3 and SR 3.8.1.13 (Volume 13, Pages 37 and 45, respectively) verify the EDG can be loaded to between 90% and 100% of the continuous rating of the EDG. 90% of the continuous rating of the Davis Besse EDGs is 2340 kW and 100% is 2600 kW. The 90% and 100% values were chosen for the Davis Besse ITS SRs to be consistent with the load values specified in Regulatory Guide 1.9, Rev. 3, Sections 2.2.2 (for SR 3.8.1.3) and 2.2.9 (for.SR 3.8.1.13).
Both of these new load values are in excess of the currently required test values in CTS 4.8.1.1.2.a.5 (Page 6) and CTS 4.8.1.1.2.d.3 (Page 8). The maximum expected accident load for EDG 1-1 is 2322 kW, which is less than the 90%value. However, the maximum expected accident value for EDG 1-2 is 2384 kW, which is slightly above the 90% value http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 (approximately 91.7%). Davis Besse believes that the proposed 90% to 100% load range is acceptable since: a) it is consistent with Regulatory Guide 1.9, Rev. 3; b) the EDGs are tested every 24 months at a value well in excess of the maximum expected accident value during performance of SR 3.8.1.13 (the 2 hour portion of the test is performed at 105% to 110% of the continuous rating of the EDGs; c) the proposed minimum value forSR 3.8.1.3 is just slightly lower than the maximum expected accident load for only one of the two EDGs; and d) the new proposed values are much greater than what the current Technical Specifications require for the same Surveillances.
NRC Response by Robert Clark ][No further questions at this time. Item closed.on 11/16/2007
____________________________
Date Created: 10/09/2007 02:46 PM by Robert Clark Last Modified:
11/16/2007 09:09 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2~sigI4&#xfd; Return to View Menu Pr~int Document RAI Screening Required:
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ITS 3.8 Robert Clark 7 35 Information ITS Nuniber: OR.: D!OC Numnber: Bases JFD-Number:
3.8.1 None None None INCAuthor Robert Clark Revise LCO 3.8.1, Condition G, to read:- "One or more trains with one load sequencer inoperable." Revise LCO 3.8.1, Condition H, to read: "One or more trains with two load sequencers inoperable." STS 3.8.1 Condition F was modified to be consistent with Davis-Besse licensing basis. The Justification for Deviation (JFD #7) states that CTS Table 3.3-3 Comment Action 15a provides the Actions for when one load sequencer per bus is inoperable and requires the sequencer to be removed within 1 hour. However, proposed ITS LCO 3.8.1, Condition G, does not clearly refer to one load sequencer per train. Condition H should use the word "load sequencers" to be consistent with Condition G.Per 1OCFR50.36(c)(2) the Tech Specs shall contain LCO's which provide remedial actions that shall be taken until the condition is met.Issue Date 10/10/2007 Close Date [11/1 5/2 0 07 Logged in User: Jerry Jones'Responses Licensee Response by Jerry Jones on 10/18/2007 The first requested action was to revise LCO 3.8.1, Condition G, to read: "One or more trains with one load sequencer inoperable." A review of the ISTS Markup for LCO 3.8.1, Condition G, shows that it already reads exactly as requested.
The second requested action was to revise LCO 3.8.1, Condition H, to read "One or more http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 trains with two load sequencers inoperable." The request is to change the word "sequencers" to "load sequencers." This request is appropriate.
A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Robert Clark No further questions at this time. Item closed.on 11/15/2007
__Date Created: 10/10/2007 03:12 PM by Robert Clark Last Modified:
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6/2/2008 NRC ITS Tracking Page I of 2 FWjs~~ ,/';Return to View Menu] Print Docuen RAI Screening Required:
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ITS 3.8 Robert Clark None 20 Information ITS Number: OSI: DOC Number: Bases JFD Number: 3.8.1 None M.5 None NRC A or] Robert Clark Revise DOC M5 to include'reference to M8 not ML. EDG load rejection surveillance CTS 4.8.1.1.2.d.1 (ITS SR 3.8.1.10) was revised to require that the EDG operate at the power factor limit when synchronized with offsite power.In addition to including the power factor limit, M5 referenced M1 for other Comment changes. Ml addresses changes to the CTS that are related to EDG steady state voltage and frequency for normal and fast starts. M5 should reference M8 which discuss the EDG frequency acceptance criterion following load rejection.
10CFR50.36(c)(3) requires Surveillance Requirements to verify that the LCO's are met.lssue Date] 10/10/2007 Close [11/15/2007 Logged in User: Jerry Jones'Responses Licensee Response by Jerry The requested action was to revise Discussion of Change (DOC)Jones on 10/18/2007 M05 (Volume 13, Page 20) to reference DOC M08 instead of the currently referenced DOC MO1 (in the last sentence of the first paragraph.
This request is appropriate.
A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Robert Clark INo further questions at this time. Item closed.http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 I on 11/15/2007 ii 11 I.Date Created: 10/ 10/2007 05: 10 PM by Robert Clark Last Modified:
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6/2/2008 NRC ITS Tracking Page I of 2 F&#xfd;g N ew Res~ponse.
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ITS 3.8 Robert Clark 3 44 Information ITS-Number:
OSI: DO C Number: Bases JFD Number: 3.8.1 None None None NRC Author] Robert Clark CTS SR 4.8.1.1.2.d.2.(c) specifies that the noncritical trips (overspeed, generator differential) are automatically bypassed upon loss of voltage on the essential bus "and/or" an SFAS signal. ITS SR 3.8.1.12 specifies EDG Comment automatic trips are bypassed on actual or simulated loss of voltage on the essential bus "or" and actual or simulated SFAS signal. Please clarify if both trip signals will be applied when performing this surveillance.
Note: ISTS SR 3.8.1.13 requires that the trip signals be concurrent.
Issue. Date ]110/19/2007 Close Date i 1 1/1 5/2 0 0 7 Logged in User: Jerry Jones"'Responses Licensee Response by Bill Bentley on 10/19/2007 CTS SR 4.8.1.1.2.d.2,(c) states "Verifying that all diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the essential bus and/or an SFAS signal." Therefore, engine overspeed and generator differential are not the non critical trips. As shown in the text associated with SR 3.8.1.12, Volume 13, page 87 of 323, engine overspeed and generator differential current are not included as part of the non critical trips. Engine overspeed and generator differential current are not bypassed on a safety start of the EDG's. A safety start results from either loss of voltage oran http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/
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6/2/2008 NRC ITS Tracking Page 2 of 2 SFAS signal. The "concurrent" terminology was changed in the ISTS 3.8.1.13 because it does not match the Davis-Besse design. It only takes one or the other condition.
The conditions do not have to be concurrent.
NRC Response by Robert Clark o further questions at this time. Item closed.[on 11/15/2007Ilo qetosathstm.tecoed Date Created: 10/19/2007 02:27 PM by Robert Clark Last Modified:
11/15/2007 09:16 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
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ITS 3.8 Robert Clark 15 43 Information ITS..N.mber:
OSI.: DOC Number: BasesJFD...Number:
3.8.1 None None None NRC Authojr Robert Clark JFD 15 is applicable to ISTS SR 3.8.1.12 as noted on page 43. However, the marked up ISTS also referenced JFD 15 for SR 3.8.1.13 (page 44), SR 3.8.1.13 (page 45), SR 3.8.1.14 (page 46), and SR 3.8.1.19 (page 48) which are not applicable.
Issue _Date 110/19/2007 Close Daate F11/1 5/2 0 0 7 Logged in User: Jerry Johes'Responses Licensee Response by Bill The last sentence of JFD #15 states "subsequent surveillances have Bentley on 10/19/2007 been renumbered, as applicable." The JFD #15 markings on page 44, 45, 46, and 48 only apply to this renumbering.
NRC Response by Robert Clark No further questions at this time. Item closed.on 11/15/2007
_Date Created: 10/19/2007 02:31 PM by Robert Clark Last Modified:
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ITS 3.8 Robert Clark 9 47 Information TS .Number: OI: DOC Number: Bases.,JFD Number: 3.8.1 None None None NRC Author Robert Clark Coninent Please explain why the Note for ISTS SR 3.8.1.18 does not apply to the DB.o.e.... load sequencers and emergency time delay relays.Issue Date 10/19/2007 Cllose-Date
[02/12/2008 Logged in User: Jerry Jones'Responses Licensee Response by Bill ISTS SR 3.8.1.18, was changed to SR 3.8.1.6, and is identical to Bentley on 10/19/2007 the CTS surveillance requirement 4.8.1.1.2.a.7.
This surveillance requirement is currently performed on a monthly basis, and that same frequency was applied to SR 3.8.1.6. The note would restrict this surveillance from being performed in Modes 1 -4. Since the surveillance is currently performed in Mode 1 on a monthly basis, the note does not apply.NRC Response by Robert Clark Your response states that the current surveillance for the load on 11/15/2007 sequencer and emergency time delay relays are performed in Mode 1 on a monthly basis. Please provide additional information or basis as to why the Note for STS SR 3.8.1.18 is not applicable to 1313DB.Licensee Response by Bill Bentley on 01/25/2008 The 31 day frequency has been in place since the beginning of plant operation and has been routinely performed with the plant in http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 mode 1. The Sequencer Functional Unit in the particular SFAS Channel is tested as part of the monthly functional test for a particular SFAS Channel. The load sequencer is declared inoperable during testing of the load sequencer.
In addition to the equipment actuated by SFAS, ITS SR 3.8.1.6 also includes the emergency time delay relays for the Makeup Pumps. These relays cause the operating Makeup pump to trip after a loss of offsite power, and to restart 2.5 seconds after the associated EDG output breaker closes. These relays are separate from the SFAS Load sequencers, are tested separately, andwere added as a modification to the plant several years ago: The testing of the Make Up Pump 1 Time Delay Relay renders EDG 1 INOPERABLE if Make Up Pump I is running. The testing of the Make Up Pump 2 Time Delay Relay renders EDG 2 INOPERABLE if Make Up Pump 2 is running. There is no override capability if a valid signal is received during testing. USAR Table 8.3-1 contains information related to the sequential loading of the EDGs.NRC Response by Robert Clark j[No further questions at this time. Item closed.o 2 Date Created: 10/19/2007 02:32 PM by Robert Clark Last Modified:
02/12/2008 02:29 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
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ITS 3.8 Robert Clark None 77 Information ITS Number: OS0.1:. DOC Number: Bases,.,JFD Nu.Amber: 3.8.1 None None 10 NRC Author Robert Clark The Bases for SR 3.8.1.9 should explain why Mode 1 or 2 restrictions are not Co mment applicable to SR 3.8.1.9.a i.e., verify automatic and manual transfer of AC power sources from the unit auxiliary source to the pre-selected offsite circuit.Issue Date 10/22/2007 Close Date [ 03/17/2008 Logged in User: Jerry Jones'Responses Licensee Response by Jerry Jones on 10/25/2007 ITS SR 3.8.1.9 (Volume 13, Page 38) is modified by a Note that states SR 3.8.1.9.b shall not normally be performed in MODES 1 and 2. The Note explanation in the Bases for ITS SR 3.8.1.9 (Page 78) explains why this Note is applicable to SR 3.8.1.9.b.
The format of the ISTS does not include explaining why Notes are not applicable to other SRs. For example, ISTS SR 3.8.1.15 (Page 46)is required to be performed at a similar Frequency (18 months) as ITS SR 3.8.1.9, but does not include a MODE restriction Note.Similarly, the ISTS SR 3.8.1.15 Bases do not include a reason why the SR can be performed in MODES 1 and 2. Furthermore, SR 3.8.1.9.a is a test of the 13.8 kV bus transfers between the unit auxiliary source and the startup transformers.
The unit auxiliary source is normally powered by the main generator.
It is part of normal plant operations to perform manual transfers between these http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 sources in MODE 1. For example, when performing a plant startup, once the main generator is producing output at a specified power level, the 13.8 kV buses are manually transferred from the Startup Transformers to the unit auxiliary source. Thus, explaining why the MODES I and 2 restriction is not applicable to SR 3.8.1.9.a is not necessary.
NRC Response by Robert Clark on 11/15/2007 The Bases for SR 3.8.1.9 explains why performance of SR 3.8.1.9.b (verify automatic and manual transfer of AC power from the normal offsite circuit to the alternate offsite circuit) is restricted in MODE 1 or 2. However, the Bases does not explain why performance of SR 3.8.1.9.a (verify automatic and manual transfer of AC power from the unit auxiliary source to the pre-selected offsite circuit) is not restricted in MODE 1 or 2. The Bases should clarify why Note 2 for SR 3.8.1.9 is not applicable to SR 3.8.1.9.a.
Licensee Response by Bryan Kays on 03/16/2008 After further review, Davis-Besse has determined that Note 2 (Volume 13, Page 38) should apply to the automatic transfer portion of SR 3.8.1.9.a in addition to 3.8.1.9.b.
Therefore, Note 2 for SR 3.8.1.9 (Page 32) has been changed to be applicable to the automatic portion of SR 3.8.1.9.a and all of SR 3.8.1.9.b.
Additionally, Discussion of Change L08 (Page 28), Justification for Deviation 20 (Page 53) and the Bases for SR 3.8.1.9 (Page 78)have also been revised. A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Robert Clark on 03/17/2008 No further questions at this time. Item closed.Date Created: 10/22/2007 02:17 PM by Robert Clark Last Modified:
03/17/2008 01:07 PM http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/lfddcealOd3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2[ ssig ] Return to View Menu] 4 Print Documient RAI Screening Required:
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TB POC.:. JFD ..Number:. Page Number(s)*:
ITS 3.8 Robert Clark None 109 Information ITS Number: OS.I.: DOC. Number: Bases- JFD Number: 3.8.1 None L.2 None I_ C Athor[Robert Clark Provide justification for deleting the suspension of CORE ALTERATIONS from CTS 3.8.1.2.CTS 3.8.1.2 (ITS 3.8.2 A.2.1, B.1) deleted the requirement to suspend CORE ALTERATIONS upon loss of AC power until the minimum required AC C ommePqnt!A sources are restored to OPERABLE status. The DOC for L02 should supplement the justification for the TS change by referencing TSTF-471,"Eliminate use of term CORE ALTERATIONS in ACTIONS and Notes." Per 10CFR50.36(c)(2) the Tech Specs shall contain LCO's which provide remedial actions that shall be taken until the condition is met.IssueDate 10/24/2007 Close Date[ 11/27/2007 Logged in User: Jerry Jones'Responses Licensee Response by Bill Bentley on 10/26/2007 Discussion of Change (DOC) L02 (Volume 13, page 109) provides the justification for deleting the requirement to suspend "CORE ALTERATIONS" upon loss of the required AC sources. The justification provided is consistent with the justification provided for deleting the requirement in the Improved Standard Technical Specifications (ISTS), as documented in TSTF-471.
However, Davis-Besse has not included any references to the ISTS as part of the justifications for changing the CTS to the ITS, except for that http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1 fddcea 1 Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 provided in the generic DOC A01. Basically, almost every single change to the CTS could include a statement such as "this change is consistent with NUREG-1430.".
Since this type of statement does not appear to add any further justification, it has not been added. In addition, the ISTS markup (Page 115) clearly shows that_ _ _ __ the requirement has been deleted from the ISTS by TSTF-471.NRC Response by Robert Clark No further questions at this time. Itemclosed.
on 11/27/2007]__
Date Created: 10/24/2007 10:52 AM by Robert Clark Last Modified:
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6/2/2008 NRC ITS Tracking Page I of 2 sign, .. Return to View Men]u a.Print Document RAI Screening Required:
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No catego[y BSI -Beyond Scope Issue ITS.. TB POC: JFD Number: Page Number(s):
ITS 3.8 Robert Clark Vijay Goel None 21 Information ITS Number: OSI: DO.C.Number:
Bases.. JFD Number: 3.8.1 None M.8 None NRC Author] Vijay Goel Regulatory Bases: 10CFR50.36(c)(3), GDC 18"The purpose of ISTS SR 3.8.1.9.b and c limits is to ensure the EDG is ready to take on the next safety load quickly (within the designed voltage and Comment frequency) if the largest load gets tripped off due to some reason during load sequencing.
The TS of many plants comply to the requirements of ISTS SR 3.8.1.9.b and c limits. Explain why it is not possible for Davis-Besse to comply to the requirements of ISTS SR 3.8.1.9.b and c limits." Issue Date] 12/17/2007 Close Date] [01/14/2008 Logged in User: Jerry Jones'Responses Licensee Response by Bryan Kays on 01/10/2008 CTS 4.8.1.1.2.d.
I (Volume 13, Page 8) requires verification that each EDG can reject a load equivalent to the largest single emergency load without tripping the EDG. ITS 3.8.1.10 directly reflects this CTS requirement.
Furthermore, as stated in Justification for Deviation (JFD) 13 (Page 51), the Davis-Besse EDGs were designed to meet NRC Safety Guide 9, dated March 10, 1971. Safety Guide 9 did not include limits on the steady state frequency and voltage following a single load rejection (i.e., the EDGs were not originally designed to meet these requirements).
These new requirements are now part of Regulatory Guide 1.9, http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/
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6/2/2008 NRC ITS Tracking Page 2 of 2 Rev. 3 (essentially Safety Guide 9, Rev. 3). Furthermore, this requirement is a design requirement, not a requirement assumed in the safety analysis.
In addition, this deletion of ISTS 3.8.1.9.b and c have been approved by the NRC in the ITS conversion for James A. Fitzpatrick Nuclear Power Plant and the Monticello Nuclear Generating Plant. Therefore, Davis-Besse believes that no change is necessary.
Furthermore, since this is consistent with the Davis-Besse current licensing basis, it is not a beyond scope issue.NRC Response by Vijay Goel on 1 EEEB agrees that this issue is not an BSI, and has no further 01/14/2008 question at this time.Date Created: 12/17/2007 03:14 PM byVijay Goel Last Modified:
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ITS 3.8 Robert Clark None 62 Information ITS..-Nu.m!be-r: .OS:. DOC Numfl belr: Bas.es..J5,HD Numbher: 3.8.1 None None 12 NRC Robert Clark The Bases for Required Actions 3.8.1, A.2, B.2, and C.2 where modified in accordance with TSTF 402-T to state: "These redundant required features are those that are assumed to function to mitigate an accident, coincident with a loss of offsite power, in the safety analyses, such as the Emergency Core Cooling System and Auxiliary Feedwater System. These redundant required features do not include monitoring requirements, such as Post Accident Comment Monitoring and Remote Shutdown." The staff does not believe the proposed Bases changes are necessary because LCO 3.0.2 requires, an evaluation be performed in accordance with STS 5.5.14, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program (e.g., due to loss of redundant required features), the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.Issuej Date 1j02/04/2008 Close Date 103/14/2008 Logged in User: Jerry Jones'Responses Licensee Response by Jerry Jones on 02/13/2008 Davis-Besse believes the additional information added to the Bases for ITS 3.8.1 Required Action A.2 (Volume 13, Page 62), Required Action B.2 (Page 64), and Required Action C.2 (Page 66) is necessary.
As stated in TSTF-402-T, the existing Bases do http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/l fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 3 not provide sufficient guidance in determining which systems are, and are not, considered "required feature(s)" when implementing Required Actions A.2, B.2.and C.2. Therefore, Davis-Besse has added this information to clarify the Bases. Furthermore, as stated in the ISTS Bases, the purpose of Required Actions A.2, B.2, and C. 1 is to ensure that a complete loss of safety function does not occur should a loss of offsite power occur. This means that the"required feature(s)" must have a safety function to provide in the event of a loss of offsite power. Safety functions fall into two categories:
initial condition of an accident and mitigation of an accident.
The Required Actions are, by definition, not consistent with initial conditions of the accident analyses.
The initial conditions of the accident analyses are represented in the LCOs, and the Required Actions only apply when an LCO is not met.Therefore, Required Actions A.2, B.2, and C. 1 are concerned with verifying that the safety function of accident mitigation can be performed.
This is only a concern if the safety function is assumed to be performed during an accident coincident with a loss of offsite power. This is appropriate because without a loss of offsite power coincident with the accident, the existing electrical power configuration would persist and a full train of emergency power would be available.
Lastly, systems which do not perform a function assumed in the safety analysis, such as Power Accident Monitoring and Remote Shutdown, are not considered "required feature(s)" consistent with the existing statements in the Bases. In addition, the safety analyses do not assume an accident is initiated while in a Technical Specification Condition.
That is why the time spent in Technical Specification Conditions is normally limited.Required Actions A.2, B.2, and C. 1 provide additional assurance that the required safety functions could be performed should an accident occur during the limited Completion Times. Therefore, the safety analyses are not affected by this change. Based on this information, Davis-Besse feels that this additional information should be maintained in the Bases for ITS 3.8.1. This editorial change has been approved by the TSTF group.Licensee Response by Bryan Based on the 3/5/2008 phone call, Davis-Besse has decided to Kays on 03/09/2008 remove the changes that were made to incorporate TSTF-402T.
Therefore, the Bases pages for Required Action A.2 (Volume 13, Page 62), Required Action B.2 (Page 64), and Required Actions C. 1 and C.2 have been revised. Additionally, the Justification for Deviations (JFD) 12 (Page 100) has been deleted. A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Robert Clark Draft markup of the Bases for Required Actions A.2, B.2, Cl, and on 03/12/2008 C2 not included in attachment.
Please submit draft markup.[Licensee Response by Bryan The markup is now attached.Kays on 03/14/2008
__NRC Response by Robert Clark No further questions at this time. Item closed.on 03/14/2008 http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 3 of 3 Date Created: 02/04/2008 05:18 PM by Robert Clark Last Modified:
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ITS 3.8 Robert Clark None 137 Information ITS Number: 0SI: DOCNumber:
Bases JFD Number;3.8.3 None M.2 None NRC Author] [Robert Clark Provide justification for the apparent discrepancy between ITS Required Action 3.8.3.D.1 and the Bases for SR 3.8.3.3.ITS Required Action 3.8.3.D.1 which specifies minimum requirements for EDG new fuel oil properties was added to the CTS per DOC M02. The DOC for M02 stated that if the new fuel oil properties are not within the required limits, a period of 30 days is allowed to restore the "stored" fuel oil properties.
This period provides sufficient time to test the "stored" fuel oil to determine that the "new" fuel oil did not cause the "stored" fuel oil to be outside of the required limits. This restoration may involve feed and bleed procedures, filtering, or combinations of these procedures.
Comment However, the Bases for SR 3.8.3.3 states that if "new" fuel oil properties are within acceptable limits (ASTM Standards), fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. In addition, the Bases also states in part that: Failure to meet any of the above limits (ASTM Standards) is cause for rejecting the new fuel oil, but does not represent a failure to meet the LCO concern since the fuel oil is not added to the storage tanks.Please explain why DOC for M02 implies that "new" fuel oil properties not within limits can be added to the storage tanks whereas the Bases for SR 3.8.3.3 states this condition is not allowed.10CFR50.36(c)(3) requires Surveillance Requirements.to verify that the LCO's are met.Issue Date[ 10/24/2007 http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 C.los.Q. e Date 11/14/2007 Logged in User: Jerry Jones'Responses Licensee Response by Bill Bentley on 10/26/2007 The SR 3.8.3.3 Bases (Volume 13, Pages 150 and 151) words in the third paragraph, first sentence of the NRC reviewer's comment is a lead-in to the tests that must be accomplished and pass prior to adding new fuel oil to the storage tanks. The tests are listed in the SR 3.8.3.3 Bases as (a), (b), and (c). These are the same tests referred to in ITS 5.5.12.a, Acceptability of new fuel oil for use prior to addition to storage tanks. If the new fuel oil does not pass these acceptance tests, then the new fuel will be rejected and cannot be added to the onsite EDG fuel oil storage tanks. If the new fuel oil passes these acceptance tests, then it can be added to the onsite fuel oil storage tanks. However, as described in the SR 3.8.3.3 Bases, further additional fuel oil tests are required on the new fuel. The results of these tests are required to be analyzed within 31 days after the initial sample. This means that the new fuel would have already been added to the stored fuel oil tanks, and if the additional "new" fuel oil properties are not met, then ITS 3.8.3 ACTION D (Page 142) must be taken. The Improved Standard Technical Specifications (ISTS) allow the "new" fuel to be added to the onsite storage tanks prior to the completion of the additional fuel oil properties analyses, recognizing that the additional "new" fuel oil properties might be out of limit. The period of 30 days in the Required Action applies when any of the additional fuel oil properties are found to be not within limits on the new fuel oil that was added to the fuel oil storage tanks. Thus, the "new" fuel oil properties are not met, but they affect the"stored" fuel oil since the fuel oil has already been added to the storage tanks. Since Davis-Besse currently does not require this additional "new" fuel oil testing, Discussion of Change M02 provides the justification for the more restrictive change and is explaining how SR 3.8.3.3 and ACTION D works. This is also explained in the Bases for ACTION D. 1 (Page 149). The above description is consistent with the manner in which the ISTS works.NRC Response by Robert Clark on 11/14/2007 ISTB currently have no further questions on this section.Date Created: 10/24/2007 01:31 PM by Robert Clark Last Modified:
11/14/2007 06:01 PM http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 3 JJNASsigjj&#xfd;eReturn to View Menuj 14 rnt Documnent RAI Screening Required:
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3.8 Robert Clark Matthew 1 167 iTS ITS Number: McConnell DOCNumber:
Bases JFD Number: Information 3.8.4 .OR: None None None NRC Autho]r Matthew McConnell Provide an endorsement letter from your Battery Manufacturer(s) to show that float current monitoring can be used to identify a battery's state-of-charge and that the proposed float current value will remain adequate throughout the expected service life of the existing batteries.
Furthermore, provide assurance that the float current value will remain adequate for future batteries (i.e.,.the TS float current value will hold true for the life of the nuclear power plant).Also provide assurance that the equipment used to monitor float current will have the necessary accuracy and capability to measure electrical currents in the expected range. This is consistent with the industry resolution that was reached following the July 12, 2006, Technical Specifications Task Force (TSTF)-360 public meeting.Commnenit 10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.http://www.excelservices.com/exceldbs/itstrack_davisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 3 F Issue 11/16/2007 Close Date 04/10/2008 Logged in User: Jerry Jones'Responses Licensee Response by Jerry Jones on 12/20/2007 The 2 amp limit was previously discussed with the NRC as part of obtaining approval for its inclusion in the CTS. CTS Table 4.8-1 Note (c) (Volume 13, Page 211) was added to the Davis-Besse CTS as part of License Amendment 158. This change allowed charging current to replace the gravity requirement in the table.The NRC Safety Evaluation for this amendment, dated July 16, 1991, specifically states the following: "The Licensee requested information from the station battery manufacturer on the validity of using less than two amps battery charging current as indication of sufficient charge on the station battery as discussed in the IEEE Standard.
In February 1991, the manufacturer conducted testing to determine the state of charge on a battery if, following a battery service test or performance discharge test, a battery is recharged until a battery charging current of less than two amps is reached.The manufacturer stated that at this point, a battery, such as the station battery, would be approximately 95 percent fully charged.""The current station battery has a 20-year design life and is certified by the manufacturer for a nuclear service life of 16 years.The station battery was sized to include a 25-percent margin to account for aging. The licensee has reviewed the station battery loading calculations and has determined that, even including the 95 percent factor discussed above, the station battery will still be able to satisfy its design load requirements at the end of service life." "The NRC staff has reviewed this issue and finds that use of a stabilized battery charging current of less than 2 amps in lieu of specific gravities to indicate that the station battery is charged is allowed by IEEE 450-1980, is in accordance with the NRC guidance of July 16, 1981, and will assure that the station battery will be able to meet its design load requirements.
Therefore, the staff finds the proposed amendment to be acceptable." Thus, the NRC has already asked and Davis-Besse provided the answers to the first two questions.
As far as the third question concerning ensuring the values are sufficient for any replacement batteries, the change control process will ensure that new batteries meet the current requirements related to charging current. If they do not, then a Technical Specification change would be required prior to battery change-out.
Furthermore, since Amendment 158 was approved, Davis-Besse has replaced the batteries with batteries that are equivalent to the previous batteries and continues to meet the 2 amp charging current requirements in the CTS. In addition, Davis-Besse has been using less than two amps battery charging current as indication of sufficient charge of the station battery since.NRC approval of License Amendment 158. The desire to further quantify the current measuring accuracy was previously http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/lfddcea1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 3 of 3 identified during the development of the ITS submittal.
During the upcoming refueling outage which starts on December 30 of this year, a controlled current will be supplied to the circuit and the accuracy of the current reading will be quantified.
This will provide data for Davis-Besse to ensure the equipment used to monitor float current will have the necessary accuracy and capability to measure electrical currents in the expected range. For information, License Amendment
.158, including the NRC Safety Evaluation, is attached.NRC Response by Matthew During a 1/29/08 conference call, the staff informed the licensee McConnell on 02/21/2008 that the referenced LAR allowed the use of float current monitoring to determine a battery's state-of-charge following a discharge and not for everyday use (i.e., the revised TS did not eliminate routinely taking specific gravity measurements).
The licensee agreed to follow-up on this item as a restult of the call.Licensee Response by Bryan An endorsement letter from the battery manufacturer is attached.Kays on 03/16/2008 The equipment used to monitor float current will have the necessary accuracy and capability to measure electrical currents in the expected range.NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on 04/10/2008 Branch (EEEB)staffs request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 11:46 AM by Matthew McConnell Last Modified:
04/10/2008 07:52 AM http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON.
D. C. 20555 EXT-C Lol)1-03537 90. 35QI RECEIVED July 16, 1991 Docket No. 50-346 JUL 3 0 W9 TOLEDO EDISON Mr. Donald C. Shelton, Vice President Nuclear -Davis-Besse Centerior Service Company c/o Toledo Edison Company 300 Madison Avenue Toledo, Ohio 43652
 
==Dear Mr. Shelton:==
 
==SUBJECT:==
AMENDMENT NO.158 TO FACILITY (TAC NO. 79672)The Commission has issued Amendment No.NPF-3 for the Davis-Besse Nuclear Power revises the Technical Specifications in March 1, 1991.OPERATING LICENSE NO. NPF-3 158 to Facility Operating License No.Station, Unit No. 1. The amendment response to your application dated This amendment allows an alternate method of determining battery operability following service or performance discharge surveillance testing.A copy of the Safety Evaluation is also enclosed.included in the Commission's next biweekly Federal Notice of issuance will be Register notice.Sincerely, I Adon B. Hopkins, Sr. Project Manager Project Directorate 111-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
 
==Enclosures:==
: 1. Amendment No. 158 to License No. NPF-3 2. Safety Evaluation cc: See next page UNITED STATES 0( REGULATORY COMMISSION 9, WASHINGTON, D. C. 20555 C,, TOLEDO EDISON COMPANY CENTERIOR SERVICE COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 158 License No. NPF-3 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by the Toledo Edison Company, Centerior Service Company, and the Cleveland Electric Illuminating Company (the licensees) dated March 1, 1991 complies with the standards and requirements of the Atomic Energy Act of 1954,* as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance-(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
: 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read'as follows:  (a) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 158 , are hereby incorporated in the license.The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications..
: 3. This license amendment is effective as of its date of issuance and shall be implemented not later than 45 days after issuance.FOR THE NUCLEAR REGULATORY COMMISSION Hopkins, Sr. Project Manager Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
 
==Attachment:==
 
Changes to the Technical Specifications Date of issuance:
July 16, 1991 UNITED STATES NUCLEAR REGULATORY COMMISSION."
D.C. 20555 SAFETY EVALUATION.BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO-AMENDMENT NO. 158 TO FACILITY OPERATING LICENSE NO. NPF-3 TOLEDO EDISON COMPANY CENTERIOR SERVICE COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT.O. I DOCKET NO. 50-346
 
==1.0 INTRODUCTION==
 
By letter dated.March 1, 1991, Toledo Edison Company-,(thelicensee) proposed an amendment to the Technical Specifications (TS) for the Davis-Besse Nuclear Power Station, Unit No. 1. The proposed change involves TS Table 4.8-1, "Battery Surveillance Requirements," and its bases., The amendment would allow an alternate method of determining battery operability following service or performance discharge surveillance testing.2.0 EVALUATION The' amendment would add a footnote to TS Table 4.8-1 which allows the use of a battery charging current of less than-2 amps when on a float charge in lieu of ,specific gravities as a means of determining station battery operability following battery service or performance discharge"surveillance testing. The proposed change is in accord with guidance provided in a letter from the NRC (T. Novak) to Toledo Edison (R. Crouse) dated July 16, 1981.IEEE Standard 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement-of Large Lead Storage Batteries for Generating'Stations and Substations," allows for the use of a stabilized battery charging current as indication that a battery is charged in lieu of specific gravities.
The licesnee requested information from the station battery manufacturer on the validity, of using less than two amps battery charging current as indication of sufficient charge on the station battery as discussed in the IEEE Standard.In February 1991, the manufacturer conducted testing to determine the state of charge on a battery if, following a battery service or performance ,discharge test, a battery is recharged until a battery charging current of  less than two amps is reached. The manufacturer stated that at this point, a battery, such as the station battery, would be approximately 95 percent fully charged.The current station battery has a 20-year design life and is certified by the manufacturer for a nuclear service life of 16 years. The'station battery was sized to include a 25-percent margin to account for aging. The licensee has reviewed the station battery loading calcu~lations arid has determined that, even including the 95 percent factor discussed above, the station battery will still be able to satisfy its design load requirements at the end of service life.The NRC staff has reviewed this issue and finds-that use of a stabilized battery charging current of less that 2 amps in'lieu of specific gravities to indicate that the station battery is charged is allowed by IEEE 450-1980, is in accord with the NRC guidance of July.16,.
1981, and will assure that the station battery will be able to meet its design load requirements.
Therefore, the staff finds the proposed amendment to be acceptable.
 
==3.0 STATE CONSULTATION==
 
In accordance with the Commission's regulations, the Ohio State official.was notified of the proposed issuance of the amendment.
The State official had no comments.4.0 -ENVIRONMENTAL CONSIDERATION This amendment changes a requirement with respect to installation or useof a facility component located within the restricted area as defined in 10 CFR Part 20 or changes a surveillance requirement.
The staff has deternlinedthat the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative, occupational radiation exposure.
The Commission has prev'iously issued a proposed finding.that the amendment involves no significant hazards consideration and there has been no public comment on such finding (56 FR 13671). Accordingly, the amendment meets thc eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment nieed be prepared in connection with the issuance of the amendment.
 
==5.0 CONCLUSION==
 
The staff has concluded, based on the considerations discussed above, that: (I1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2)'such activities will be conducted in compliance with the Commission's&sect;'regulations, a'nd (3) the issuarnce of this amendment will not be inimical to the common defense and security or to the health and safety of the-public.
Principal Contributor:
Jon B. Hopkins, PDIII-3 Date: July 16, 1991 ATTACHMENT TO LICENSE AMENDMENT NO. 158 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
Remove Insert 3/4 8-10 3/4 8-10 B 3/4 8-1 B 3/4 8-1 TABLE 4.8-1.BATTERY SURVEILLANCE REQUIREIENTS CATEGORY A(1) CATEGORY B (3)Parameter Limits for each Limits for each Allowable-designated pilot connected cell value for each cell connected cell Electrolyte
>Minimum level '" >Minimum level Above top of Level indication mark, indication mark, plates, and _ &" above and < V" above and not maximum level maximum level overflowing indication mark indication mark Float Voltage > 2.13 volts > 2.13 volts(b) > 2.07 vol~ts Not more than.020 below the average of all (c) > 1.195 connected cells Specific)
> 1.200 -Gravitya Average of all Average of all connected cells connected cells> 1.205 > 1.19 5 (c)(a) Corrected for electrolyte temperature and level..(b) .,Corrected for average electrolyte temperature.(c) Or battery charging current, following a service or performance discharge test; is less than two amps, when on'a float charge.(1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements.
are taken and found to be within their allowable values,-and provided all parameter(s) are restored to within limits within the next 6 days.(2) For any Category B parameter(s) outside the limit(s) shown, the battery* may be considered OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits, within 7 days.(3) Any Category B parameter not within its allowable value indicates an inoperable battery..I I DAVIS-BESSE., UNIT I 3/4 8-10 Amendment No. 100,158 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: 2. Verifying total battery terminal voltage is greater than or equal to 129 volts on float charge.b. At least once per 92 days and within 7 days after a battery discharge (battery terminal voltage below 110 Volts),. or:battery overcharge (battery terminal voltage above 150 volts), by: 1. Verifying that the parameters in Table 4.8-1 meet the Category B limits, 2. Verifying that there is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10-6 ohms, and 3. Verifying that the average electrolyte temperature of every sixth connected cell is above 60 0 F.c. At least once per 18 months by verifying that: 1. The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, 2. The cell-to-cell and terminal connections are clean, tight and coated with anti-corrosion material, 3. The resistance
'f. each cell-to-cell and terminal connection is less than or equal to 150 x 10-6 Ohms, and 4. The battery charger will supply at least 475 amperes at a minimum of 130 volts for at least 8 hours.d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the-actual or simulated emergency loads for the design duty cycle when the battery is subjected to. a battery service test.e. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60 month interval this performance discharge test may be performed in lieu of the battery service test.f. Every 18 months, during shutdown, performance discharge tests of battery capacity shall be given to' any battery that shows signs of degradation or has reached 85% of the service life expected for the:application.
Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.DAVIS-BESSE, UNIT 1 3/4 8-9 Amendment No. 100 3/4.8 ELECTRICAL POWER SYSTEMS BASES The OPERABILITY of the A.C. and D.C. power sources and associated distribu-tion systems during operation ensures that sufficient power will be available to supply the safety related equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident conditions within the facility.
The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix "A" to 10 CFR 50.The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation.
The OPERABILITY of the power sources are consistent with the initial condltion assumptions of the safety analyses and are based upon maintaining at least one of each of the onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite A.C. source.The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown-and refueling ensures that 1) the facility can be maintained in the shutdown or refueling condition for extended time periods and 2) sufficient instrumentation and control icapability is available for monitoring and maintaining the facility status.iThe Surveillance Requirements for demonstrating the OPERABILITY of the Istation batteries are based on the recommendations of Regulatory Guide 1.129, "Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants", February 1978, and IEEE Std. 45.0-.80, "IEEE Recommended Practice for. Maintenance, Testing, and Replacement of large Lead Storage Batteries for Generating Stations and Substations".
Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connec-tion resistance values and the performance of battery service and discharge tests. ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity.Table 4.8-1 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and .015 below the manufacturer's full charge specific gravity or a battery charger current of less than two amps is characteristic of a charged cell with adequate capacity.
The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than .020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than .010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery. Exceptions to the specific gravity requirements are taken to allow for the normal deviations experienced after a battery discharge and subsequent recharge associated with a service or performance discharge test. The specific gravity deviations are recognized and discussed in IEEE 450-1980.DAVIS-BESSE, UNIT I B 3/4 8-1 Amendment No. 100, 158 3/4.8 ELECTRICAL POWER SYSTEMS BASES ,Operation with a battery cell's parameter outside the normal limit but within the allowable-value specified in Table 4.8-1 is permitted for up to seven days. During this seven-day period: (1) the allowable value for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than .020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than .040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.DAVIS-BESSE, UNIT 1B3/ 8-AenmtNo 10 B 3/4 8-2 Amendment No. 100 NUCLEAR LOGISTICS INC March 3, 2008, Float Currenf Monitoring
 
==Subject:==
Davis-Besse Nuclear Power Station, GNB NCN Product
 
==Dear Mr. Johnson,==
Please see the attached GNB letter from Robert J. Schmitt addressing Float Current Monitoring in response to your, inquires.If any further information is needed please let me know.Sincerely, Devin Shrode Project Engineer Nuclear Logistics Inc.7450 Whitehall Street
* Fort Worth, Texas 76118
* 817.284.0077
&deg; Fax 817.590.0484
* 800.448.4124
* generalinfo@nuclearlogistics.com www.nuclearlogistics.com GNB.INDUSTRIAL POWER A Division of -EXIDE Technologies 3 March 2007 GNB Industrial Power 3950 Sussex Avenue Float Current Monitoring Aurora, IL 60504-7932 USA From: Robert J. Schmitt 630.862.2200 tel 800.872.0471 tol free Staff Engineer 630.862.2325 fax GNB Network Power www gnb.com To: Mr. Archie Bell NLI
 
==Subject:==
Davis-Besse Nuclear Power Station, GNB NCN Product GNB's position on the use of float current measurements to determine the state of charge of flooded stationary lead-calcium batteries is as follows:* The concept of utilizing float current levels of a flooded, stationary string battery to determine a state of charge throughout the life of the battery is reasonable." There is a relationship between percentage of ampere-hours returned following a successful discharge capacity test and battery state of charge.* Proper follow-up and verification of satisfactory float charge voltage, current and specific gravities is necessary to determine whether the battery is operating properly per GNB's Installation and Operating Manual, section 93.10.* The charge current of each battery and can be affected by impurity levels, age, operating environment and maintenance history.I hope this addresses your concerns on this matter and that you will contact me with any further questions.
Best regards, RjS NRC ITS Tracking.Page I of 4 Return to View Menu Pr int Document RAI Screening Required:
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..4 01: None None None NRC Matthew McConnell Does the proposed two-ampere float current limit indicate a fully charged battery? If not, provide a regulatory commitment to maintain a design margin to ensure that two amperes is indicative of a fully charged battery. This is consistent with the industry resolution that was reached following the July 12, 2006, TSTF-360 public meeting.10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 Comment CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue Date 11/16/2007 Close Date ;04/25/2008 Logged in User: Jerry Jones-'Responses http://www.excelservices.com/exceldbs/itstrackda.visbesse.nsf/1 fddceal Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 4 Licensee Response by Jerry Jones on 12/20/2007 As described in the Davis-Besse response to 200711161146, the 2 amp limit indicates a battery is approximately 95 percent fully charged. Furthermore, the NRC Safety Evaluation for the License Amendment states: "The current station battery has a 20-year design life and is certified by the manufacturer for a nuclear service life of 16 years. The station battery was sized to include a 25-percent margin to account for aging. The licensee has reviewed the station battery loading calculations and has determined that, even including the 95 percent factor discussed above, the station battery will still be able to satisfy its design load requirements at the end of service life." In addition, UFSAR Section 8.3.2.1.2 (Page 2684 of 4076) also describes the 1.25 age factor (i.e., design margin) used in the battery sizing analysis.
Therefore, since this information is provided in the above described NRC Safety Evaluation (a copy of which is provided in the Davis-Besse response to question 200711161146), as well as UFSAR Section 8.3.2.1.2, both of which are regulatory controlled documents, no additional changes are needed.NRC Response by Matthew During a 1/29/08 conference call, the staff informed the licensee McConnell on 02/21/2008 that the referenced LAR allowed the use of float current monitoring to determine a battery's state-of-charge following a discharge and not for everyday use (i.e., the revised TS did not eliminate routinely taking specific gravity measurements).
The staff also questioned the basis for the float current limit value, particularly, the design margins being credited.
The licensee agreed to follow-up on this item as a restult of the call.Licensee Response by Bryan This additional information is being supplied at the request of the Kays on 02/21/2008 NRC. Calculation C EE-002.01-010 states: "The results indicate thata 21 plate, 1500 amp-hour battery, at a temperature of 60 F, degraded to 80% of 95% manufacturers rated capacity, has sufficient capacity to supply the equipment necessary for mitigating the effects of a LOCA followed by a loss of offsite power with essential bus lockout." Calculation C-EE-002.01-010 also states: "A battery aging correction factor of 1.32 shall be used to accommodate for a battery aged to 80% capacity and charged to 95% capacity (see Licensing Correspondence serials 2481 and 2492, and Log 5203)." [A 1.25 aging factor, plus additional conservatism to compensate for a battery charged to only 95% of capacity yields the value of 1.32. (1.25 x 100%/95%=1.32).]
Since the design margin requirement is already in the calculation together with reference to the previous correspondence with the NRC, no further action is required to ensure the requirement is maintained.
Licensee Response by Bryan Kays on 03/16/2008 In addition to the Response on 2/21/2008, Davis-Besse commits to maintain a 5% recharge margin to ensure that the 2 amp float current limit is conservative for the life of the battery. The DC Calculation (C EE-002.01-101, referenced in the previous response) utilizes a temperature correction factor set at 1.11 to'correspond to a minimum temperature of 60 degrees F. The aging http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/I fddcea1 Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 3 of 4 factor is set with a base of 125%, as recommended by IEEE-485, and then modified for Station Batteries only being charged to 95%capacity.Licensee Response by Bill Bentley on 04/14/2008 After a phone call with the reviewer on 4/11/08, Davis-Besse was requested to provide information related to how our batteries are sized relative to the IEEE 485 sizing calculation requirements.
The following is provided as a response:
Calculation C-EE-002.01-010, DC CALC, utilizes IEEE 485-1983.
IEEE 485 Section 6.2 recommends considering temperature, design margin, compensation for aging, and initial capacity when sizing a battery.DC Calc software follows the recommended IEEE 485 format and allows for input of Temperature Correction Factor, Design Margin, and Aging Factor. Calculation C-EE-002.01-010 Section 1.3 discusses the additional considerations included in the evaluation performed by the DC Calc software.
Temperature Correction Factor is set at 1.11 to correspond to a minimum temperature of 60 Degrees Fahrenheit; Aging Factor is set with a base of 125% as recommended by IEEE 485 and then modified to account for Station Batteries only being charged to 95%. Design Margin is set to unity, which implies no additional design margin is considered.
NRC Response by Matthew Does the 2-Amp float current value equate to a 95% charged McConnell on 04/16/2008 battery on the battery manufacturer's recharge curve? If not, describe how the 2-Amp value was calculated.
NRC Response by Matthew In your last response for 200711161146, you stated that Davis-McConnell on 04/22/2008 Besse would supply a controlled current to the circuit and the accuracy of the (float) current reading will be quantified during the next refueling outage. Describe how this task will be formally captured (e.g., Regulatory Commitment).
Licensee Response by Jerry Response to NRC question dated 4/16/2008 Yes, the 2 amp float Jones on 04/25/2008 current value equates to a 95% charged battery based upon testing performed by the battery manufacturer.
The NRC Safety Evaluation for Amendment 158 (copy provided.in the 12/20/2007 response to question 200711161146) states: "In February 1991, the manufacturer conducted testing to determine the state of charge on a battery if, following a battery service or performance discharge test, a battery is recharged until a battery charging current of less than two amps is reached. The manufacturer stated that at this point, a battery, such as the station battery, would be approximately 95 percent fully charged." Response to NRC question dated 4/22/2008 The test described in the NRC reviewer's comment has been completed and Davis-Besse has determined that the equipment that will be used to monitor float current has the necessary accuracy and capability to measure electrical currents in the expected range.NRC Response by Matthew McConnell on 04/25/2008 Based on the licensee's April 25, 2008, response, the staff understands that the 2-amp float current limit equates to a 95%charged battery on the battery manufacturer's recharge curve. The licensee noted that the battery manufacturer performed testing that demonstrated that the 2-amp float current equates to a 95%http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1 fddcea1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 4 of 4 charged battery. The staff finds that the licensee adequately responded to the Electrical Engineering Branch (EEEB)staffs request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 11:47 AM by Matthew McConnell Last Modified:
04/25/2008 04:09 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2 sir Retu'rn to View, MeinuiLPitDcmn RAT Screening Required:
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ITS 3.8 Robert Clark Matthew None 223 ITS ITS Number: McConnell
!OC Number:. BasesJFD.
Nunmber:.Information 3.8.4 OSI: None None None NRC Authdor Matthew McConnell Identify the value of the minimum established design limit for electrolyte level (e.g., low level line on cell jar) in the Technical Specification (TS) Bases. This is consistent with the industry resolution that was reached following the July 12, 2006,.TSTF-360 public meeting. None -page 223 -Volume 13 10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 Comment CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for_._ operation will be met.IssueDate 11/16/2007 Close Date [02/01/2008 Logged in User: Jerry Jones Responses http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/
I fddcea 1 Od3bdbb5 85256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 Licensee Response by Jerry Jones on 12/10/2007 CTS Table 4.8-1 (Volume 13, Page 211) identifies the Category A and B minimum established design limit for electrolyte level as greater than the minimum level indication mark. While the ISTS Bases for SR 3.8.6.3 (Page 233) does not include this specific value (it essentially allows the value to be in a plant-controlled document), Davis-Besse will agree to include this value in the ITS Bases, similar to including the minimum established design limit for cell temperature in the ITS Bases (Page 234). In addition, during the development of the Davis-Besse response to this question, it was noted that the CTS Markup had an administrative error, in that the insert describing the deletion of the Category A and B values and replacement with the phrase "greater than or equal to the minimum established design limits" is incorrectly annotated as replacing the third column value. It should be replacing the first and second column values. A draft markup regarding these changes is attached.
These changes will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on 02/01/2008 Branch (EEEB)staff s request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 11:49 AM by Matthew McConnell Last Modified:
02/01/2008 11:02 AM http://www.excelservices.cOm/exceldbs/itstrack davisbesse.nsf/1 fddceal Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking PageJ of 2 F&#xfd; Iv,&#xfd;Return to View Menu Print DQcu -en RAI Screening Required:
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ITS 3.8 Robert Clark Matthew None 162 ITS Number: McConnell D OC N.umber: Bases JFD Number: Information 3.8.4 OSI: L.1 None None NRC Author Matthew McConnell Provide the basis for the proposed 7-day battery charger Completion Time.Consistent with the industry resolution that was reached following the July 12, 2006, TSTF-360 public meeting, the staff requires that a risk-informed evaluation be performed in accordance with Regulatory Guide 1.174 and 1.177 to support extending the battery charger Completion Time beyond 72 hours when using a non-Class 1E battery charger that is not capable of being supplied power from a source independent of the offsite power system (e.g., diesel generator).
Furthermore, describe the 'alternate means' that is being credited for the proposed extended Completion Time.10 CFR 50.36, "Technical Specifications," requires that the technical CGo!mment specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.I Issue Date I 11/16/2007 http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/l fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 Close Date 02/01/2008 Logged in User: Jerry Jones'Responses Licensee Response by Jerry Davis-Besse has submitted the completion Time for ITS 3.8.4 Jones on 12/20/2007 Required Action A.3 (Volume 13 Page 167), consistent with the currently approved NRC NUREG-1430, Revision 3.1.Furthermore, NRC Administrative Letter 96-04, "Efficient Adoption of Improved Standard Technical Specifications," specifically states that beyond scope issues, which are generally characterized as changes that differ from both the existing technical specifications and the improved STS (thus, the NRC described change is a beyond scope issue), tend to unnecessarily complicate and delay the conversion review process. In addition, during discussions with the NRC concerning the Davis-Besse ITS submittal, held prior to the date of the Davis-Besse ITS submittal, the NRC reiterated this fact and cautioned Davis-Besse about the effect beyond scope changes could have on the ITS approval date.Thus, Davis-Besse did not include these "agreements," which had not been submitted to the NRC in TSTF form yet, in our ITS submittal.
Subsequent to the Davis-Besse ITS submittal, the industry submitted TSTF-500, which proposed implementation of the agreements between the NRC and the industry.
Furthermore, Davis-Besse is not aware of when, or if, the TSTF will be approved by the NRC. However, Davis-Besse has reviewed the NRC meeting minutes for the July 12, 2006 meeting and proposed TSTF and will adopt the more restrictive Completion Time (72 hours versus 7 days), provided this new beyond scope change does not delay NRC approval of the Davis-Besse ITS. A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on 02/01/2008 Branch (EEEB)staffs request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 11:50 AM by Matthew McConnell Last Modified:
02/01/2008 11:04 AM http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/lfddcea1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2 R eturn to View Menu Print Do eI RAI Screening Required:
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ITS 3.8 Robert Clark Matthew 9 176 TITS Number: McConnell DOC.Number:
Bas.es..JFD Nuniber: Information 3.8.4 .OS: None None None NRC Matthew McConnell Clarify what is meant by the ability to use an inoperable yet functional replacement charger (LCO 3.8.4.2 TS Bases).10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established Comment for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue Date 11/16/2007 Close Date [02/01/2008 Logged in User: Jerry Jones wResponses I ....Licensee Response by Jerry Jones on 12/06/2007 The Bases for ISTS 3.8.4 Required Actions A. 1, A.2, and A.3 (Volume 13, Page 176) states a balance of plant non-Class 1E http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 battery charger may be used as an alternate means for restoring battery terminal voltage (as required by Required Action A. 1).Therefore, if a non-class 1 E battery charger is allowed to be used, then a class 1E battery charge that is technically inoperable, but still functional (i.e., it can still provide adequate current and voltage to maintain the associated battery terminal voltage greater than or equal to the minimum established float voltage), should be allowed to be used. For example, a class lE battery, charger could be inoperable due to a mounting problem (i.e., it is found to not meet the class 1E requirements), but it is still perfectly capable of performing its normal function.
Therefore, the Bases was changed to allow the use of inoperable, but functional Class 1E battery charger.NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on 02/01/2008
[Branch (EEEB)staffs request for additional information.
_[Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 12:03 PM by Matthew McConnell Last Modified:
02/01/2008 11:07 AM http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/1 fddcea 10d3bdbb585256e8...
6/2/2008 NRC ITS Tracking Pagel of3~ss~gn~Return to View Menu Print Docuen RA-I Screening Required:
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Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC ITS TRACKING NRC Reviewer ID 200711161204 Conference Call Requested?
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3.8 Robert Clark Matthew None 178 ITS Number: McConnell DOC Nunb.er.:.
Bases..,IJFD Number:.Information 3.8.4 OS: None None None NRC A orl Matthew McConnell Describe why the word 'design' is being deleted in the TS Bases for SR 3.8.4.2.10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established' for each structure, system, or component that is part of the primary success Conmm1ent path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCQs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue Date] 11/16/2007 Close D 04/22/2008 Logged in User: Jerry Jones'Responses Licensee Response by Jerry Jones on 12/06/2007 The word "design" is being deleted from ITS Bases SR 3.8.4.2 (Volume 13, Page 178) because the value specified in ITS SR 3.8.4.2 (Page 168) for the battery charger test is less than the http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 3 design value. The design output of the battery chargers is 600 amps. This current licensing basis value (475 amps) in CTS 4.8.2.3.2.c (Page 160) was approved for Davis-Besse in License Amendment 100, dated March 16, 1987. As stated in the ITS Bases for SR 3.8.4.2 (Page 178), the 475 amps and 8 hour test duration ensures that the charger can restore the battery from the minimum design charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences.
Therefore, Davis-Besse believes no changes to the ITS submittal are required.
Note that this question is similar to question 200711161204.
" NRC Response by Matthew McConnell on 02/21/2008 During a 1/29/08 conference call, the staff asked the licensee to confirm whether a current limiting device was used to prevent the battery charger from supplying greater than 475 amps. The licensee agreed to follow-up onthis item as a restult of the call.Licensee Response by Bryan Kays on 02/21/2008 This additional information is being supplied at the request of the NRC. As previously stated, the design output of the battery chargers is 600 amps. The battery chargers have a current limiter to limit the output of the chargers to a maximum of 110 percent of capacity (660 amps). The battery chargers have a maximum short circuit contribution of 750 amps. At 475 amps output, the battery chargers are supplying all steady state DC loads required under any conditions while recharging the batteries to a fully charged condition over a period of eight hours from a discharge condition of 105 volts per battery.Licensee Response by Bryan Kays on 03/16/2008 This response supersedes the response on 2/21/2008.
The battery charger current limit is set at 575 amps. Davis-Besse currently tests the battery charger at 475 amps for 8 hours. This amount of energy (8 X 475 or 3800 amip-hours) exceeds the'energy required to meet USAR 8.3.2.1.3, which states "each charger is capable of supplying all steady-state DC loads required under any condition of operation while charging the battery to a fully charged condition over a period of 12 hours from a discharge of 750 Amperes to the minimum limit of 105 Volts per battery (which correlates to the manufacturer specification of the one-hour discharge rate)." Since the current method of testing exceeds the energy delivery requirements, the design requirement of the charger is being verified..
Licensee Response by Jerry This response is an addition to the response on 3/16/2008.
Davis-Jones on 04/21/2008 Besse has determined that the wording for the Bases Surveillance Requirement (SR) 3.8.4.2 (Volume 13, Page 178) should state the SR verifies the "required design" capacity of the required chargers.A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Matthew The licensee adequately respondedto the Electrical Engineering McConnell on 04/22/2008 Branch (EEEB)staff s request for additional information.
Therefore, EEEB has no further questions at this time.http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/I fddcea 1 Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 3 of 3 Date Created: 11/16/2007 12:04 PM by Matthew McConnell Last Modified:
04/22/2008 01:40 PM http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1fddcea1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking.Pagel of 2 Return to View Menula Print Docmn RAI Screening Required:
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Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC ITS TRACKING NRC Reviewer ID 200711161205 Conference Call Requested?
No CIAeo[ ESI -Emergent Staff Issue ITS Section: TB POC: JFD Number:. Page.Number(s):
ITS 3.8 Robert Clark Matthew 8 168 ITS Number: McConnell DOC Number:. Bases JFD Number: Information 3.8.4 OSI: None None None NRC ][ Matthew McConnell Adding the word 'required' to TS Surveillance Requirement (SR) 3.8.4.2 is not consistent with the ISTS. Provide a detailed justification for adding the word'required' to TS SR 3.8.4.2 and describe how the proposed change would affect performance of this SR on the spare battery charger or alternate means described in the TS Bases.10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 Comment CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue Date 111/16/2007 Close Date 02/01/2008 Logged in User: Jerry Jones.'Responses http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/lfddcea1Od3bdbb585256e8...
-6/2/2008 NRC ITS Tracking Page 2 of 2 Licensee Response by Jerry Jones on 12/05/2007 ITS SR 3.8.4.2 (Volume 13, Page 168) added the word "required" to be consistent with the TSTF-GG-05-01, Writer's Guide for Plant-Specific Improved Technical Specifications.
TSTF-GG 01, Section 4.1.3.b states, in part, "required" is specifically used in Conditions, Required Actions and Surveillances to denote reference to equipment which is "required" by the LCO for the specific existing Applicability.
In a case where the LCO only requires some of all possible components be used to satisfy the LCO requirement, then the clarification of "required" is used in the Conditions, Required Actions, and Surveillances.
At Davis-Besse, there are two normal battery chargers and one spare battery charger available to meet the LCO requirements for each train. As stated in the ITS Bases (Page 174), the LCO requires "each train consisting of two batteries, one battery charger for each battery and the corresponding control equipment and interconnecting cabling to the associated bus within the train to be OPERABLE." Therefore, the LCO only requires some of all of the possible components installed to be OPERABLE (either both normal chargers or one normal and one spare charger can meet the LCO requirements for each train). Based on this fact, the wording of required is appropriate.
This addition of the word "required is also explained and justified in Justification for Deviation (JFD) 8 (Page 169). Based on the use of the word "required," ITS SR 3.8.4.2 would not be required to be met for the spare battery charger unless the spare is being used to satisfy the requirements of the LCO. At that time, the spare charger would become the "required" charger and the ITS SR 3.8.4.2 must be met by either of the two means stated in the Surveillance Requirement.
NRC Response by Matthew McConnell on 02/01/2008 The licensee adequately responded to the Electrical Engineering Branch (EEEB)staff's request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 12:05 PM by Matthew McConnell Last Modified:
02/01/2008 11:09 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 10d3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2 1[&#xfd; Jgn OReturn to View Menu QPint DocumuentI RAI Screening Required:
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TB.-PO.C; JFD.Numb.er.:.
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ITS 3.8 Robert Clark Matthew None 163 ion ITS Number: McConnell DOC Number: BasesJPFD Numni-ber:n Information 3.8.4 OSI: L.3 None None NRC Matthew McConnell The purpose of SR 3.8.4.2 is to test the design capacity of the battery charger.The test amperage is based on the output rating of the charger. Justify replacing
'based on' with 'also well within' in the TS Bases for SR 3.8.4.2.10 CFR 50,36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established C omament for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue Date 11/16/2007 C[lose Date 1 04/22/2008 Logged in User: Jerry Jones'Responses Licensee Response by Jerry The words "based on" are being changed to "also well within" in http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking.Page 2 of 2 Jones on 12/06/2007 the ITS Bases SR 3.8.4.2 (Volume 13, Page 178) because the value specified in ITS SR 3.8.4.2 (Page 168) for the battery charger test is less than the design value (thus it cannot be based on the output rating of the charger).
The design output rating of the battery chargers is 600 amps. This current licensing basis value (475 amps) in CTS 4.8.2.3.2.c (Page 160) was approved for Davis-Besse in License Amendment 100, dated March 16, 1987. As stated in the ITS Bases for SR 3.8.4.2 (Page 178), the 475 amps and 8 hour test duration ensures that the charger can restore the battery from the minimum design charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences.
Therefore, Davis-Besse believes no changes to the ITS submittal are required.
Note that this question is similar to question 200711161204.
NRC Response by Matthew McConnell on 02/21/2008 During a 1/29/08 conference call, the staff asked the licensee to confirm whether a current limiting device was used to prevent the battery charger from supplying greater than 475 amps. The licensee agreed to follow-up on this item as a restult of the call.Licensee Response by Bryan This additional information is being supplied at the request of the Kays on 02/21/2008 NRC. As previously stated, the design output of the battery chargers is 600 amps. The battery chargers have a'current limiter to limit the output of the chargers to a maximum of 110 percent of capacity (660 amps). The battery chargers have a maximum short circuit contribution of 750 amps. At 475 amps output, the battery chargers are supplying all steady state DC loads required under any conditions while recharging the batteries to a fully charged condition over a period of eight hours from a discharge condition_of 105 volts per battery.Licensee Response by Bryan This response supersedes the response on 2/21/2008.
See the Kays on 03/16/2008 response to RAI 200711161204.
NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on' 04/22/2008 Branch (EEEB)staffs request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 12:06 PM by Matthew McConnell Last Modified:
04/22/2008 01:40 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2 SF.Assign .Return to View Menu a RAI Screening Required:
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No Category J ESI -Emergent Staff Issue ITS Section: TB POC: JF.D Number: Page Number(s):
ITS 3.8 Robert Clark Matthew "1 178 ITS.Number:
McConnell DOC. Number.:.
BasesJF.D Number: Information 3.8.4 OSI: None None None NRC Author ][Matthew McConnell Describe the basis for the minimum design float voltage of 129 volts (i.e., manufacturer recommended volts per cell multiplied by the number of cells).This information should be incorporated in the revised TS Bases for SR 3.8.4.1.10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 Comment CFR 50.34. A TS limiting condition for operation (LCO) must be established
............................................
for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue.D!ate 11/16/2007 Close Date [02/06/2008 Logged in User: Jerry Jones'Responses http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1 fddcea 10d3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 Licensee Response by Jerry Jones on 01/30/2008 The basis for the 129V value in the ITS Bases for SR 3:8.4.1 (Volume 13, Page 178) is based upon the value from Amendment 100 to the Davis-Besse CTS. This Amendment revised the Davis-Besse Technical Specifications for station batteries based upon the NRC's model Technical Specifications of July 16, 1981. The.Amendment added Specification 4.8.2.3.2.a.2 to verify the total battery terminal voltage is greater than or equal to 129 volts on float charge. The NRC Safety Evaluation associated with the Amendment states that since the value is consistent with the STS, it is acceptable.
However, this value will be changed to reflect the minimum design voltage of 130.2V to match the value given in the ITS Bases for SR 3.8.6.2 and SR 3.8.6.5 (Page 233). The value in SR 3.8.6.2 and SR 3.8.6.5 ensures optimal long term battery performance is obtained by maintaining a float voltage greater than or equal to the minimum established design limits provided by the battery manufacturer.
A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Matthew McConnell on 02/06/2008 The licensee adequately responded to the Electrical Engineering Branch (EEEB)staffs request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 12:07 PM by Matthew McConnell Last Modified:
02/06/2008 09:29 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2 FK;&#xfd;'-'&#xfd;Return to View Menu a Prnt Do~n RAI'Screening Required:
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ESI -Emergent Staff Issue ITS Section: TB POC: JFD Number: Pa geNumber(s):
3.8 Robert Clark Matthew 2 222 ITS ITS Number: McConnell DOC Nun.ber: Bases.FD....Number:
Information 3.8.6 .OS: None None None NRC Author[] Matthew McConnell Provide further justification for deviating from Improved Standard TS (ISTS)Limiting Condition for Operation (LCO) 3.8.6 Conditions A, B, C, D, and E (i.e., proposed 'more' instead of 'two' and removed 'on one train'). JFD 2 -Pages 222/223 -Volume 13 10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 Com~ment CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issu.e.Dat][
11/16/2007 Close Date 102/01/2008 Logged in User: Jerry Jones'Responses http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 Licensee Response by Jerry Jones on 01/28/2008 The bracketed words "or two" and "on one train" in ISTS 3.8.6 Conditions A, B, C, and D (Volume 13, Pages 222 and 223) have been changed to be consistent with the allowance in ISTS 3.8.6 Condition E (Volume 13, Page 223). Specifically, Condition E allows batteries in redundant trains (i.e., in two trains) to be inoperable for up to 2 hours. Thus, the words "on one train" are not necessary to be included in Conditions A through D. The proposed wording is also consistent with the wording in the ISTS 3.8.9 ACTIONS (Pages 285 and 286). The ISTS 3.8.9 ACTIONS A, B, and C (Page 285) state "one or more" buses or distribution subsystems inoperable.
Thus, the Conditions cover one or more trains -i.e., they do not restrict entry to only one train of distribution.
If multiple trains of distribution are inoperable such that a safety function is affected, this is covered by ISTS 3.8.9 Condition D (Page 286). This identical change was previously approved by the NRC for both the DC Cook Units 1 and 2 and Monticello ITS submittals.
Therefore, for consistency with the manner in which these types of Conditions are described, Davis-Besse believes that the change to the bracketed words is correct.NRC Response by Matthew McConnell on 02/01/2008 The licensee adequately responded to the Electrical Engineering Branch (EEEB)staff's request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 11:55 AM by Matthew McConnell Last Modified:
02/01/2008 11:19 AM T http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2 Fy ;;;J'-"Return to View Menu~ I Pint Do~cuientI RAI Screening Required:
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ITS 3.8 Robert Clark Matthew None 213 ITS Number: McConnell DOC Number: Bases..JFD..
Nunber.: Information 3.8.6 OSI: M.1 None None NRC Author Matthew McConnell It is the staff's understanding that float current monitoring provides an indication of the battery's state of charge. Clarify the intent of the following statement: "The purpose of SR 3.8.6.1 is to assist in the determination of the state of charge of the battery to assure the battery can provide the required current and voltage to meet the design requirements." 10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the Comment safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.IssuejDate
[11/16/2007 Close Date 1 02/01/2008 Logged in User: Jerry Jones http://www.excelservices.com/exceldbs/itstrack davisbesse.nsf/lfddcealOd3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2"Responses Licensee Response by Bryan The NRC reviewer is correct. The float current monitoring does Kays on 12/07/2007 provide an indication of the battery's state of charge. Therefore, the statement in Discussion of Change MO1 (Volume 13, Page 213)will be changed to delete "to assure the battery can provide the required current and voltage to meet the design requirements." A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
Licensee Response by Jerry Based on a recent phone conversation with the NRC reviewer, Jones on 01/30/2008 which clarified the NRC reviewer's concern, the first sentence has been modified to further delete the word "assist." An amended draft markup regarding this and the original change is attached and supersedes the draft markup attached to the first Davis-Besse response.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on 02/01/2008 Branch (EEEB)staff s request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 11:59 AM by Matthew McConnell Last Modified:
02/01/2008 11:22 AM http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/l1fddcealOd3bdbb585256e8...
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ITS 3.8 Robert Clark Matthew None 223 ITS Number: McConnell DOC Nunber:. Bases .JFD. Nuniber.:.
Information 3.8.6 OSI: None None None NRC [Matthew McConnell Describe the battery pilot cell selection process.10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success CoQrimment path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a. fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue Date 11/16/2007 CloseDate 04/10/2008 Logged in User: Jerry Jones'Responses I...Licensee Response by Bryan Kays on 12/08/2007 The Davis-Besse battery cell selection process is performed in accordance with EPRI TR 100248, Revision 1, Stationary Battery Guide. Engineering determines annually which cells are to be the http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 pilot cell using the guidance of the EPRI guide and engineering judgment.
Currently, Davis-Besse has two pilot cells per battery.The applicable pages of the EPRI guide are attached for information.
NRC Response by Matthew McConnell on 02/21/2008 During a 1/29/08 conference call, the staff informed the licensee of the original basis for rotating pilot cells (i.e., based on removing electrolyte from cells during specific gravity measurements).
The staff also informed the licensee that the TSTF-360/500 working group and NRC staff had agreed that the pilot cell(s) should be the lowest voltage cell(s). The licensee agreed to follow-up on this item as a restult of the call.Licensee Response by Bryan Davis-Besse will commit to revise the pilot cell selection criteria.Kays on 03/16/2008 The pilot cells will be selected to represent the lowest voltage cells in the battery. Pilot cell selection will be evaluated at a minimum of every two years (i.e., every 2 years or earlier) to ensure they continue to meet the selection criteria.NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on 04/10/2008 Branch (EEEB)staff s request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 12:01 PM by Matthew McConnell Last Modified:
04/10/2008 07:53 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 10d3bdbb585256e8...
6/2/2008 EPRI Licensed Material Vented Lead-Acid Battery Inspections The general inspections will be referred to here as monthly inspections for consistency with industry standards and manufacturer's recommendations..
This section presents the following information r elated to monthly inspections: " The purpose of each inspection and what the inspection does or does not accomplish
&deg; Battery degradation and failures that are detectable by each inspection" A step-by-step description of how to perform each inspection
* Precautions and limitations to observe while performing each inspection Monthly inspections provide a general assessment of the battery's condition and verify that it is being maintained within normal operating limits.. All inspections should be made with the battery operating under normal float conditions.
Monthly inspections include the following:
* Visual inspection of the batfery, battery rack or cabinet, and battery area* Measurement of float voltage at the battery terminals* Check of charger output current and voltage* Visual check of electrolyte levels Measurement of battery area ambient temperature and check of ventilation equipment condition* Measurement of pilot cell voltage, specific gravity, and electr olyte temperature
* Check for battery grounds The monthly inspection results should be recorded on a data sheet or form to support a'long-term battery trending program, Typical data sheets are provided by the manufacturer., 11.1.1 Pilot Cell Selection Considerations The following monthly checks are normally performed on vented lead-acid battery pilot cell(s): " Cell voltage" Electrolyte specific gravity" Electrolyte temperature Pilot cells are intended to provide a general indication of the battery status with regard to voltage, specific gravity, and temperature Quarterly and annual inspections usually involve some check of all the cells. The actual number of pilot cells is determined by the plant; battery manufacturers sometimes specify only a single pilot cell for each battery installation..
If the battery uses more than one rack, one pilot cell per rack is sometimes 11-2 EPRI Licensed Material -Vented lead-Acid Inspections selected to ensure that the effects of ambient temper ature variations in the area are detected., Pilot cells should be periodically rotated, but generally no more frequently than annually..
Because the electrolyte checks can remove small amounts of electrolyte from the cell, peiiodic rotation ensures that excessive electrolyte is not removed from any single cell., The following factors should be considered in selecting pilot cells:.Is the pilot cell accessible?
Although the cell might be representative of the general state of the battery, personnel safety considerations should preclude its selection if it is in a hard to reach spot" Is the pilot cell near a ventilation duct? The temperature of the cell should be near the aver age battery temperature,* Is the pilot cell one of the lower-voltage cells? Typically, the pilot cell hasone of the: lowest voltages to ensure'that minimum specifications are met.. The lowest-voltage cell is often selected to be the pilot cell.. Periodic rotation of a pilot cell should be.based on annual inspection data..* Has the cell been a pilot cell in the past? Pilotcells should be periodically rotated, and the same cell should not be used again unless there is a compelling reason for its selection:
11.1.2 General Visual Inspection 11.1..2.1 Purpose of Inspection The purpose of the monthly visual inspection is to assess the gener al condition of the battery and battery area. This monthly inspection is intended to check for the more conspicuous problems that can be encountered.
The following types of battery degradation can be detectedduring the'general visual inspection:
o Abnormal battery area ambient temperature
* Impiopei operation of the battery area ventilation system* .Electrolyte leakage or presence of eleztiuolyte on cell covers-Cracked cells or flame arrestors ,A detailed visual inspection of each cell is recommended on an annual basis See Section 11.31 for, annual visual inspections., 11-3 NRC ITS Tracking Page 1 of 2 Assignj Y0` Return to View Menuj Print Documenmt RAI Screening Required:
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NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC Reviewer ID ][ 200711161202 Conference Call Requested?
No Caltegry ESI -Emergent Staff Issue ITS Section: TB POC: JFDNumber:
Page Number(s):, 3.8 Robert Clark Matthew None 223 ITS 1TS-Number:
McConnell DOC Number: Bases JFD Number: Information 3.8.6 OSI: None -None None NRC Author][ Matthew McConnell Describe how the battery pilot cell temperatureis representative of the entire battery's temperature.
10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established Comment for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration,,or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue Pate F[ 11/16/2007
.[Close D~ate 04/02/2008 Logged in User: Jerry Jones'Responses Licensee Response by Jerry CTS 4.8.2.3.2.b.3 (Volume 13, Page 210) requires a verification
]IJones on 12/06/2007-that average temperature of every sixth cell (not all of the cells) is http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 above 60 degrees F every 92 days. ITS SR 3.8.6.4 requires a similar verification every 31 days (Page 224). Davis-Besse adopted the ISTS requirements for this Surveillance, thus, we are not sure exactly what the NRC reviewer is requesting.
The ISTS Bases for this SR (Page 234) states the following: "This Surveillance verifies that the pilot cell temperature is greater than or equal to the minimum established design limit (i.e., 40OF [Note: changed to 60 in the Davis-Besse ITS submittal]).
Pilot cell electrolyte temperature is maintained above this temperature to assure the battery can provide the required current and voltage to meet the design requirements.
Temperatures lower than assumed in battery sizing calculations act to inhibit or reduce battery capacity.
The Frequency is consistent with IEEE-450 (Ref. 1)." Therefore, Davis-Besse believes that the intent of the ISTS Surveillance is to ensure the battery temperature is not so low that it cannotperform its intended function.
Furthermore, since the battery cells for a given battery are all in the same room, whose temperature is controlled, then sampling only the pilot cells should give adequate confirmation that the entire battery is above the minimum limit. While there could be a large cell-temperature deviation due to a shorting condition, IEEE-450 states in Section D.3 that large cell-temperature deviations, which are usually caused by shorting conditions, are also evident by the cell voltage.ISTS SR 3.8.4.1 (Page 168) requires a weekly check of overall cell voltage. This SR has been maintained in the Davis-Besse ITS.Therefore, the ITS includes a Surveillance that is capable of detecting a shorting problem.NRC Response by Matthew During a 1/29/08 conference call, the staff requested the licensee McConnell on 02/21/2008 to provide additional information as to how pilot cell(s)temperature is representative of the entire battery's temperature.
The licensee agreed to follow-up on this item as a restult of the-_call.Licensee Response by Bryan Since the 5 degree F temperature deviation criteria can not always Kays on 03/16/2008 be demonstrated, Davis-Besse will make a commitment to either use cell temperature as one of the criteria for selecting
'the pilot cells or a separate pilot cell will be selected to reflect average_battery temperature.
NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on 04/02/2008 Branch (EEEB)staffs request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 12:02 PM by Matthew McConnell Last Modified:
04/02/2008 04:04 PM http://www.excelservices.com/exceldbs/itstrack, davisbesse.nsf/1 fddceal Od3bdbb585256e8
... 6/2/2008 NRC ITS Tracking Page I of 2 I E1- s-sgn V Return to View Menu W Print Docuen RAI Screening Required:
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NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC Reviewer D[200711161209 Conference Call Requested?
No CAtI FJ ESI- Emergent Staff Issue ITS Section: TB POC: JFD Number: Page Number(s):.
ITS 3.8 Robert Clark Matthew None 230 ITS Number: McConnell DOC..Number:
Bases. JIFD N.umnb.er:
Information 3.8.6 OS!" None 1 None NRC Author [Matthew McConnell Describe the intent of removing the following statement from the TS Bases for Condition B.1 and B.2 of LCO 3.8.6: "but there are one or more battery cells with float voltage less than [2.07] V, the associated "OR" statement in Condition F is applicable and the battery must be declared inoperable immediately.
If float voltage is satisfactory and there are no cells less than [2.071 V" 10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the Comment safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue.Date I1 11/16/2007 Close Date 102/01/2008 Logged in User: Jerry Jones http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/
1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2'Resnionses Licensee Response by Jerry Jones on 12/07/2007 ISTS 3.8.6 ACTION B (Volume 13, Page 222), provides the Required Actions for when a battery has a float current greater than 2 amps. The Required Actions are to perform SR 3.8.4.1, which requires verification that the entire: battery float voltage is.within limits (i.e., it checks the overall battery voltage, not the individual cell voltage), and to restore the float current to within limits: The Bases statement in question is describing a Condition that is not covered by the Required Actions. It is basically a reminder that if an additional condition is discovered while float current is not within limits, to make sure ACTION F is entered.Furthermore, individual cell voltage is not required to be checked while in ACTION B. This type of reminder is not needed in the Bases, and is inconsistent with the format of the ISTS. For example, ISTS 3.8.6 ACTION A (Page 222) provides the Required Actions for when an individual cell voltage is not within limits.When.this Condition occurs concurrent with a float current problem on the same battery, ACTION F is required to be entered.However, the ISTS Bases for Required Actions A. 1, A.2, and A.3 (Page 229) does not have a similar reference as the one being deleted from the ITS Bases for Required Actions B.1 and B.2.Thus, the deletion of the words in'question is consistent with the other related Required Action Bases. Therefore, Davis-Besse does not desire to maintain this inconsistent reminder in the Required Actions B. 1 and B.2 Bases. ITS 3.8.6 ACTION F (Page 224) and associated Bases (Page 232) is clear as to when ACTION F is to be taken. In addition, Davis-Besse does not understand how this Bases deletion is classified as an Emergent Staff Issue. Davis-Besse believes that this is an administrative type issue.NRC Response by Matthew McConnell on 02/01/2008 The licensee adequately, responded to the Electrical Engineering Branch (EEEB)staffs request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 12:09 PM by Matthew McConnell Last Modified:
02/01/2008 11:25 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8....
6/2/2008 NRC ITS Tracking.Page I of 2 Assign 1 Return to View Men]u I dPrint.Docciient RI Screening Required:
Yes This Document will be approved by: George Wilson This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC ITS TRACKING NRC Reviewer FD /200711161210 Conference Call Requested?
No Cat /I ESI -Emergent Staff Issue ITS Section: TB POC: JFD Number: Page-Number(s):
3.8 Robert Clark Matthew None 233 I TS Number: McConnell DOC Number: Bases JFD Number: Information 3.8.6 OSI: None 2 None NRC.Author][
Matthew McConnell The TS Bases for SR 3.8.4.1 identifies the minimum established float voltage provided by.the battery manufacturer to be 129 volts whereas the TS Bases for SR 3.8.6.2 and 3.8.6.5 identifies the value as being 130.2 volts. Justify the discrepancy.
10 CFR 50.36, "Technical Specifications,".
requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. ATS limiting condition for operation (LCO) must be established Comment for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue Dt[11/16/2007 Close Date [~,oII o Cloe a: iel02/06/2008 Logged in User: Jerry Jones'Responses http://www.excelservices.com/exceldbs/itstrack_.davisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 Licensee Response by Jerry The 130.2V value provided in ITS SR 3.8.6.2 and SR 3.8.6.5 is the Jones on 01/30/2008 correct value. The voltage limit for ITS SR 3.8.4.1 will be changed, as described in the response for question 200711161207.
NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on 02/06/2008 Branch (EEEB)staffs request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 12:10 PM by Matthew McConnell Last Modified:
02/06/2008 09:32 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2.1 411 A~ssign~ Return to View Menu~ I Print Documen RAI Screening Required:
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NRC ITS TRACKING NRC Reviewer IDD 200711161212 Conference Call Requested?
No Category ESI -Emergent Staff Issue ITS SIection:;
TB POC: JFD.Number:
Page Nulmbero(s):
ITS 3.8 Robert Clark Matthew None 234 ITS Number: McConnell D.OCN.unmber:.
Bases JFD Nu.nmber: Information 3.8.6 0SI: None 4 None NRC.Author Matthew McConnell The following proposed TS Bases wording for SR 3.8.6.6 is not consistent with the ISTS or IEEE Std. 450: "Initial conditions for the modified performance discharge test should be identical to those specified for a performance discharge
[emphasis added]test.""Since the ampere-hours removed by a one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the modified [emphasis added] performance discharge
[emphasis added] test without compromising the results of the performance discharge test." a) Does the modified performance test envelope the service test?b) The staff does not consider these changes to be editorial.
Justify the Comment deviation from the ISTS and IEEE Std. 450.10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for http://www~excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 1 operation will be met.Issue Date 11/16/2007 Close Date 02/06/2008 Logged in User: Jerry Jones"'Responses Licensee Response by Jerry Question b response:
Davis-Besse originally believed the changes Jones on 12/20/2007 in question made to the ISTS provided clarity of the requirements.
However, Davis-Besse has re-evaluated the changes in question.
1)The first sentence in question:
The words (Volume 13, Page 234)will be changed back to "service test." The reference to IEEE-450 will be maintained, since this is where the information concerning the initial conditions is specified., 2) Second sentence in question: The Word "modified" (Page 234) will be deleted (the change adding* the word"'"discharge" will be maintained since the test is a performance discharge test, not a performance test). Question a response:
As stated in the ISTS Bases for SR 3.8.6.6, fourth paragraph (Page 234), and maintained in the Davis-Besse ITS Bases, the modified performance discharge test consists of two rates, the one minute rate for the battery or the largest current -load of the duty cycle, followed by the test rate employed for the performance test. "both of which envelop the duty cycle of the service test." Furthermore, IEEE-450, Section 5.4 states that it is permissible to perform a modified performance test "if the test's discharge rate envelops the duty cycle of the service test." In addition, Davis-Besse does not understand how corrections needed to the Bases are classified as an Emergent Staff Issue. Davis-Besse believes that this is an administrative type issue. A draft markup regarding these changes is attached.
These changes will be reflected in the supplement to this section of the ITS Conversion ,Amendment.
NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on 02/06/2008 Branch (EEEB)staff s request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 12:12 PM by Matthew McConnell Last Modified:
02/06/2008 09:43 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/
1 fddcea 10d3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2 IFFRetur~n to View..MenuJJQ Print Document RAI Screening Required:
Yes This Document will be approved by: Carl Schulten This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC Reviewer ID 200805211335 Conference Call Requested?
No Catego [ In Scope ITS Section: TB POC: JFD Number: Page Number(s):, ITS 3.8 Robert Clark Matt McConnell None 223 Information ITS.Number:
OS!: D.OC Number: Bases JFD Number: 3.8.6 None None None The licensee provided the following Regulatory Commitment in response to staff RAI 200711161201: "Davis-Besse will commit to revise the pilot cell selection criteria.
The pilot cells will be selected to represent the lowest voltage cells in the battery. Pilot cell selection will be evaluated at a minimum of every two years (i.e., every 2 years or earlier) to ensure they continue to meet the selection criteria." The NRC staff position requires that licensees perform quarterly (i.e., every 92 days) assessments to ensure that the pilot cell continues.to be the lowest voltage.cell. This frequency is consistent with proposed new TS SR 3.8.6.5 and IEEE Std. 450 guidance for measuring each battery cell voltage.Please consider this position or provide adequate justification for performing Comment assessments every 2 years..................
10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs)' are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue Date 05/21/2008 http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e..-
7/17/2008 NRC ITS Tracking Page 2_of 2 Closei ! ate 06/20/2008 Logged in User: Anonymous"'Responses Licensee Response by Jerry Davis-Besse made the previous commitment to be consistent with Jones on 06/16/2008 both the EPRI guideline referenced in RAI 200711161201 and the proposed TSTF-500, Section 4.5.2, which states that pilot cell selection should be evaluated at a minimum of each outage to ensure they continue to meet the selection criteria.
Davis-Besse also reviewed the NRC meeting minutes for the July 12, 2006 meeting, which is what proposed TSTF-500 is based upon, and the meeting minutes did not include any specific requirements concerning battery pilot cell selection time. This TSTF is currently under review by the NRC. Davis-Besse does not believe that any further change to the time period should be made at this time.However, Davis-Besse will further commit to changing the battery pilot cell selection evaluation time period to be consistent with the requirements of TSTF-500 within 6 months after the NRC has approved the TSTF as documented in the Federal Register (Davis-Besse assumes that the TSTF will be approved and documented in the Federal Register as part of the Consolidated Line Item Improvement Program.)NRC Response by Gerald Waig The licensee adequately responded to the Electrical Engineering on 06/20/2008 Branch (EEEB) staffs request for additional information.
Therefore, EEEB has no further questions at this time. (Posted by Aron Lewin for Matthew McConnell as requested)
Date Created: 05/21/2008 01:35 PM by Gerald Waig Last Modified:
06/20/2008 08:25 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsfl1 fddceal Od3bdbb585256e...
7/17/2008 NRC ITS Tracking Page I of 2 1~j~gn ~Return to View MenulI Print Documen RAI Screening Required:
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NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC Reviewer ID 1200711161157 Conference Call Requested?
No Category ESI -Emergent Staff Issue ITS Section: TBPOC.: JFD Number: Page Number(s).:.
ITS 3.8 Robert Clark Matthew None 246 ITS Number: McConnell DOC.Nuniber:.
Bases JFD Number: Information 387 OSI: M.1 None None NRC Autr[Matthew McConnell Provide a detailed justification for the proposed 8-hour Completion Time for TS LCO 3.8.7 Condition B.10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses, and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established Comment for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.IssujeDate
[11/16/2007 Close DateI 02/01/2008 Logged in User: Jerry Jones'Responses I Licensee Response by Jerry This question is essentially a duplicate of 200710261400.
See Jones on 11/30/2007 response to RAI 200710261400.
II II II http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceaIOd3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 NRC Response by Matthew McConnell on 02/01/2008 The licensee adequately responded to the Electrical Engineering Branch (EEEB)staff s request for additional information.
Therefore, EEEB has no further questions at this time..--1 Date Created: 11/16/2007 11:57 AM by Matthew McConnell Last Modified:
02/01/2008 11:28 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2 'Ass' New Resiponse
'tReturn to View Menu Print Document RAI Screening Required:
No This is a Non RAI Dialogue This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
Status: Approval Not Required This document will not be relied upon by staff for disposition of the LAR Yes NRC ITS TRACKING NRC Reviewer ID 200801171153 Conference Call Requested?
No CFategor In Scope ITS Section; TB POC.: JFD Number: Page .Number(s):
ITS 3.8 Robert Clark None 252 Information ITS Number: OSI: DOC.Number:
Bases JFD Number: 3.8.7 None None 1 NRCAuthor
[Robert Clark The Bases for LCO 3.8.7 ties the definition for inverter operability to the 120 VAC vital buses. There is no need to define inverter operability in terms of the 120 VAC vital buses. Operability of the 120 VAC vital buses are discussed the Bases for LCO 3.8.9 (Distribution Systems -Operating) i.e., the 120 VAC vital buses are OPERABLE if they are energized to their proper voltage from the Comment associated inverter or Class 1E constant voltage transformer.
The definition for OPERABLE inverters should state: OPERABLE inverters require the inverter output voltage and frequency be within tolerance, and power input to the inverter be supplied from a 125 VDC station battery.Alternatively, power supply may be from an AC source or battery charger via rectifier as long as the station battery is available as the uninterruptible power supply.Issue Date 01/17/2008 Close Date 102/12/2008 Logged in User: Jerry Jones.'Responses Licensee Response by Jerry Jones on 02/11/2008 ISTS 3.8.7 LCO Bases (Volume 13, Page 252) correctly ties the definition for inverter OPERABILITY to the 120 VAC vital buses.As stated in the Applicable Safety Analyses section of the ISTS 3.8.7 Bases (Page 250), second paragraph, "The OPERABILITY of the inverters is consistent with the initial assumptions of the http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 accident analyses and is based on meeting the design basis of the unit. This includes maintaining required AC vital buses OPERABLE during accident conditions..." Thus, the inverters must be powering the 120 VAC vital buses for the inverters to be considered OPERABLE.
If the inverters were not tied to the 120 VAC vital buses, and a design basis accident coincident with a loss of offsite power occurred, then the 120 VAC vital buses would be de-energized until the EDGs started and ties to the buses. The accident analysis assumes the 120 VAC vital buses remain energized at all times subsequent to the loss of offsite power. The only way this can occur is for the 120 VAC vital buses to be energized from the inverters.
ISTS LCO 3.8.7 (Page 245) ensures this requirement is maintained.
If an inverter is not powering its associated 120 VAC vital bus, then the inverter is inoperable and Condition A (Page 245) is entered. 24 hours is provided to restore the inverter to OPERABLE status (i.e., place it back on the 120 VAC vital bus). When in this condition, the 120 VAC vital bus is still required to be OPERABLE (i.e., energized) by ISTS LCO 3.8.9 (Page 285). The 120 VAC vital bus is considered OPERABLE as long as it is energized from a qualified source.Furthermore, the Note to ISTS 3.8.7 ACTION A (Page 245)specifically states to enter applicable Conditions and Required Actions of LCO 3.8.9 if any 120 VAC vital bus is de-energized.
Davis-Besse believes that the wording in the ISTS Bases is correct and should be maintained.
If the wording was changed to that proposed in the NRC question, then if the inverter is fully energized but was not powering the 120 VAC vital bus, the inverter would be considered OPERABLE.
Thus, as long as the 120 VAC vital bus was energized by a qualified source described in the ISTS 3.8.9 Bases, all LCO requirements would be met and no ACTIONS would be required.
However, as stated above, if a design basis accident concurrent with a loss of offsite power occurred, the safety analyses assumptions would not be met (i.e., the 120 VAC vital buses must be energized at all times following the loss of power, not just after the EDGs restore power).NRC Response by Robert Clark No further questions at this time. Item closed.onR ReposebyRoer/Car0frte Date Created: 01/17/2008 11:53 AM by Robert Clark Last Modified:
02/12/2008 03:22 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2~Assig6Ne~w Respns~eF4 Return to View Menuli ZPrint Documient RAI Screening Required:
No This is a Non RAI Dialogue This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING Status: Approval Not Required This document will not be relied upon by staff for disposition of the LAR Yes NRC Reviewer ID 200801171201 Conference Call Requested?
No Category In Scope ITS Section: TBYPOC: JFD Number: Page Number(s):
ITS 3.8 Robert Clark None 269 Information ITS Number.: OSI:. D!O.C. NunmMber:
Ba ses.. JFD NuMber-: 3.8.8 None None 1 NRC Author [Robert Clark The Bases for LCO 3.8.8 ties the definition for inverter operability to the 120 VAC vital buses. There is no need to define inverter operability in terms of the 120 VAC vital buses. Operability of the 120 VAC vital buses are discussed the Bases for LCO 3.8.10 (Distribution Systems -Shutdown) i.e., the 120 VAC vital buses are OPERABLE if they are energized to their proper voltage either from a)the associated inverter, via inverted 125 VDC voltage or the Class 1E Comment constant voltage transformer, or b) the associated non-essential power source-(regulated instrumation distribution panel YAR or YBR).The definition for OPERABLE inverters should state: OPERABLE inverters require the inverter output voltage and frequency be within tolerance, and power input to the inverter be supplied from a 125 VDC station battery.Alternatively, power supply may be from an AC source or battery charger via rectifier as long as the station battery is available as the uninterruptible power supply.Issue Date 01/17/2008 Close Date[ 02/12/2008 Logged in User: Jerry Jones'Responses Licensee Response by Jerry The response to 200801171153 applies, with the Applicable Jones on 02/11/2008 sections of ISTS 3.8.8 applied in lieu of ISTS 3.8.7 and ISTS 3.9.10 applied in lieu of ISTS 3.8.9.http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 NRC Response by Robert Clark oN further questions at this time. Item closed.on 02/12/2008 Date Created: 01/17/2008 12:01 PM by Robert Clark Last Modified:
02/12/2008 03:24 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2[1 s&#xfd; jgn ,&#xfd;Return to View Menu1Q Print Document RAI Screening Required:
Yes This Document will be approved by: Tim Kobetz This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC Reviewer ID 200710261400 Conference Call Requested?
No Catego&#xfd;J In Scope ITS Section: TB POC: JFD Number: Page Number(s):
ITS 3.8 Robert Clark 3 287 Information ITS Number: OS1: DOC Number: Bases JFD Number;: 3.8.9 None. None None NRC Author] Robert Clark F Provide justification for the 8 hour Completion Time (CT) for LCO 3.8.9 Condition B, One or more AC vital buses inoperable.
JFD 3 states that the 8 CT is provided for all AC buses -4.16 kV, 480V and the 120VAC vital buses and that the consequences of a loss of a train of the 120 VAC vital buses would be similar to the loss of a train of the above higher voltage buses. A loss of one 120 VAC vital bus would place the unit in a half trip conditions.
Loss of more than one 120 VAC vital bus would lead to a reactor trip and possibly trigged an SFAS signal. However, loss of one 4.16 Kv CommeilneAt bus and .its associated 480V bus would not necessarily lead to a half reactor trip because the station batteries should provide power to the vital buses via the inverters for at least 2 hours. It appears that a loss of one or more 120VAC vital bus should have a CT of 2 hours consistent with LCO 3.8.9. Condition C, One DC electrical power distribution subsystems inoperable.
If one 120VAC circuit is inoperable, a CT of 2 hours should be sufficient time to place the affected RPS and SFAS channels in the bypass mode.10CFR50.36(c)(3) requires Surveillance Requirements to verify that the LCO's are met.IssuieDate 10/26/2007 Close Date[ 01/14/2008 Logged in User: Jerry Jones' Responses http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 Licensee Response by Jerry Jones on 11/30/2007 CTS 3.8.2.1 Action (Volume 13, Page 278) requires that with less than the required electrical buses OPERABLE, to restore the inoperable bus to OPERABLE status within 8 hours. ITS 3.8.9 Condition B (Page 285) directly reflects this CTS requirement, by requiring the AC vital buses to be restored to OPERABLE status within 8 hours (i.e., the 8 hours is consistent with current licensing basis). As stated in Justification for Deviation (JFD) 3 (Page 287), this is acceptable because the consequences of a loss of a train of the 120 VAC vital bus would be similar to the loss of a train of the higher voltage (4.16kV and/or 480V) electrical power distribution system. The Safety Features Actuation System (SFAS) and the Reactor Protection System (RPS) have two channels per train. If one channel in Train 1 or Train 2 is inoperable (i.e., one channel without 120 VAC vital bus power), then Condition B would be entered. Loss of one channel reduces the system logic from two-out-of-four to one-out-of-three, but the safety function is still maintained and the logic will still withstand a single failure.Furthermore, a trip in any of the remaining channels would cause the system to actuate and perform its safety function.
If one 120 VAC vital bus in Train 1 and one 120 VAC vital bus in Train 2 were inoperable (i.e., both buses concurrently inoperable), then Condition E (Page 286) would be entered since the safety function would not be met. Condition E was added by Discussion of Change (DOC) M01 (Page 280) expressly to cover a degradation of the electrical distribution system that would cause a required safety function to be lost. Condition E gives no additional time for continued operation and requires that LCO 3.0.3 be entered immediately.
Based on the above, 8 hours is the appropriate Completion Time for Condition B of ITS 3.8.9. Therefore, Davis-Besse believes that no change is necessary and that an 8 hour Completion Time when one or both AC vital buses in a train are inoperable is acceptable.
NRC Response by Robert Clark on 01/14/2008 ISTB currently have no further questions on this section.Date Created: 10/26/2007 02:00 PM by Robert Clark Last Modified:
01/14/2008 11:21 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 Section 3.9 RAIs NRC ITS Tracking Page I of 3 F Retur~n to View Mejnu [Q rintociet RAI Screening Required:
Yes This Document will be approved by: Tim Kobetz This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC ITS TRACKING NRC Reviewer I D 1200711271441 Conference Call Requested?
No CAo Other Technical Challenge ITS. Seetion:, 'rB POC: JFD Number: Page Number(s);
ITS 3.9 Bill Cartwright Bill Cartwright 1 28 Information ITS.Numfbe;.r:
OR: D , " DOC N.umber:;
Bases JFD Number: 3.9.2 None, None None To meet operability requirements, source range instruments must have the required sensitivity.
Provide the analysis or assessment to show that Davis Besse is not susceptable to neutronic decoupling during a core offload, with fuel removed from the core around a credited source range detector, and that the detectors maintain the adequate sensitivity.
In the proposed LCO 3.9.2, the bracket text was added to the ISTS: Two source range neutron flux monitiors
[,one from each side to the reactor Commen. core,] shall be operable., Many PWRs have addressed a potential situation called "neutronic decoupling", a condition during a core offload where the distance between a source range detector and active fuel causes the detector to loose sensitivity to the point where it can no longer meet its design function.The language in the ISTS compensates for this by the general, non-location specific requirement for operability, while the proposed ITS language implies location requirements on operability that may not be valid.Issue Date 11/27/2007 Close Date 02/25/2008 Logged in User: Anonymous'Responses F. _______________________
Licensee Response by Bryan Kays on 11/28/2007 The ITS wording in question for Tech Spec 3.9.2. is consistent with the current license basis approved by the NRC in License Amendment Request (LAR) 172 and LAR 269. Since the ITS is http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal0d3bdbb585256e...
5/30/2008 NRC ITS Tracking Page 2 of 3 consistent with the approved license basis, Davis-Besse believes the wording to be acceptable.
Copies of LAR 172 and LAR 269 are attached for information.
NRC Response by Bill Please see the attached inspection report from Diablo Canyon, p 19 Cartwright on 11/29/2007 of the pdf file. The reference is ADAMS ML060260012 (publically available).
As a clarification, my concern is that by defining operability of the detectors in your proposed improved technical specifications (ITS) by orientation (180 apart) a core offload may result in a condition where only one of the two detectors can be operable due to loss of sensitivity (neutronic decoupling).
The Improved Standard Technical Specifications accomidate this through a general operability requirement that is not location specific.Licensee Response by Bill In the original license submittal for LAR 172, it was statdd that for Bentley on 12/03/2007 CTS 3/4.9.2: "In order to ensure that the neutron flux in the core is appropriately monitored during refueling operations, the LCO has been modified to require that the two operable neutron monitors be from seperate channels (and, therefore, from opposite sides of the core)." For example, there is an RPS instrument and a Gamma Metrics instrument on each side of the core. Each set of instruments is powered from the same 120 VAC source (Y1 on one side, and Y2 on the other side). By stipulating that the instruments be from opposite sides of the core, a single power supply failure (single channel failure) will not remove all indication.
Making the suggested change would alter the basis of the previous approval provided by the NRC.NRC Response by Bill Discussions with the Technical Branch indicate that there are not Cartwright on 12/14/2007 regulatory requiremtnts for independent electrical supplies for the refueling source range detectors.
This is reinforced by the allowance for temporary source range detectors discussed in the basis. The amendment
#172 quoted also incorporated post accident monitoring requirements which do have independent electrical requirements.
The statements that discuss changing count rates under the channel check Surveilance Requirement in the basis can not modify the operability requirements for these detectors.
If specific locations are specified for refueling source range detectors, they must be capable of performing their design basis function(s), ie to detect an inadvertant dilution.
The original_question is still open.Licensee Response by Jerry Jones on 02/23/2008 In response to 12/14/2007 comment from the NRC, CTS 3.9.2 (Volume 14, Page 23) has been revised to delete the statement"one from each side of the reactor core." To support this change, a new Discussion of Change (DOC) L02 (Page 26) has been written, ITS LCO 3.9.2 (Page 28) has been changed to delete the statement"one from each side of the reactor core," Justification for Deviation (JFD) 1 has been deleted, the appropriate Bases (Pages 32 and 33) changes have been made, and Bases JFD 2 has been deleted. Furthermore, the Applicable Safety Analyses Bases (Page 32) have been changed to be consistent with similar wording in the http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsfl1 fddceal Od3bdbb585256e...
5/30/2008 NRC ITS Tracking Page 3 of 3 ITS 3.3.9 Bases. Specifically, as stated in the proposed DOC, the source range neutron flux monitors are not assumed in any safety analyses, thus they cannot meet Criterion
: 3. The specific reason for maintaining the monitors in ITS 3.9.2 has been provided.NRC Response by Bill The response is acceptable, this question is closed.Cartwright on 02/25/2008 1 Date Created: 11/27/2007 02:41 PM by Bill Cartwright Last Modified:
02/25/2008 09:16 AM http.://www.excelservices.co.m/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e...
5/30/2008 NUCLEAR UNITED STATES REGULATORY COMMISSION WASHINGTON, D.C. 20555 RECEIVED AUG 31ir2 TOLEDO EDISON August 24, 1992 Docket No. 50-346 Mr. Donald C. Shelton Vice President, Nuclear -Davis-Besse Centerior Service Company c/o Toledo Edison Company 300 Madison Avenue Toledo, Ohio 43652 EXT-9 2-0.28 72
 
==Dear Mr. Shelton:==
 
==SUBJECT:==
AMENDMENT NO.172 TO FACILITY (TAC NO. M75236)The Commission has issued Amendment No.NPF-3 for the Davis-Besse Nuclear Power revises the Technical Specifications in February 2, 1990.OPERATING LICENSE NO. NPF-3 172 to Facility Operating License No.Station, Unit No. 1. The amendment response to your application dated This amendment revises TS 3/4.3.3.6, Post-Accident Monitoring Instrumentation, by adding nPttron flux (wide range) and neutron flux (source range) instru-mentation to Tables 3.3-10 and 4.3-10 to reflect the appropriate surveillance requirements for the new monitors.
Additionally this amendment revises TS 3/4.9.2, Refueling Operations
-.Instrumentation, by adding a requirement to calibrate the neutron flux monitors prior to entry into Mode 6.A copy of the Safety Evaluation is also enclosed.included in the Commission's next biweekly Federal Sincerely, Notice of issuance will be Register notice.o~n 8Hopkins, Sr. Project Manager Project Directorate 111-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
 
==Enclosures:==
 
I.- Amendment No. 172 to License No. NPF-3 2. Safety Evaluationn cc w/enclosures:
See next page Mr. Donald C. Shelton Toledo Edison Company Davis-Besse Nuclear Power Station Unit No. I cc: Mary E. O'Reilly Centerior Energy Corporation 300 Madison Avenue Toledo, Ohio 43652 Mr. Robert W. Schrauder Manager, Nuclear Licensing Toledo Edison Company 300 Madison Avenue Toledo, Ohio 43652 Gerald Charnoff, Esq.Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.W.Washington, D.C. 20037 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division 1700 Rockville Pike, Suite 525 Rockville, MD 20852 Resident Inspector U. S. Nuclear Regulatory Commission 5503 N. State Route 2 Oak Harbor, Ohio 43449 Mr. Murray R. Edelman Executive Vice President
-Power Generation Centerior Service Company 6200 Oak Tree Boulevard Independence, Ohio 44101 Radiological Health Program Ohio Department of Health Post Office Box 118 Columbus, Ohio 43266-0149 Attorney General Department of Attorney General 30 East Broad Street Columbus, Ohio 43215 Mr. James W. Harris, Director Division of Power Generation Ohio Department of Industrial P. 0. Box 825 Columbus, Ohio 43216 Regulations 0 Ohio Environmental Protection Agency DERR--Compliance Unit ATTN: Zack A. Clayton P. 0. Box 1049 Columbus, Ohio 43266-0149 President, Board of Ottawa County Commissioners Port Clinton, Ohio. 43452 State of Ohio Public Utilities Commission 180 East Broad Street Columbus, Ohio .43266-0573 Mr. James R. Williams State Liaison to the NRC Adjutant General's Department Office of Emergency Management Agency 2825 West Granville Road Columbus, Ohio 43235-2712 o4_ UNITED STATES 9 &deg;NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20565 gC , TOLEDO EDISON COMPANY CENTERIOR SERVICE COMPANY., AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR-POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 172 License No. NPF-3 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by the Toledo*Edison Company, Centerior Service Company, and the Cleveland Electric Illuminating Company (the licensees) dated February 2, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954. as amended (the* Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health ard safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical-to thecommon defense and security or to the health and safety of the public; and E. -The issuance of this amendment is in accordance with.10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
: 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment;'
and paragraph 2.C.(2) of Facil~ity Operating License No. NPF-3 is hereby amended to read as follows: S 2 (a) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.172 , are hereby incorporated in the license.The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.
: 3. This license amendment is effective as of its date of issuance and shall be implemented not later than 90 days after issuance.FOR THE NUCLEAR REGULATORY COMMISSION on B. Hopkins, Sr. Project Manager Project Directorate 111-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
 
==Attachment:==
 
Changes to the Technical Specifications Date of issuance:August 24, 1992 0 ATTACHMENT TO LICENSE AMENDMENT NO. 172 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
Remove Insert 3/4 3-48 3/4 3-48 3/4 3-48a 3/4 3-50 3/4 3/50 3/4 9-2 3/4 9-2 0 0 0 TABLE 3.3-10 POST-ACCI DENT MONITORING INSTRUMENTATION co C=CL C+-'W-4 INSTRUMENT
: 1. SG Outlet Steam Pressure 2. RC Loop Outlet Temperature
: 3. RC Loop Pressure 4. Pressurizer Level 5. SG Startup Range Level 6. Containment Vessel Post-Accident Radiation 7. High Pressure Injection Flow 8. Low Pressure Injection (DHR) Flow 9. Auxiliary Feedwater Flow Rate 10. RC System Subcooling Margin Monitor ll.. PORV Position Indicator 12. PORV Block Valve Position Indicator 13. Pressurizer Safety Valve Position Indicator 14. BWST Level 15. Containment Normal Sump Level 16. Containment Wide Range Water Level MINIMUM CHANNELS OPERABLE I/Steam Generator 2/Loop 2/Loop 2 2/Steam Generator 2 I/Channel.
1/Channel 2/Steam Generator-1 1 1 1/Valve 3 1 1 0 LA (D C+.'ba'-4.I'i TABLE 4.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUI REMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
: 1. SG Outlet Steam Pressure M R 2. RC Loop Outlet Temperature M R 3. RC Loop Pressure M R 4. Pressurizer Level M R 5. SG Startup Range Level M R 6. Containment Vessel Post-Accident Radiation M R 7. High Pressure Injection Flow M R 8. Low Pressure injection (DHR) Flow M R 9. Auxiliary Feedwater Flow Rate M R 10. RC System Subcooling Margin Monitor M R 11. PORV Position Indicator M R 12. PORV Block Valve Position Indicator M R 13. Pressurizer Safety Valve Position Indicator M R 14. BWST Level S R 15. Containment Normal Sump Level M R 16. Containment Wide Range Water Level M R 3/4.9 REFUELING OPERATIONS
*BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met: a. Either a K f of 0.95 or less, which includes a 1% Ak/k conservati allowance for uncertainties, or b. A boron concentration, of > 1800 ppm, which includes a 50 ppm conservative allowance for uncertainties.
APPLICABILITY:
M1ODE 6.ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at > 10 gpm of 8750 ppm boric acid solution or its equivalent until Keff is reduced to < 0.95 or the boron concentration is restored to > 1800 ppm, whichever is the more restrictive.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to: a. Removing or unbolting the, reactor vessel head, and b. Withdrawal of any safety or regulating rod in excess of 3 feet from its fully inserted position within the reactor pressure vessel.4.9.1.2 The boron concentration of the reactor pressure vessel and the refueling canal shall be determined by chemical analysis at least once each 72 hours.0 DAYIS-BESSE, UNIT 1 3/4 9-1 Amendment No. 143 UNITED STATES*NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20556 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 172 TO FACILITY OPERATING LICENSE NO. NPF-3 TOLEDO EDISON COMPANY CENTERIOR SERVICE COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346
 
==1.0 INTRODUCTION==
 
By letter dated February 2, 1990, the Toledo Edison Company (the licensee)requested changes to Technical Specification (TS) 3/4.3.6, Post Accident Monitoring Instrumentation, which would add neutror flux (wide range) and neutron flix (source range) instrume:,tation to Tables 3.3-10 a:nd 4.3-10 reflect the appropriate surveillance requirements for the new monitors.
In addition this proposed amendment would revise TS 3/4.9.2, Refueling Operations
-Instru-mentation, by adding a requirement to calibrate the neutron flux monitors prior to entry into Mode 6.The licensee has been using the excore neutron flux monitors supplied with the Babcock & Wilcox (B&W) nuclear instrumentation system to meet both TS 3./4.9.2 and 3/4.3.1.1, Reactor Protection System Instrumentation, Table 3.3-1 and 4.3-1, Item 11 (Source Range, Neutron Flux and Rate -Startup and Shutdown).
However, the B&W-supplied nuclear instrumentation system is not environmentally qualified.
Two excore neutron flux monitors (each with wide range and source range capa-bility) qualified to meet the requirements of Regulatory Guide 1.97, "Instru-mentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Acc-ident," were installed during the fifth refueling outage.2.0 EVALUATION The nuclear instrumentation (NI) system is designed to provide neutron flux information over the full range of reactor operations.
To provide total monitoring, three ranges of neutron flux detectors are furnished:
source range, intermediate range and power range. The NI .system consilsts of two source range channels, two intermediate range channels and four power range channels.
'This arrangement allows continuous monitoring of neutron flux level from source range  low to 125% of rated power. A minimum of one decade overlap between ranges is provided.
The power range detectors are required by the reactor protection system (RPS) to perform safety functions, and are part of the RPS.During refueling (Mode 6), the operability of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect reactivity changes in the core.The purpose of the post-accident monitoring system (PAMS) is to follow the course of an accident condition with wide range instrumentation which provides the operators the essential safety status information needed to return the plant to a maintained, safe shutdown condition.
The operability of the post-accident monitoring system ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.Supplement I to NUREG-0737, "Clarification of TMI Action Plant Requirements," directs the installation of various instrumentation systems as specified in Regulatory Guide 1.97, and specifies that neutron flux measurements must be made to indicate whether plant safety functions are being accomplished and to provide information required to mitigate the consequences of an accident.
Regulatory Guide 1.97 further specifies that the neutron flux measurements must be made with components/devices that meet certain criteria among which are that the equipment is environmentally qualified per Regulatory Guide 1.89, "Qualification of Class 1E Equipment for Nuclear Power Plants," and the methodology described in NUREG-0588, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment." Although the B&W-supplied NI system components do not meet the Regulatory Guide 1.97 criteria, Supplement 1 to NUREG-0737, paragraph 6.1.b, permits plants to rely on currently installed equipment even if it is presently not environmentally qualified.
The equipment is required to eventually be replaced with environ-mentally qualified components.
The new excore neutron flux monitoring system was installed to comply with Supplement 1 to NUREG-0737 and it meets the requirements that are specified in Regulatory Guide 1.97. The components are: (a) Environmentally qualified per IEEE 323-1974, "Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations," (b) Seismically qualified per IEEE 344-1975, "Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations," (c) Independently separated per-IEEE 384-1977, "Standard Criteria for Independence of Class 1E Equipment and Circuits," and (d) System cables splices, and connections are qualified per IEEE 383-1974,"Standard for Type Test of Class IE Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations" 0  The new Gamma-Metrics supplied excore neutron flux monitoring system provides neutron flux measurement from reactor shutdown to reactor full power level. The new system consists of two independent channels each with a source range and wide range display. Prior to the modification which installed the new excore neutron flux monitoring system in addition to the original B&W-supplied nuclear instru-mentation system, the NI system also provided in the control room and containment during refueling operations an audible indication of the source range counts. The two new excore monitors will be used to provide the audible indication of source range counts in the control room and containment during refueling operations.
The two new excore monitors will also be available to be used in additionto the orig-inal NI system to provide the visual indication of source range counts in the control room during refueling operations.
The new monitors should be included in the technical specifications and appropriate surveillance testing be reflected to demonstrate operability of the monitors.
Specifically, two new line items "36.Neutron Flux (Wide Range)" and "37. Neutron Flux (Source Range)" are proposed to be added to Tables 3.3-10 and 4.3-10. This will require a minimum of one channel of each to be operable in Mode I (Power Operation) through Mode 3 (Hot Standby), that a channel check-be performed monthly and a channel calibration be performed at each refueling.
These requirements are similar to those of other PAMS instru-ments. The neutron detectors must be excluded from the channel calibration due to their non-adjustability (channel gain is adjustable).
Since, during refueling operations, the new monitors will provide the audible indication of source range counts in the control room and containment and will be available in addition to the B&W-supplied NI system to provide the visual indica-tion of source range counts in the control room, appropriate changes are required for TS 3/4.9.2. In order to ensure that the neutron flux in the core is appro-priately monitored during refueling operations, the LCO has been modified to require that the two operable neutron monitors be from separate channels (and, therefore, from opposite sides of the core). A channel calibration (TS 4.9.2d) of the monitors will be required to be performed prior to entry into Mode 6 (Refueling), if not performed within the last 18 months. The addition of this requirement is necessary because a channel calibration requirement for the new monitors does not exist, while the original NI system is calibrated by TS 3/4.3.3.1 requirements on a refueling basis. As the proposed Surveillance Requirement will apply to both the original and new flux monitors, the phrases "if not performed within the last 18 months" must be included to provide necessary flexibility in scheduling the channel calibration of the original monitors.
These monitors are typically calibrated during the latter stages of a refueling outage and not prior to Mode 6 entry. It is intended that TS 4.0.2. (1.25 criterion) would also be applicable to the channel calibration frequency of 18 months. It should be noted that the original NI system will continue to be utilized to meet its previous TS 3/4.3.1.1 requirements.
The proposed changes would not increase the probability of equipment degradation because there would be no decrease in TS operability and surveillance require-ments. The B&W-supplied NI will continue all of its present functions and the new excore neutron flux monitoring system will provide an independent and qualified system with additional TS operability and surveillance requirements.
The proposed changes do not inhibit the function of existing Class 1E equipment and there is no S  radiological consequence associated with the increased TS requirements.
The proposed changes will ensure that the new system is adequately tested at the appropriate frequency.
The staff has reviewed the proposed changes to TS 3/4.3.6 and 3/4.9.2 and finds that the addition of a qualified excore neutron flux, monitoring system is acceptable.
 
==3.0 STATE CONSULTATION==
 
In accordance with the Commission's regulations, the Ohio State official was notified of the proposed issuance of the amendment.
The State official had no comments.4.0 ENVIRONMENTAL CONSIDERATION This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes a surveillance requirement.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individu'al or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (56 FR 43813). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
 
==5.0 CONCLUSION==
 
The staff has concluded, based on the considerations discussed above, that:' (1)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the commondefense and security or to the health and safety of the public.Principal Contributor:
James J. Lombardo Date: August 24, 1992 9P.0 0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Lqj 47-," September 12, 2005 Mr. Mark B. Bezilla Vice President-Nuclear, Davis-Besse FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station 5501 North State Route 2 Oak Harbor, OH 43449-9760 D E f E iW SEP 19 2005 DBNPS
 
==SUBJECT:==
DAVIS-BESSE NUCLEAR POWER STATION, UNIT 1 -ISSUANCE OF AMENDMENT RE: REFUELING OPERATIONS
-INSTRUMENTATION (TAC NO. MC5473)0
 
==Dear Mr. Bezilla:==
The Commission has issued the enclosed Amendment No. 269 to Facility Operating License No. NPF-3 for the Davis-Besse Nuclear Power Station, Unit 1. The amendment revises the Technical Specifications (TSs) in response to your application dated December 20, 2004, as supplemented by letter dated April 6, 2005.This amendment revises TS 3/4.9.2, "Refueling Operations
-Instrumentation." Specifically, the changes revise TS 3/4.9.2 concerning source range flux monitors to be more consistent with improved Standard Technical Specifications.
A copy of the Safety Evaluation is also enclosed.
The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.Sincerely, William A. Macon, Jr., Project Manager, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-346
 
==Enclosures:==
: 1. Amendment No. 269 to NPF-3 2. Safety Evaluation cc w/encls: See next page Davis-Besse Nuclear Power Station, Unit 1 0 cc: Mary E. O'Reilly FirstEnergy Corporation 76 South Main St..Akron, OH 44308 Manager -Regulatory Affairs FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station 5501 North State -Route 2 Oak Harbor, OH 43449-9760 Director, Ohio Department of Commerce Division of Industrial Compliance Bureau of Operations
& Maintenance 6606 Tussing Road P.O. Box 4009 Reynoldsburg, OH 43068-9009 Regional Administrator U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60523-4351.
Michael A. Schoppman Frarnatome ANP 24 Calabash Court Rockville, MD 20850 Resident Inspector U.S. Nuclear Regulatory Commission 5503 North State Route 2.Oak Harbor, OH 43449-9760 Barry Allen, Plant Manager FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station 5501 North State -Route 2 Oak Harbor, OH 43449-9760 Dennis Clurn Radiological Assistance Section Supervisor Bureau of Radiation Protection Ohio Department of Health P.O. Box 118 Columbus, OH 43266-0118 Carol O'Claire, Chief, Radiological Branch Ohio, Emergency Management Agency 2855 West Dublin Granville Road Columbus, OH 43235-2206 Zack A. Clayton DERR Ohio Environmental Protection Agency P.O. Box 1049 Columbus, OH 43266-0149 State of Ohio Public Utilities Commission 180 East Broad Street Columbus, OH 43266-0573 Attorney General Office of Attorney General 30 East Broad Street Columbus, OH 43216 President, Board of County Commissioners of Ottawa County Port Clinton, OH 43252 President, Board of County Commissioners of Lucas County One Government Center, Suite 800 Toledo, OH 43604-6506 The Honorable Dennis J. Kucinich United States House of Representatives Washington, D.C. 20515 The Honorable Dennis J. Kucinich United States House of Representatives 14400 Detroit Avenue Lakewood, OH 44107 Mr. Lew W. Myers Chief Operating Officer FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station 5501 North State Route 2 Oak Harbor, OH 43449-9760 UNITED STATES NUCLEAR REGULATORY COMMISSION 0WASHINGTON, D.C. 20555-0001 o* p FIRSTENERGY NUCLEAR OPERATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 269 License No. NPF-3 The U.S. Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by the FirstEnergy Nuclear Operating Company (the licensee) dated December 20, 2004, as supplemented by letter dated April 6, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii)that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
: 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended'to read as follows:  '(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 269 , are hereby incorporated in the license.FirstEnergy Nuclear Operating Company shall operate the facility in accordance with the Technical Specifications.
: 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of the date of issuance.FOR THE NUCLEAR REGULATORY COMMISSION 4e-411 &#xfd;- .-"L X Gene Y. Suh, Chief, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation
 
==Attachment:==
 
Changes to the Technical Specifications O Date of Issuance:
September 12, 2005 0 ATTACHMENT TO LICENSE AMENDMENT NO. 269 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change.Remove Insert 3/4 9-2 3/4 9-2 REFUELING OPERATIONS INSTRUMENTATION LIMITJNG CONDInON.FOR OPERATION 3.9.2 Two source range neutron flux monitors, one from each side of the reactor core, shall be OPERABLE.APPLICABILITY:
MODE 6.ACTION: a. With only one of the required OPERABLE source range neutron flux monitors, 1. Immediately suspend CORE ALTERATIONS, and 2. Immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than the RCS boron concentration requirement of LCO 3.9.1.b. With no OPERABLE source range neutron flux monitor, I. Perform ACTION a., and 2. Immediately initiate action to restore one source range neutron flux monitor to OPERABLE status, and 3. Once per 12 hours verify that the RCS boron concentration meets the requirement of LCO 3.9.1, using chemical analysis to determine the boron concentration of the reactor pressure vessel and the refueling canal.SURVEILLANCEREQUIREMENTS 4.9.2 As a minimum, two source range neutron flux monitors, one from each side of the reactor core, shall be demonstrated OPERABLE by performance of: a. Deleted b. Deleted c. A CHANNEL CHECK at least once per 12 hours, and d. A CHANNEL CALIBRATION prior to entry into MODE 6 if not performed within the last 18 months. Neutron detectors are excluded from CHANNEL CALIBRATION.
I DAVIS-BESSE.-
UNrT 1 3/4 9-2 Amendment No. 172, 269 "UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 269 TO FACILITY OPERATING LICENSE NO. NPF-3 FIRSTENERGY NUCLEAR OPERATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT 1 DOCKET NO. 50-346
 
==1.0 INTRODUCTION==
 
By application to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated December 20, 2004 (Agencywide Documents Access and Management System (ADAMS)Accession No. ML043580232), as supplemented by letter dated April 6, 2005 (ADAMS Accession No. ML050980109), FirstEnergy Nuclear Operating Company (the licensee)requested changes to the Technical Specifications (TSs) for the Davis-Besse Nuclear Power Station, Unit 1 (DBNPS). The supplement dated April 6, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on February 15, 2005 (70 FR 7765).S The proposed changes would revise TS 3/4.9.2, "Refueling Operations
-Instrumentation." Specifically, the proposed changes would revise TS 3/4.9.2 concerning source range flux monitors to be more consistent with improved Standard Technical Specifications (STSs). The proposed TS changes would achieve consistency with corresponding requirements in NUREG-1430, "Standard Technical Specifications Babcock and Wilcox Plants," Revision 3, dated June 2004, with exceptions to account for plant-specific design differences and retention of current licensing basis requirements and commitments.
In this safety evaluation (SE), the NRC staff refers to the current DBNPS TS as a CTS; a standard TS contained in NUREG-1430 as an STS; and the proposed DBNPS TS, which results from changes addressed in this SE, as a PTS. In addition to the requirements and considerations contained in NUREG-1 430, the NRC staff based its evaluation of the PTS changes on the Commission's "Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors" (Final Policy Statement), published on July 22, 1993 (58 FR 39132), and Title 10 of the Code of Federal Regulations (10 CFR)Section 50.36, "Technical specifications," as amended July 29, 1996 (61 FR 39299).Consistent with the Final Policy Statement, the licensee proposed transferring some CTS requirements to licensee-controlled documents, such as the TS Bases, for which changes are controlled by a regulation such as 10 CFR 50.59, "Changes, tests, and experiments." Accordingly, if 10 CFR 50.59 does not require prior NRC approval, such transferred requirements may be changed by the licensee without it. However, NRC-controlled documents, such as TSs, may not be changed by the licensee without prior NRC approval.
In addition, the licensee emphasized that the PTSs were requested to more closely match the improved STSs.During its review, the NRC staff relied on the Final Policy Statement,.
10 CFR 50.36, the STSs, and DBNPS TSs as guidance for accepting proposed CTS changes. This SE provides a
.1-2-summary basis for the NRC staff's conclusion that the licensee has developed the PTSs based on the STSs and CTSs, except for plant-specific considerations, and that the use of the PTSs are acceptable for continued operation of DBNPS. This SE also explains the NRC staff's conclusion that the PTSs are consistent with the DBNPS current licensing basis and conform to 10 CFR 50.36. In this SE, NRC staff conclusions regarding the conformance and consistency of the PTSs with these requirements are limited tospecifications in the CTSs related to the specific changes proposed by the licensee in the subject application, as supplemented.
The NRC staff also acknowledges that it is acceptable for the PTSs to differ from the STSs and DBNPS TSs to retain CTS provisions, that are based on the current licensing basis for DBNPS, which the NRC staff has previously reviewed and approved.For the reasons stated in this SE, the NRC staff finds that the PTSs issued with this license amendment comply with Section 182a. of the Atomic Energy Act (the Act), as amended, 10 CFR 50.36, and the guidance in the Final Policy Statement, and that they are in accordance with the common defense and security and provide adequate protection of the health and safety of the public.
 
==2.0 REGULATORY EVALUATION==
 
Section 182a. of the Act requires that applicants for nuclear power plant operating licenses will state:[S]uch technical specifications, including information of the amount, kind, and source of special nuclear material required, the place of the use, the specific S characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization
... of special nuclear material will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public. Such technical specifications shall be a part of any license issued.In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TSs. In doing so, the Commission placed emphasis on those matters related to the prevention of accidents and the mitigation of accident consequences; the Commission noted that applicants were expected to incorporate into their TSs "those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity," as set forth in the Statement of Consideration, "Technical Specifications for Facility Licenses; Safety Analysis Reports," (33 FR 18610, December 17, 1968). Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation:
(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) SRs; (4) design features; and (5) administrative controls.However, the rule does not specify the particular requirements to be included in a plant's TSs.NRC and industry representatives have developed guidelines for improving the content and quality of nuclear power plant TSs. On February 6, 1987, the Commission issued an interim policy statement on TS improvements, "Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors" (52 FR 3788). During the period from 1989 to 1992, the utility owners groups and the NRC staff developed improved STSs, such as* NUREG-1430, that would establish model TSs consistent with the Commission's policy for each primary reactor, nuclear steam supply system, and type. In addition, the NRC staff, licensees, and owners groups developed generic administrative and editorial guidelines in the form of a I-3-"Writer's Guide" for preparing TSs, which gives greater consideration to human factors principles.
FirstEnergy followed this guidance in the development of the TS changes proposed in the subject application for DBNPS.In June 2004, the Commission issued NUREG-1 430, Revision 3, which was developed using the guidance and criteria contained in the Commission's Interim Policy Statement.
The improved STSs in NUREG-1430 were established as a model for developing improved TSs for Babcock and Wilcox plants in general. The improved STSs reflect the results of a detailed review of the application of the interim policy statement criteria to generic system functions, which werepublished in a,"Split Report" issued to the nuclear steam supply system owners groups in May 1988. The improved STSs also reflect the results of extensive discussions between the owners groups and the NRC staff concerning a number of drafts and revisions, so that application of the Writer's Guide and the TS content criteria in 10 CFR 50.36 would consistently reflect detailed system configurations and operating characteristics for all nuclear steam supply system designs. As such, the improved STS Bases presented in NUREG-1430 provide an abundance of generally applicable information regarding the extent to which NUREG-1430 presents requirements that are necessary to protect public health and safety.With respect to the subject application, Section 3.9, "Refueling Operations," in NUREG-1430 applies to DBNPS.On July 22, 1993, the Commission issued its Final Policy Statement (58 FR 39132), expressing the view that satisfying the guidance in the policy statement also satisfies Section 182a. of the Act and 10 CFR 50.36. The Final Policy Statement described the safety benefits of the improved STSs, and encouraged licensees to use the improved STSs as the basis for S plant-specific license amendments.
The scope of such amendments that are based on the STSs range from partial conversions involving changes to one or more individual specifications, to complete conversions of a plant's TSs. The subject application for DBNPS proposes a partial conversion of the DBNPS CTSs related to refueling operations.
Further, the Final Policy Statement gave guidance for evaluating the required scope of a plant's TSs and defined four guidance criteria to be used in determining which existing TS limited condition for operation (LCO) requirements, including any associated Actions and Surveillance Requirements (SRs), should remain in the TSs. The Commission noted that, in allowing certain items to be relocated to licensee-controlled documents while requiring that other items be retained in the TSs, it was adopting the qualitative standard enunciated by the Atomic Safety and Licensing Appeal Board in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979).There, the Appeal Board observed:[WIhere is neither a statutory nor a regulatory requirement that every operational detail set forth in an applicant's safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is legally binding upon the licensee unless and until changed with specific Commission approval.
Rather, as best we can discern it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.By this approach, existing LCO (and associated) requirements that fall within or satisfy any of the criteria in the Final Policy Statement should be retained in the TSs; those LCO (and associated) requirements that do not fall within or satisfy these criteria may be relocated to  W licensee-controlled documents.
The Commission codified the four criteria set out in the Final Policy Statement in 10 CFR 50.36(c)(2)(ii)
(60 FR 36953, July 19, 1995). The four criteria are as follows: Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room (CR), a significant abnormal degradation of the reactor coolant pressure boundary.Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a design-basis accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.Criterion 3 A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to-mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.Criterion 4 A SSC which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.Part 3.0 of this SE provides the basis for the NRC staff's conclusion that the conversion of the DBNPS refueling operations CTSs to PTSs based on corresponding requirements in the STSs* and the CTSs, as modified by plant-specific considerations, is consistent with the DBNPS current licensing bases and the requirements and guidance of the Final Policy Statement and 10 CFR 50.36.3.0 TECHNICAL EVALUATION The NRC staff has organized this SE by identifying each proposed TS change as belonging to one of the following TS change categories:
Administrative A change that neither reduces nor increases the existing operational limitations and administrative controls for the facility.More Restrictive A change that increases an existing operational limitation or administrative control, or that adds a new operational limitation or administrative control for the facility.Less Restrictive A change that reduces~or deletes an existing (Specific) operational limitation or administrative control for the facility.Less Restrictive A change that involves moving detailed technical (Generic) information or requirements, which are inappropriate or unnecessary for inclusion in TSs, to licensee-controlled documents.
Grouping TS changes in these four categories is customary for evaluating applications that* propose to convert a facility's TSs to improved TSs, modeled on the STSs. The following subsections provide detailed discussions of the CTSchange categories, and the NRC staff's evaluations of the acceptability of changes under each category.
i 0 3.1 Administrative Changes Administrative changes, which are incidental to adopting STS format or phrasing, are intended to incorporate human factors principles into the form and structure of the TSs making them easier to understand and use by plant operations personnel.
These changes involve reorganizing, reformatting, and clarifying CTS requirements without affecting technical content or operational restrictions.
Among the administrative-type changes proposed by the licensee in the present application, and found acceptable by the NRC staff, are: " Deleting the words "As a minimum" in LCO 3.9.2, consistent with the wording in STSs;" Retaining the words "one from each side of the reactor core" in LCO 3.9.2, which is an appropriate plant-specific departure from the wording in STSs;" Changing the word "operating" to "OPERABLE" in LCO 3.9.2, consistent with the wording in STSs and the standard definition of OPERABLE in the CTSs and STSs;* Clarifying that CHANNEL CALIBRATION in SR 4.9.2 excludes the neutron detectors themselves, consistent with the wording in STSs;* Clarifying that in the case where there is no OPERABLE source range neutron flux monitor, ACTION a for only one OPERABLE source range neutron flux monitor must also be performed* The NRC staff reviewed all of the administrative changes proposed by the licensee and finds them acceptable because they are consistent with the Writer's Guide, STSs, and CTSs; do not result in any substantive change in operating requirements; and are consistent with the Commission's regulations.
3.2 More Restrictive Changes The licensee, in electing to implement the various specifications based on the STSs and the CTSs, proposed requirements that are more restrictive than those in the CTSs. PTSs in this category include requirements that are either new, more conservative than corresponding requirements in the CTSs, or that have additional restrictions that are not in the CTSs but are in the STSs.Specifically, the DBNPS LCO 3.9.2 ACTION statement would be revised to be similar to the STS ACTION statement, but continue to use the DBNPS TS format and add the STS wording to use the reactor coolant system refueling boron concentration requirement specified in DBNPS LCO 3.9.1. STS ACTION B.2 has no corresponding requirement in the CTSs to verify that boron concentration is within refueling limits. The corresponding STS specification (SR 3.9.1.1) refers to boron concentration limits specified in the COLR [Core Operating Limits Report], whereas the CTS boron concentration is determined from the kef, required by LCO 3.9.1 and this difference is reflected in the PTSs. Another proposed change is that the CHANNEL CHECK surveillance be performed every 12 hours throughout MODE 6 operation consistent with the STSs instead of just during CORE ALTERATIONS.
Changes such as these that are categorized as more restrictive are acceptable because they place additional limitations
* on plant operation that enhance safety.
Is-6-3.3 Less Restrictive Changes (Specific)
Less restrictive requirements include changes, deletions, and relaxations to CTS requirements.
When requirements have been shown to give little or no safety benefit, their removal from the TSs may be appropriate.
In most cases, relaxations previously granted to individual plants on a plant-specific basis were the result of (1) generic NRC actions, (2) new staff positions that have evolved from technological advancements and operating experience, or (3) resolution of the owners groups comments during the development of the STSs. The NRC staff reviewed generic relaxations contained in the STSs and found them acceptable because they are consistent with current licensing practices and the Commission's regulations.
The licensee did not propose any specific TS changes in this category.3.4 Less Restrictive Changes (Generic)When requirements have been shown to give little or no benefit, their removal from the TSs may be appropriate.
In most cases, relaxations previously granted to individual plants on a plant-specific basis were the result of,(1) generic NRC actions, (2) new staff positions that have evolved from technological advancements and operating experience, or (3) resolution of the owners group comments on STSs.A proposed change to the CTSs involved revision of the DBNPS SR 4.9.2 to be equivalent to the STS SRs, including deletion of the independent mention of the CHANNEL FUNCTIONAL TEST. Consistent with the STSs, the CHANNEL FUNCTIONAL TEST requirements are not* necessary to verify that the source range flux monitors are capable of satisfying the LCO requirements.
In MODE 6, the source range flux monitors are required for indication only; there are no required setpoints; and, the source range neutron flux detectors have no control function.
The NRC staff reviewed these generic relaxations contained in the PTSs and found them acceptable because they are consistent with current licensing practices and the Commission's regulations.
3.5 DBNPS TS Bases The licensee proposed conforming changes to the Bases for the specifications revised in this amendment.
The NRC staff has no objection to these Bases changes.The licensee stated in its supplemental letter dated April 6, 2005, that the information removed from the TSs and moved to the TS Bases will be adequately controlled under the TS Bases Control Program specified in TS 6.17. This program provides for evaluation of changes in accordance with 10 CFR 50.59 to ensure the Bases are properly controlled.
TheNRC staff notes that NUREG-1 430 Bases B 3.9.2, "Nuclear Instrumentation," clearly states that the source range neutron flux detectors "also provide continuous visual indication in the control room and an audible alarm to alert operators to a possible dilution accident." The licensee further statedthat the visual and audible indication features will remain installed in the plant and no changes to the source range indication features are currently contemplated.
The NRC staff expects that any future design change reviewed under 10 CFR 50.59 that reduces or eliminates these indication features would warrant prior NRC approval. 4.0 REGULATORY COMMITMENTS The licensee's letter dated December 20, 2004, contained the following regulatory commitment:
Associated changes to the TS Bases would be made under the provisions of the TS Bases Control Program. These changes are expected to include additional detail, including a discussion of the visual and audible indication features formerly referenced in the LCO.
 
==5.0 STATE CONSULTATION==
 
In accordance with the Commission's regulations, the Ohio State official was notified of the proposed issuance of the amendment.
The State official had no comments.6.0 ENVIRONMENTAL CONSIDERATION This amendment changes a requirement with respect to installation or use of a facility component located with 'in the restricted area as defined in 10 CER Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (70 FR 7765). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51 .22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
 
==7.0 CONCLUSION==
 
The NRC staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.Principal Contributor:
W. Macon Date: September 12, 2005 UNITED STATES oNUCLEAR REGULATORY COMMISSION REGION IV 611 RYAN PLAZA DRIVE, SUITE 400 Y0 ,ARLINGTON, TEXAS 76011-4006 4 January 25, 2006 Jack S. Keenan, Senior Vice President of Generation and Chief Nuclear Officer Mail Code B32 Pacific Gas and Electric Company P.O. Box 770000 San Francisco, CA 94177-0001
 
==Dear Mr. Keenan:==
 
==SUBJECT:==
ERRATA OF NRC INSPECTION REPORT 05000275/2004005 AND 05000323/2004005 This errata corrects the volume of water that was lost from the spent fuel pool on December 23, 2004, from 36,000 gallons to 3600 gallons. Please replace the first page of the Summary of.Findings and pages 14-16 of NRC Inspection Report 05000275/2004005 and 05000323/2004005, dated February 11, 2005, with the enclosed revised pages.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).Should you have any questions concerning this inspection, we will be pleased to discuss'them with you.Sincerely, IRA!William B. Jones, Chief'Project Branch B Division of Reactor Projects Dockets: 50-275 50-323 Licenses:
DPR-80 DPR-82
 
==Enclosure:==
 
Revised pages of NRC Inspection Report 05000275\2004005 and 05000323\2004005 Pacific Gas and Electric Company cc w/enclosure:
David H. Oatley, Vice President and General Manager Diablo Canyon Power Plant P.O. Box 56 Avila Beach, CA 93424 Donna Jacobs Vice President, Nuclear Services Diablo Canyon Power Plant P.O. Box 56 Avila Beach, CA 93424 James R. Becker, Vice President Diablo Canyon Operations and Station Director, Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 3 Avila Beach, CA 93424 Sierra Club San Lucia Chapter ATTN: Andrew Christie P.O. Box 15755 San Luis Obispo, CA 93406 Nancy Culver San Luis Obispo Mothers for Peace P.O. Box 164 Pismo Beach, CA 93448 Chairman San Luis Obispo County Board of Supervisors Room 370 County Government Center San Luis Obispo, CA 93408 Truman Burns\Robert Kinosian California Public Utilities Commission 505 Van Ness Ave., Rm. 4102 San Francisco, CA 94102-3298 Pacific Gas and Electric Company-3-Diablo Canyon Independent Safety Committee Robert R. Wellington, Esq.Legal Counsel 857 Cass Street, Suite D Monterey, CA 93940 Ed Bailey, Chief Radiologic Health Branch State Department of Health Services P.O. Box 997414 (MS 7610)Sacramento, CA 95899-7414 Richard F. Locke, Esq.Pacific Gas and Electric Company P.O. Box 7442 San Francisco, CA 94120 City Editor The Tribune 3825 South Higuera Street P.O. Box 112 San Luis Obispo, CA 93406-0112 James D. Boyd, Commissioner California Energy Commission 1516 Ninth Street (MS 34)Sacramento, CA 95814 Jennifer Tang Field Representative United States Senator Barbara Boxer 1700 Montgomery Street, Suite 240 San Francisco, CA 94111 Chief, Technological Services Branch FEMA Region IX Department of Homeland Security 1111 Broadway, Suite 1200 Oakland, CA 94607-4052 Pacific Gas and Electric Company-4-Electronic distribution by RIV: Regional Administrator (BSMI)DRP Director (ATH)DRS Director (DDC)DRS Deputy Director (RJCI)Senior Resident Inspector (TWJ)Branch Chief, DRP/B (WBJ)Senior Project Engineer, DRP/E (RAK1)Team Leader, DRP/TSS (RLNI)RITS Coordinator (KEG)DRS STA (DAP)V. Dricks, PAO (VLD)J. Dixon-Herrity, OEDO RIV Coordinator (JLD)ROPreports DC Site Secretary (AWC1)W. A. Maier, RSLO (WAM)SUNSI Review Completed:
_wbj__ ADAMS: Yes El No Initials:
_wbj_Publicly Available El Non-Publicly Available El Sensitive Non-Sensitive R:\ REACTORS\
DC\2004\DC2004-05RP Errata
: DRP/!B ._ _ <> _.....TWJackson;df WBJones T- WBJ IRA/ .1/25/06 1/25/06 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
 
==SUMMARY==
OF FINDINGS IR 05000275/2004-005, 05000323/2004-005; 10/01/04 -12/31/04; Diablo Canyon Power Plant Units 1 and 2; Operability Evaluations, Event Followup, Personnel Performance Related to Nonroutine Plant Evolutions and Events, Equipment Alignment, Access Control To'Radiologically Significant Areas, Other.This report covered a 13-week period of inspection by resident inspectors and announced inspections in the areas of inservice inspections, emergency preparedness, and radiation protection.
Five self-revealing, four NRC-identified Green noncited violations, and one unresolved item with potential safety significance greater than Green were identified.
The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609 "Significance Determination Process." Findings for which the.Significance Determination Process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1 649, "Reactor Oversight Process," Revision 3, dated July 2000.A. NRC-Identified and Self-Revealing Findinqs Cornerstone:
Initiating Events Green. A self-revealing noncited violations was identified for the failure to appropriately implement the procedure for spent fuel pool skimmer filter replacement, as required by Technical Specification 5.4.1 .a. On December 23, 2004, operators cleared the spent fuel pool skimmer system using Section 6.3.1 of Procedure OP B-7:111, "Spent Fuel Pool System -Shutdown and Clearing and Filter Replacement," Revision 15, instead of the appropriate section, which was Section 6.3.2. A human performance crosscutting aspect was identified for the failure on two occasions to address configuration control concerns with the system.This finding impacted the Initiating Events Cornerstone and was considered more than minor using Example 5.a of IMC 0612. Specifically, Valve SFS-2-3 was mis-positioned due to the use of the wrong section of Procedure OP B-7:111 and then returned to service. Additionally, operators had two opportunities to identify the mis-positioning of Valve SFS-2-3 but failed to identify the condition.
The mis-positioned valve resulted in a loss of approximately 3600 gallons of water from the spent fuel pool. This finding was reviewed by NRC management in accordance with IMC 0609 and 0612 and determined to be of very low safety significance (Section 1 R1 4.2).Cornerstone:
Mitigating Systems Green. A self-revealing, noncited violation was identified for the failure to setup phase sequence test equipment according to procedure, as required by 10 CFR Part 50, Appendix B, Criterion V. This failure resulted in the momentary de-energization of Vital 4kV Bus G and the auto-start of Diesel Engine Generator 2-1. Subsequent investigation by Pacific Gas & Electric Company revealed that the primary side of the test transformer was wired in a wye configuration instead of a' delta configuration.
This Enclosure  were also evident for the feedwater level controller malfunction.
The inspectors, determined that with the information provided in the procedure and the plant conditions, that there was sufficient evidence to result in the shift foreman deciding to trip the reactor and close the main steam isolation valves: Furthermore, the inspectors observed that PG&E had not developed a procedural bases for the actions specified by Step 5.1.1. A human performance crosscutting aspect (resources) was identified for the inadequate alarm procedure.
The inspectors are reviewing the adequacy of alarm response Procedure AR PK 10-21 to address a feedwater heater level control malfunction as an unresolved item.Analysis.
No analysis was performed for this unresolved item.Enforcement.
Unresolved Item (URI) 50-323/04-05-03,'
Adequately of Alarm Procedure For Feedwater Heater Level Control Malfunctions.
.2 Unit 2 Spent Fuel Pool (SPF) Level Drop a. Inspection Scope On December 23, 2004, the Unit 2 SPF level dropped approximately 4 inches as a result of Valve SFS-2-3, SFP skimmer pump casing drain to miscellaneous equipment drain tank, being left open following a filter replacement.
The inspectors observed operator actions and equipment performance following the event. The inspectors also interviewed operations personnel and reviewed the event for corrective actions, violation of requirements, and generic issues.b. Findings Introduction.
A Green, self-revealing NCV was identified for the failure to appropriately implement the procedure for SFP skimmer filter replacement, as required by Technical Specification 5.4.1 .a. This failure resulted in a loss of approximately 3600.gallons of water from the SFP.Description.
On December 23, 2004, operators implemented Clearance 79718 for replacing the SFP skimmer filter. Attached to the clearance was Procedure OP B-7:11I,"Spent Fuel Pool System -Shutdown and Clearing and Filter Replacement," Revision 15. Section 6.3.1 of the procedures for shutting down and clearing the skimmer pump and strainer had been marked for implementation.
Following the implementation of the clearance, the work control lead observed that Section 6.3.1 of Procedure OP B-7:111 was used, when Section 6.3.2, steps 'a' through 'e', should have been used. Section 6.3.2 of the procedure specifically addressed replacement of the SFP skimrner filter. The work control lead marked steps 'g' .through T of Section 6.3.2 Enclosure  for returning the SFP skimmer pump back to service. He noticed that, because Section 6.3.1 had been used to clear the pump, 4 valves would be potentially mis-positioned.
The work control lead discussed the potential for the 4 valves to be potentially mis-positioned with the oncoming shift work control lead.Following SFP skimmer filter replacement, the oncoming shift work control lead informed operators to restore the SFP skimmer system using Section 6.3.2. The work control lead also informed the operators that he was not sure how the SFP skimmer system had been cleared by the previous shift. Operators restored the SFP skimmer system, and when they started the system, they found 3 valves mis-positioned.
Approximately 3 hours later operators noticed a steady increasing level in the miscellaneous equipment drain tank. Operators then found that Valve SFS-2-3 was still mis-positioned from the clearance of the skimmer pump. For the 3 hours that Valve SFS-2-3 was mis-positioned, approximately 3600 gallons of water was drained from the SFP.The inspectors determined that PG&E failed to properly implement Procedure OP B-7:111 when clearing the SFP skimmer system. Section 6.3.2 specifically addressed replacement of the SFP skimmer filter. The inspectors also observed that other operators were aware of a potential mis-position of valves. However, the need for checking the alignment of these valves had not been adequately communicated to and/or carried out by the operators who restored the SFP skimmer system. The operators who restored the SFP skimmer system recognized and corrected the 3 mis-positioned valves, but failed to adequately investigate the reason for the mis-position, which was a missed opportunity to discover the 4 th mis-positioned valve. A human performance cross cutting aspect was identified for the failure on two occasions to address configuration control concerns with the system.Analysis.
The performance deficiency associated with this event is the failure to properly implement Procedure OP B-7:111 as required by Technical Specification 5.4.1 .a.This deficiency impacted the Initiating Events Cornerstone that limit the likelihood of events that upset plant stability during shutdown and affected the configuration control attribute for operating equipment lineup. The finding was considered more than minor using Example 5.a of Inspection Manual Chapter 0612. Specifically, Valve SFS-2-3 was mis-positioned due to the use of the wrong section of Procedure OP B-7:lII and then returned to service. Additionally, operators had two opportunities to identify the mis-positioning of Valve SFS-2-3 but failed to identify the condition.
The mis-positioned valve resulted in a loss of approximately 3600 gallons of water from the spent fuel pool.This finding was reviewed by NRC management in accordance with Inspection Manual Chapter 0609 and 0612 and determined to be of very low safety significance.
This determination was based on the performance deficiency would not have resulted in a loss of spent fuel pool inventory below the Technical Specification required level on a loss of spent fuel pool cooling.Enclosure  Enforcement.
Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. Item 3.h of Regulatory Guide 1.33, Appendix A recommends procedures for startup, operation, and shutdown of fuel storage pool purification and cooling systems. Contrary to the above, PG&E failed to properly implement Procedure OP B-7:111 with regards to replacing the SFP skimmer filter. The failure to properly implement this procedure resulted in mis-position of Valve SFS-2-3 and the loss of approxirhately 3600 gallons of water from the SFP. Because the failure to properly implement Procedure OP B-7:111 is of very low safety significance and has been entered into the corrective action system as AR A0628635, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 50-323/04-05-04, Failure to Properly Implement Procedure for Spent Fuel Pool Skimmer Filter Replacement.
1R15 Operability Evaluations (71111.15).
: a. Inspection Scope The inspectors reviewed seven inspection samples of operability evaluations.
These reviews of operability evaluations and/or prompt operability assessments and supporting documents were performed to determine if the associated systems could meet their intended safety functions despite the degraded status. The inspectors reviewed the applicable Technical Specification, Codes/Standards, and Final Safety Analysis Report Update sections in support of this inspection.
The inspectors reviewed the following AR's and operability evaluations: (Unit 2) Environmental qualification of auxiliary feedwater flow indication cable (ARs A0620857, A0621502)(Unit 1) Emergency core cooling system (ECCS) voiding (AR A0621502)(Unit 1) Startup Transformer 1-1 automatic tap changer in manual due to unexpected step increases (AR A0625650)(Unit 2) Residual Heat Removal Pump 2-2 socket weld crack at suction pressure instrument line (AR A0624790)(Units 1 and 2) Valve FW-2-LCV-1 10 failed closed (AR A0624790)(Unit 2) DEG 2-3 lube oil instrument line crack, (AR A0617419)(Unit 1) Small water drip on feedwater pipe lead 2-2 (AR A0628484)Enclosure Chapter 5.0 RAIs NRC ITS Tracking Page I of 2~1 sgn~.Return to View Menu~ d Print Document, RAI Screening Required:
Yes Status: Closed This Document will be approved by: Tim Regulatory Basis must be included in Comments Kobetz section of this Form This document has been reviewed and Yes information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING NRC Reviewer ID 200711061308 Conference Call Requested?
No Category In Scope ITS Section:.
TB POC.: JFD Number: Page Number.(s):
ITS 5.0 Gerald Weig Gerald Waig 18 111 Information ITS Number:n.
0S1.:. DOC.Number:
Bases.-JFD Number:.5.5 None None None NRC Author Gerald Waig Question: Please provide additional information to clarify ITS Section 5 JFD number 18 by providing a definition of the word "temporary"'
as used in IST 5.5.11.b and identify which outdoor liquid storage tanks are considered to be temporary per this specification.
Also identify the source document where this information can be found.Discussion:
As written, IST 5.5.11.b would appear to only apply to "all outdoor'temporary' liquid storage tanks .......The CTS LCO 3.11.1, from which 5.5.11.b is taken, in part, applies to "Liquid Holdup Tanks" that is further clarified by asterisk footnote as "Tanks Comment included in this specification are those outdoor tanks that are not surrounded
.........................
by liners, dikes or walls capable of holding the tank contents or that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system." Note that the CTS 3.11.1a refers to"...tank" (singular vs plural in other references).
The ISTS clearly applies to all liquid radwaste tanks not surrounded by liners, dikes, or walls...connected to the liquid radwaste treatment system" to ensure compliance with the limits of 10 CFR 20, Appendix B, Table 2, Column 2 as it relates to water supply in an unrestricted area in the event of an uncontrolled release of the tank contents.Addition information is needed to define "outdoor temporary storage tanks to deterine if the JFD adequately supports the deviation from ISTS since the ITS wording is not as it appears in CTS.http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 Regulatory
 
==Reference:==
 
10 CFR 50.36(d)(5)
Administrative Controls Issue Date l 11/06/2007 Close Date [01/30/2008 Logged in User: Jerry Jones"'Responses Licensee Response by Jerry There is no specific definition of "temporary" as it applies to the Jones on 11/30/2007.
Davis-Besse Current Technical Specification (CTS) 3.11.1 (Volume 16, Page 68). As stated in Justification for Deviation (JFD) 18, "temporary" is consistent with the current licensing basis, as CTS LCO 3.11.1 states "The quantity of radioactive material contained in each of the following..." and CTS LCO 3.11.1.a states that the only tank is the "Outside temporary tank." A reasonable definition of temporary would be "lasting or serving for a limited time." In other words, not part of the Davis-Besse permanent plant structures.
Even though CTS 3. 11.1.a refers to"tank" in the singular, every other reference to tanks is plural within CTS 3.11.1. As an additional piece of information, UFSAR 12.1.3.4 (included as an attachment to this response) describes the sources of radioactivity stored outside the station buildings.
The only tank that can contain radioactivity is the Borated Water Storage Tank (BWST), which is a permanent plant structure.
Furthermore, UFSAR 3.8.1.1.5 (the applicable portion is also included as an attachment) discusses the evaluation of the potential consequences of a rupture of the BWST. Therefore, ITS 5.5.1 1.b (Page 100) only applies to "temporary" tanks, consistent with the current licensing basis of Davis-Besse, and JFD 18 is accurate -the change is consistent with the current licensing basis. Therefore,_no changes to the ITS submittal are required.Licensee Response by Bryan After further review, Davis-Besse has decided to change CTS Kays on 01/27/2008 3.11.1 (Volume 16, Page 68) to include all outdoor liquid storage tanks. This will change ITS 5.5.11 .b (Page 100) back to the ISTS.Additionally, CTS LCO 3.11.1 has been 'changed to match the ISTS 5.5.12.b, in regards to the 10 CFR 20 requirements.
A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
Date Created: 11/06/2007 01:08 PM by Gerald Waig Last Modified:
01/30/2008 11:02 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/I fddcea 1Od3bdbb585256e8...
6/2/2008 D-B with fluid identical to the effluent from the nearest upstream major piece of equipment and in some instances, to the letdown from the primary system.12.1.3.2 Process Piping Normally, no process piping is field-run.
When it is necessary to add piping, layouts for such piping are planned and drawn up, and then reviewed by the engineering staff for adequacy prior to 7 approval.
Once approved, the layouts are used in conjunction with existing process piping layouts and radiation zone diagrams to guide field personnel in the safe and proper installation of the new piping.12.1.3.3 Spent Fuel Assembly The source term for a spent fuel assembly is given in Table.15A-3.
12.1.3.4 Radioactivity Stored Outside 20 Sources of radioactivity normally stored outside the station buildings, other than the Dry Fuel Storage Facility are small amounts contained in the low level radwaste storage areas and in the 24 borated water storage tank. There is no dose at the unrestricted area boundary (site boundary)associated with these areas.The Dry Fuel Storage Facility (DFSF) is being used under the licensing provisions of I0CFR72, 20 Subparts.K and L. The radiological impact of this facility has been analyzed and the site boundary dose remains below the 10CFR100 guidelines.
12.1.3.5 Turbine Building Nitrogen-16 Gamma Dose Since Davis-Besse utilizes a PWR, the contribution of N-16 to the gamma dose within the turbine building is negligible.
12.1.4 Area Monitoring The area radiation monitors installed for station personnel protection are listed in Table 12.1-3.The areas selected and number of points monitored have been coordinated with the station radiation access control requirements to provide the operating personnel with a knowledge of the areas containing high radiation levels. In general, area radiation monitors are placed in areas where radiation levels could feasibly increase due to postulated occurrences.
Refer to Section 20 11.4 for design and maintenance details common to all radiation monitors.123' Each monitor channel consists of three remotely located, interconnected subsystems:
: 1. the detector, located in the monitored area, 2. the local readout unit, located in or adjacent to the monitored area, and 12.1-7 REV 24 06/04 D-B, 2. The nitrogen storage tank, which is the nearest potential missile source is enclosed in a structure capable of sustaining potential missiles from this source.Evaluation of the consequences of a postulated rupture of the BWST demonstrated that expected exposures will be within the limits set by 1OCFR100 and 1OCFR20.3.8.1.1.6 Influence of Seismic Class II Structures on Seismic Class I Structures There is no significant influence of any ClasslI structure on the Class I structures.
The following Class I equipment/systems or structures are protected from any possible failure of Class II structures which enclose them.a. Class I service water pumps and piping in the Intake Structure are below the top concrete slab and the Class 11 intake steel superstructure, The top concrete slab is 21 inches thick and is the protecting membrane against the unlikely failure of the Class II structure.
In addition, the Class HI Intake Structure Gantry Crane is 24 analyzed for SSE and found to have adequate capacity and therefore will not collapse onto the Intake Structure.
: b. Class I service water piping in the Class II Turbine Building is enclosed in Class I reinforced concrete pipe tunnel that is completely surrounded by a granular compacted fill with a minimum top cover of four feet. The 10 inch concrete ground floor slab bears on the compacted fill. The reinforced concrete tunnel, 4 feet of compacted fill cover and the 10 inch concrete ground floor slab protect the Class I piping against the unlikely failure of the Class II Turbine Building superstructure.
: c. A small portion of the Class I reinforced concrete Auxiliary Building supports the structural steel framing for the heater bay of the Class 11 Turbine Building.Multi-level steel floor framing, the elevated and ground floor concrete slabs in the heater bay, and the reinforced concrete Auxiliary.Building walls and slabs protect the Class I structure from the unlikely failure of the Class II structure and/or equipment.
3.8.1.1.7 Load Combinations The design of the above Class I structures is based on the load combinations presented in Subsection 3.8.1.3, ACI Code 318-63 and seismic analysis for Class I structures.
All of the reinforced concrete structures were designed by the Ultimate Strength Method. All of the structural steel was designed by the Working Stress Method.3.8-9 REV 24 06/04 NRC ITS Tracking Page I of-4[1 s Jgn &#xfd; Return to View Menu 4.Print Documenti RAI Screening Required:
Yes This Document will be approved by: George Wilson This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC ITS TRACKING NRC Reviewer ID 200711161152 Conference Call Requested?
No Category BSI -Beyond Scope Issue ITS Section: TB POC; JFD Number: Page Number(s):
ITS 5.0 Gerald Weig Matthew None 107 ITS Number: McConnell D OC Numnber: Bases JFD Numnber: Information 5.5 OSI: None None None NRC Author Matthew McConnell TS 5.5.16 is not consistent with the wording that was agreed upon at the July 12, 2006, TSTF-360 public meeting. Provide a detailed justification for deviating from the industry approved wording.10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established Comment for each structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Issue Date 11/16/2007 CloseI!)ate 04/24/2008 Logged in User: Jerry Jones'Responses I Licensee Response by Jerry Davis-Besse has submitted ITS Section 5.5.16, consistent with the 1 http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 4.Jones on 12/11/2007 currently approved NRC NUREG-1430, Revision 3.1.Furthermore, NRC Administrative Letter 96-04, "Efficient Adoption of Improved Standard Technical Specifications," specifically states that beyond scope issues, which are generally characterized as changes that differ from both the existing technical specifications and the improved STS (thus, the NRC described change is a beyond scope issue), tend to unnecessarily complicate and delay the conversion review process. In addition, during discussions with the NRC concerning the Davis-Besse ITS submittal, held prior to the date of the Davis-Besse ITS submittal; the NRC reiterated this fact and cautioned Davis-Besse about the effect beyond scope changes could have on the ITS approval date.Thus, Davis-Besse did not include these "agreements," which had not been submitted to the NRC in TSTF form yet, in our ITS submittal.
Subsequent to the Davis-Besse ITS submittal, the industry submitted TSTF-500, which proposed implementation of the agreements between the NRC and the industry.
Furthermore, Davis-Besse is not aware of when, or if, the TSTF will be approved by the NRC. However, Davis-Besse has reviewed the NRC meeting minutes for the July 12, 2006 meeting and proposed TSTF and will adopt the new requirement that requires actions to verify the remaining cells are greater than 2.07 V when a pilot cell or cells have been found to be less than 2.13 V, provided this new beyond scope change does not delay NRC approval of the Davis-Besse ITS. A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
Additionally, Davis-Besse does not believe that the ITS 5.5.16 change deleting the IEEE 450-1995 reference should be made since the NRC has specifically endorsed the use of this version of IEEE 450 for Davis-Besse, as described in our response to Question 200711161200.
Also, the NRC meeting minutes states to move the IEEE-450 reference to the ISTS Bases. However, ISTS Chapter 5.0 does not have a Bases, nor is the Battery Program specifically referenced or required as part of any LCO or Surveillance.
Thus, there is no Bases to put the IEEE 450 requirement in. Davis-Besse believes that the Battery Program should reference what it is based upon. Therefore, the IEEE 450-1995 reference will be maintained in the Davis-Besse ITS submittal.
NRC Response by Matthew McConnell on 02/21/2008 During a 1/29/08 and a 2/20/08 conference call, the staff informed the licensee of the revised TS wording for the battery monitoring and maintenance program that was agreed upon by the TSTF-360/500 working group and the NRC. The licensee agreed to follow-up on this item as a restult of the call.Licensee Response by Bryan Kays on 04/01/2008 This response supersedes the response of 12/11/2007.
Davis-Besse has submitted ITS Section 5.5.16, consistent With the currently approved NRC NUTREG- 1430, Revision 3.1. Furthermore, NRC Administrative Letter 96-04, "Efficient Adoption of Improved Standard Technical Specifications," specifically states that beyond scope issues, which are generally characterized as changes that http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddCea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 3 of 4 differ from both the existing technical specifications and the improved STS (thus, the NRC described change is a beyond scope issue), tend to unnecessarily complicate and delay the conversion review process. In addition, during discussions with the NRC concerning the Davis-Besse ITS submittal, held prior to the date of the Davis-Besse ITS submittal, the NRC reiterated this fact and cautioned Davis-Besse about the effect beyond scope changes could have on the ITS approval date. Thus, Davis-Besse did not include these "agreements," which had not been submitted to the NRC in TSTF form yet, in our ITS submittal.
Subsequent to the Davis-Besse ITS submittal, the industry submitted TSTF-500, which proposed implementation of the agreements between the NRC and the industry.
Furthermore, Davis-Besse is not aware of when, or if, the TSTF will be approved by the NRC. However, Davis-Besse has reviewed the NRC meeting minutes for the July 12, 2006 meeting and proposed TSTF and will adopt the new requirement that requires actions to verify the remaining cells are greater than 2.07 V when a pilot cell or cells have been found to be less than 2.13 V, provided this new beyond scope change does not delay NRC approval of the Davis-Besse ITS. Additionally, Davis-Besse agrees to put a generic reference to IEEE Standard 450.Although there will be no mention of a specific IEEE Standard 450 in this Chapter, Davis-Besse is currently committed to IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead Acid Batteries for Stationary Applications." Furthermore, by making this change, Davis-Besse is not making any changes to the current committment.
A draft markup regarding these changes is attached.This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
Licensee Response by Jerry Jones on 04/23/2008 This response supersedes the response of 4/1/2008.
Davis-Besse has submitted ITS Section 5.5.16, consistent with the currently approved NRC NUREG-1430, Revision 3.1. Furthermore, NRC Administrative Letter 96-04, "Efficient Adoption of Improved Standard Technical Specifications," specifically states that beyond scope issues, which are generally characterized as changes that differ from both the existing technical specifications and the improved STS (thus, the NRC described change is a beyond scope issue), tend to unnecessarily complicate and delay the conversion review process. In addition, during discussions with the NRC concerning the Davis-Besse ITS submittal, held prior to the date of the Davis-Besse ITS submittal, the NRC reiterated this fact and cautioned Davis-Besse about the effect beyond scope changes could have on the ITS approval date. Thus, Davis-Besse did not include these "agreements," which had not been submitted to the NRC in TSTF form yet, in our ITS submittal.
Subsequent to the Davis-Besse ITS submittal, the industry submitted TSTF'500, which proposed implementation of the agreements between the NRC and the industry.
Furthermore, Davis-Besse is not aware of when, or if,,the TSTF will be approved by the NRC. However, http://www.excelservices.com/exceldbs/itstrack'_davisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 4 of 4 Davis-Besse has reviewed the NRC meeting minutes for the July 12, 2006 meeting and proposed TSTF and will adopt the new requirement that requires actions to verify the remaining cells are greater than 2.07 V when a pilot cell or cells have been found to be less than 2.13 V, provided this new beyond scope change does not delay NRC approval of the Davis-Besse ITS. Additionally, Davis-.Besse agrees to delete the reference to IEEE Standard 450.Although there will be no mention of IEEE Standard 450 in this Chapter, Davis-Besse is currently committed to IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead Acid Batteries for Stationary Applications." Furthermore, by making this change, Davis-Besse is not making any changes to the current commitment.
A draft markup regarding these changes is attached.This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
NRC Response by Matthew McConnell on 04/24/2008 The staff understands that the licensee plans to use the recommendations provided in the Institute of Electrical and Electronics Engineers (IEEE) Standard (Std.) 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," to develop the proposed battery monitoring and maintenance program prescribed by new TS 5.5.16. However, the staff would like to note that this version of IEEE Std. 450 has not been officially endorsed by the NRC. The licensee adequately responded to the Electrical Engineering Branch (EEEB) staffs request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 11:52 AM by Matthew McConnell Last Modified:
04/24/2008 02:48 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddceal Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2 Asin /Return to View Menu IQPrnt Doc .lent RAI Screening Required:
Yes Status: Closed This Document will be approved by: George Regulatory Basis must be included in Comments Wilson section of this Form This document has been reviewed and Yes information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING NRC Reviewer ID 11200711161153 Conference CalRequested?
No Category BSI -Beyond Scope Issue ITS Section: TB POC: JFD Number: Pa ge Nu mber(s): ITS 5.0 Gerald Weig Matthew None 107 ITS Nunmber: McConnell DOC.Nu.mb.er.:.
Bases .,D Number: Information 5.5 OSI: None None None NRC Author Matthew McConnell Provide assurance that the relocated battery parameter values (e.g., visual inspection criteria, cell-to-cell connection resistance values, specific gravity monitoring, etc.) will continue to be controlled at their current level and actions will be implemented in accordance with the licensee's corrective action program. This is consistent with the industry resolution that was reached following the July 12, 2006, TSTF-360 public meeting.10 CFR 50.36, "Technical Specifications," requires that the technical specifications (TS) must be derived from the analyses and evaluation in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. A TS limiting condition for operation (LCO) must be established for each structure, system, or component that is part of the primary success Comment path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCOs specify minimum requirements for ensuring safe operation of the unit. Surveillance requirements (SRs) are TS requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that LCOs for operation will be met.Paragraph 50.65(a)(3) of 10 CFR, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," requires, in part, that"Performance and condition monitoring activities and associated goals and preventive maintenance activities shall be evaluated at least every refueling cycle provided the interval between evaluations does not exceed 24 months ...http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/
1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 6f 2 Adjustments shall be made where necessary to ensure that the objective of preventing failures of structures, systems, and components through maintenance is appropriately balanced against the objective of minimizing unavailability of structures, systems, and components due to monitoring or preventive maintenance.", Issue Date 11/16/2007 Close Date 02/27/2008 Logged in User: Jerry Jones'Responses Licensee Response by Jerry ITS 5.5.16 (Volume 16, Page 107) states that the Battery Jones on 12/11/2007 Monitoring and Maintenance Program provides for battery maintenance based on the recommendations of IEEE 450-1995.(See response to Question 200711161200 concerning maintaining IEEE 450-1995 references.)
Thethree battery parameter values referenced in the NRC reviewer's question all are specified in IEE 450-1995.
Davis-Besse intends on maintaining the CTS Surveillances related to these parameters (CTS 4.8.2.3.2.a.
1 (gravity checks), 4.8.2.3.2.b.2, 4.8.2.3.2.c.
1, 2 and 3, and 4.8.2.3.2.b.
1 (gravity checks) -Pages 209 and 210) as part of the implementation of ITS 5.5.16. Furthermore, Davis-Besse will commit to placing these CTS Surveillances in the Battery Monitoring and Maintenance Program. This commitment will be__documented in the supplement to the ITS Conversion Amendment.
NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on 02/21/2008 Branch (EEEB)staff s request for additional information.
Therefore, EEEB has no further questions at this time.Date Created: 11/16/2007 11:53 AM by Matthew McConnell Last Modified:
02/27/2008 01:52 PM http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/
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6/2/2008 NRC ITS Tracking Page I of 2~Assign~Return to View Menu a rnt Do.tin RAI Screening Required:
Yes This Document will be approved by: George Wilson This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC ITS TRACKING NRC' Rpvipiwir ID 1200711161200 Conference Call Requested?
No Category [ BSI -Beyond Scope Issue ITS Section: TB-POC: JFD Number&#xfd; Page.Number(s).:
ITS 5.0 Gerald Weig Matthew' None 107 ITS .Number: McConnell D OC.-Number:
Bas.es.JFD..
N.unib.er:.
Information 5.5 OS: None None None NRC Author [Matthew McConnell The Institute of Electrical and Electronics Engineers (IEEE) Standard (Std.)450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," is not endorsed by the NRC. Therefore, consistency with this IEEE Std. is not an adequate justification for approving TS changes. The latest version of IEEE Comment Std. 450 that is endorsed by the NRC is IEEE Std. 450-2002 as endorsed by Regulatory Guide 1.129, Rev. 2, "Maintenance, Testing, and Replacement of Vented Lead-Acid Storage Batteries for Nuclear Power Plants." Provide a revised justification for all TS changes that use consistency with IEEE Std.450-1995 as the basis for approval.
This is consistent with the industry resolution that was reached following the July 12, 2006, Technical Specifications Task Force (TSTF)-360 public meeting.Issue =Date 11/16/2007
]Close Date [102/01/2008 Logged in User: Jerry Jones"'Responses Licensee Response by Jerry Jones on 12/11/2007 While the NRC may not endorse IEEE standard 450-1995 for use generically, the NRC has specifically endorsed the use of this version of IEEE 450 for Davis-Besse.
Davis-Besse proposed, and the NRC approved, changes to certain battery requirements in http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 License Amendment 229, dated 2/9/1999.
In the NRC Safety Evaluation for this license amendment, the NRC specifically stated that the changes were based on both the ISTS (NUREG-1430) and IEEE 450-1995, and that the changes were acceptable.
The NRC also issued the Bases as part of the license amendment, and the Bases states that the Surveillances are based, in part, on IEEE 450-1995. The License Amendment and the NRC Safety Evaluation are attached for your information.
At this time, Davis-Besse intends to maintain our commitment to the 1995 version, as approved by the NRC. Thus, no changes to the ITS submittal are required.NRC Response by Matthew The licensee adequately responded to the Electrical Engineering McConnell on 02/01/2008 Branch (EEEB)staff s request for additional information.
] Therefore, EEEB has no further questions at this time.-Date Created: 11/16/2007 12:00 PM by Matthew McConnell Last Modified:
02/01/2008 11:36 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008
-pUNITED STATES LO , C) 5 21 o NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-MiO1 February 9, 1999 C Mr. ~John K. Wood Vice President
-Nuclear FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station 5501 North State Route 2 TOLEDO EDISON Oak Harbor, OH 43449-9760
 
==SUBJECT:==
AMENDMENT NO. 229 TO FACILITY OPERATING LICENSE NO. NPF-3 -DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. I (TAC NO. MA3952)
 
==Dear Mr. Wood:==
The Commission has issued the enclosed Amendment No. 229 to Facility Operating License No. NPF-3 for the Davis-Besse Nuclear Power Station, Unit No. 1. The amendment revises the Technical Specifications (TSs) in response to your application dated October 27, 1998.This amendment revises TS 3/4.8.2.3, "Electrical Power Systems -DC Distribution
-Operating," and the associated bases. The surveillance requirements for battery testing have been revised.A copy of the Safety Evaluation is also enclosed.
Notice of issuance will be included in the Commission's next biweekly Federal Register notice.Sincerely, Allen G. Hansen, Project Manager Project Directorate 111-2 Division of Reactor Projects III/[V Office of Nuclear Reactor Regulation Docket No. 50-346
 
==Enclosures:==
: 1. Amendment No. 229 to License No. NPF-3 2. Safety Evaluation cc w/encls: See next page John K. Wood FirstEnergy Nuclear Operating Company cc: MaryE. O'Reilly FirstEnergy Davis-Besse Nuclear Power Station 5501 North State -Route 2'Oak Harbor, OH .43449-9760 James L_ Freels Manager -Regulatory Affairs FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station 5501 North-State
-Route 2 Oak Harbor, OH 43449-9760 Jay E. Silberg, Esq.Shaw, Pittman, Potts and Trowbridge.
2300 N Street, NW.Washington, DC 20037 Regional Administrator U.S. Nuclear Regulatory Commission 801. Warrenville Road Lisle, IL 60523-4351 Robert B. Borsum., Babcock & Wilcox Nuclear Power Generation Division 1700 Rockville Pike, Suite 525 Rockville, MD 20852.Resident Inspector U.S. Nuclear Regulatory Commission 5503 North State Route 2 Oak Harbor, OH 43449 James H. Lash, Plant Manager FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station 5501 North State Route 2 Oak Harbor, OH 43449-9760 Davis-Besse Nuclear Power Station, Unit 1 Robert E. Owen, Chief Bureau of Radiological Health Service Ohio Department of Health P.O. Box 118 Columbus, OH 43266-0118 James R. Williams, Chief of Staff Ohio Emergency Management Agency 2855 West Dublin Granville Road Columbus, OH 43235-2206 Donna Owens, Director Ohio Department of Commerce Division of Industrial Compliance Bureau of Operations
& Maintenance 6606 Tussing Road P.O. Box'4009 Reynoldsburg, OH 43068-9009 Ohio Environmental Protection Agency DERR--Compliance Unit ATTN: Zack A. Clayton P.O. Box 1049 Columbus, OH 43266-0149 State of Ohio Public Utilities Commission 180 East Broad Street Columbus, OH 43266-0573 Attorney General Department of Attorney 30 East Broad Street Columbus, OH 43216 President, Board of County Commissioners of Ottawa County Port Clinton, OH 43252 UNITED STATES a NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FIRSTENERGY NUCLEAR OPERATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 229 Licens'e No. NPF-3 1. The Nuclear Regulatory Commission
(:the Commission) has foundthat:
A. The application for amendment by the Toledo Edison Company, Centerior Service Company, and The Cleveland Electric Illuminating Company (the licensees on the date of application;.FirstEnergy Nuclear Operating Company became the sole licensed operator on January 1, 1999) dated October 27, 1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act)-, and the Commission's rules and regulations set forth in 10 CFR Chapter I;B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized.
by this amendment can be conducted without endangering the health and safety of the-public, and (ii) that such activities wil.l be.conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public;and.E. The issuance of this amendment is in~accordance-with 10 CFR Part.51 of.the Commission's regulations and all applicablerequirements have been satisfied.
: 2. Accordingly, the license is amended by changes to.the Technical" Specifications as indicated in the att'achment to this license'amendment, and paragraph 2.C.(2) of Facility Operating, License No. NPF-3 is hereby amended to read as follows:  (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 229, are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications.
: 3. This license amendment is effective as of its date of issuance and shall be implemented not later than 120 days after issuance.FOR THE U.S. NUCLEAR REGULATORY COMMISSION Allen G. Hansen, Project Manager Project Directorate 111-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
 
==Attachment:==
 
Changes to the Technical Specifications Date of issuance:
February 9, 1999 ATTACHMENT TO lICENSE AMENDMENT NO. 229 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.Remove Insert 3/4 8-9 3/4 8-10 B 3/4 8-la 3/4 8-9 3/4 8-10 B 3/4 8-1a ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: 2. Verifying total battery terminal voltage is greater than or equal to 129 volts on float charge.b. At least once per 92 days and within 7 days after a battery discharge (battery terminal voltage below 110 volts), or battery overcharge (battery terminal voltage above 150 volts), by: 1. Verifying that the parameters in Table 4.8-1 meet the Category B limits, 2. Verifying that there is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10.6 ohms, and 3. Verifying that the average electrolyte temperature of every sixth connected cell is above 60'F.c. At least once per 18 months by verifying that the battery charger will supply at least 475 amperes at a minimum of 130 volts for at least 8 hours; and at least once each REFUELING INTERVAL by verifying that: 1. The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, 2. The cell-to-cell and terminal connections are clean, tight and coated with anti-corrosion material, and 3. The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10-6 ohms.d. At least once each REFUELING INTERVAL, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test.Once per 60 months, a modified performance discharge test may be performed in lieu of the battery service test.e. Verify battery capacity is > 80% of the manufacturer's rating when subjected to a performance discharge test or modified performance discharge test: i. At least once per 60 months, during shutdown, when the battery shows no signs of degradation, and has not reached 85% of service life.2. At least once per 12.months, during shutdown, when the battery shows signs of degradation, or has reached 85% of service life with < 100%of the manufacturer's rated capacity.3. At least once per 24 months, during shutdown, when the battery has.reached 85% of service life with > 100% of the manufacturer's rated capacity.DAVIS-BESSE, UNIT 1 3/4 8-9 Amendment No. 10,219, 229 TABLE 4.8-1 BATTERY SURVEILL ANCE REQUIREMENTS CATEGORY A... CATEGORY B" 2'Parameter Limits for each Limits for each Allowable(3) designated pilot connected cell value for each cell c6nnected cell Electrolyte
>Minimum level >Minimum level Above top of Level indication mark, indication mark, plates, and not and < k" above and  V" above overflowing maximum level maximum level indication mark(d) indication mark(d)Float >2.13 volts >2.13 volts(" >2.07 volts Voltage Not more than Specific >1.200(c)
>1.195 .020 below the Gravity(a) average of all connected cells Average of all Average of all connected cells connected cells>1.205 >1.195(c)(a) Corrected for electrolyte temperature and level. If the level is between the high and low marks and the temperature corrected specific gravity is within the manufacturer's nominal specific gravity range, it is not necessary to correct for level.(b) Corrected for average electrolyte temperature.(c) Or battery charging current, following a service, performance discharge, or modified performance discharge test, is less than two amps, when on a float charge.(d) It is acceptable for the electrolyte level to temporarily increase above the specified maximum during equalizing charges provided it is not overflowing.(I) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all parameter(s) are restored to within limits within the next 6 days.(2) For any Category B parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits within 7 days.(3) Any Category B parameter not within its allowable value indicates an inoperable battery.DAVIS-BESSE, UNIT I 3/4,;8-10 Amendment No. i..,-.8,229 3/4.8 ELECTRICAL POWER *SYSTEMS BASES Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.c.4 verify proper starting of the Emergency Diesel Generators from standby conditions.
Verification that an Emergency Diesel Generator has achieved a frequency of 60'Hz within the required time constraints meets the requirement for 'verifying the Emergency Diesel Generator has accelerated to 900 RPM.The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that 1)the facility can be maintained in the shutdown or refueling condition for extended time periods and 2) sufficient instrumentation and control' capability is available for monitoring and maintaining the facility status-.The Surveillance Requirements for demonstrating the OPERABILITY of the stati'on batteries are based on the recommendations of Regulatory Guide 1.129,"Maintenance, Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std. 450-1995, "IEEE Recommended Practice for Maintenance, Testing; and Replacement of Vented Lead-Acid Batteries for Stationary Applications," except that 'certain tests will be performed at least once each REFUELING INTERVAL.Battery degradation is indicated when the battery capacity drops more than 10%from its capacity on the previous performance discharge or modified performance discharge test, or is below 90% of the' manufacturer's rated capacity.Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that timewith the rated capacity.Table 4.8-1 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and spec'ific gravity.'The limits for the designated pilot cell's float voltage and specific gravity, greater than. 2.13 volts and .015' below the manufacturer's
'full charge specific gravity or a battery charger current of less than two amps is characteristic of a charged cell with adequate capacity.
The normal limits for'each connected cell for float voltage and specific gravity, greater than 2.13 volts and'not more than .020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than .010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery. Exceptions to the specific gravity requirements are taken to allow for the normal deviations experienced after a battery discharge and subsequent recharge associated with-a service, performance discharge, or modified performance discharge test. The specific gravity deviations-are recognized and discussed in IEEE- Std.. 450-1995.DAVIS-BESSE, UNIT 1 B 3/4 8-1a Amendment No. 100,158,203,29229 UNITED STATES 0NUCLEAR REGULATORY COMMISSION TWASHINGTON, D.C. 2D0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 229 TO FACILITY OPERATING LICENSE NO. NPF-3 FIRSTENERGY NUCLEAR OPERATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346
 
==1.0 INTRODUCTION==
 
On October 27, 1998, the Toledo Edison Company, Centerior Service Company, and The Cleveland Electric Illuminating Company (the licensees at the time of the submittal), submitted a request for changes to the Davis-Besse Nuclear Power Station, Unit No. 1, Technical Specifications (TSs). On January 1, 1999, FirstEnergy Nuclear Operating Company (FENOC) became the licensed operator of Davis-Besse.
The proposed amendment would revise TS 3/4.8.2.3, "Electrical Power Systems -DC Distribution
-Operating," and the associated bases. The surveillance requirements for battery testing would be changed.The DC electrical power systems at Davis-Besse are described in Updated Safety Analysis Report (USAR) Section 8.3.2, "DC Power Systems.".
As stated in the USAR, this equipment consists of two 250/125V DC motor control centers, four batteries, six battery chargers, four essential distribution panels, four 480V AC/125V DC rectifiers, and four nonessential distribution panels. These systems provide reliable power for control, instrumentation and DC loads required for normal operation and shutdown of the station.This amendment was submitted to.the NRC as committed to in the licensee's letter to the NRC dated October.16, 1997. The proposed changes are based on the guidance contained in the "Improved Standard Technical Specifications (ISTS) for Babcock and Wilcox (B&W) Plants," NUREG-1430, Revision 1, and the.recommended practices of Institute of Electrical and Electronics Engineers, Inc. (IEEE) Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications."  2.0 EVALUATION 2.1 Proposed Changie to TS Section 4.8.2.3.2.d The licensee proposed to' change TS Section 4.8.2.3.2.d which currently reads as follows: At least once each REFUELING INTERVAL, during shutdown, by'verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or. simulated emergency loads for the design duty cycle when the battery is subject to a battery service test.The proposed amended TS section would read as follows: At least once each REFUELING INTERVAL, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subject to a battery service test. Once per 60 months, a modified performance dischargie test may be performed in lieu of the battery service test, The licensee stated that TS SR 4.8.2.3.2.d currently requires performance of a battery service test at least onde each refueling interval during shutdown to verify that the battery capacity is adequate to supply all of the actual emergency loads for the design duty cycle. However, this test is not required to be performed every 60 months when a performance discharge test is performed as per the current surveillance requirement.
The proposed change will allow the option of performing a modified performance discharge test rather than the current option of performing a performance discharge test in place of the.battery service test once per 60 months.The modified performance discharge test, as defined in IEEE Std. 450-1995, is a test, in the as-found condition, of a battery's ability to provide a high-rate, short-duration load (usually the highest rate of the duty cycle) that will confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity.
As such, the modified performance-discharge test is a worst-case load profile of the traditional battery service test and a performance discharge test combined.On the basis of its review, the staff finds that this change is an improvement over the existing surveillance requirement and hence acceptable.
2.2 eProposed Change to TS Section 4.8.2.3.2.e The licensee proposed to change the Davis-Besse Nuclear Power Station, Unit 1, TS Section 4.8.2.3.2.e, which currently reads as follows:  At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60-month interval this performance discharge test may be performed in lieu of the battery service test.The proposedTS 4.8.2.3.2.e would read as follows: Verify that battery capacity is >80% of the manufacturer's rating when subjected to a performance discharge test or a modified performance discharge test: 1. At least once per 60 months, during shutdown.
when the battery shows no sign of degradation, and has not reached 85% of service life.2. At least once per 12 months, during shutdown, when the battery shows signs of degradation, or has reached 85% of service life with <100% of the manufacturer's rated capacity.3. At least once per 24 months, during shutdown, when the battery has reached 85% of service life with >100% of the manufacturer's rated capacity.The licensee stated that TS SR 4.8.2.3.2.e currently requires completion of a performance discharge test at least once per 60 months, during shutdown, to verify that the battery capacity is at least 80% of the manufacturer's rating.The proposed change will revise the current method of monitoring station battery capacity to allow the option of conducting either a performance discharge test or a modified performance discharge test. Additionally, the requirements of current TS SR 4.8.2.3.2.f are modified and combined with TS SR 4.8,2.3.2.e.
The proposed TS SR 4.8.2.3.2.e increases the frequency of performing a performance discharge test or modified performance discharge test to an annual or biennial frequency under certain specified battery conditions.
Additionally, in accordance with the recommended practices of IEEE Std. 450-1995, measurement of battery capacity degradation is proposed to be based on the last discharge test instead of on an average of the previous discharge tests, as is current practice.IEEE Std. 450-1995 states that annual performance tests of battery capacity should be made on any battery that shows signs of degradation or that has reached 85% of the service life expected for the application.
Degradation is indicated when the battery capacity drops'more than 10% from its capacity on the previous performance test,, or is below 90% of the manufacturer's rating.If the battery has reached 85% of its service life with a capacity of 100% or more of the manufacturer's rated capacity, and has no sign of degradation, performance testing at 2-year intervals is acceptable. The licensee stated that this proposed change to TS SR 4.8.2.3.2.e is a line item improvement that adopts the increased battery test frequency of the B&W ISTS and IEEE Std. 450-1995.
The proposed surveillance requirement change increases the frequency of battery testing fo provide increased monitoring of battery capacity once degradation due to age and use is noted, thereby increasing reliability of the battery to perform its safety function.On the basis of its review, the staff finds that this change is an improvement over the existing surveillance requirement and, hence, is acceptable.
2.3 Proposed Change to TS Section 4.8.2.3.2.f' The licensee proposed to change TS Section 4.8.2.3.2.f, which currently reads as follows: Every REFUELING INTERVAL, during shutdown, performance discharge tests of battery capacity shall be given to any battery that shows.signs of degradation or has reached 85% of the service life expected for-the application.
Degradation is indicated when the battery capacity'drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.The licensee proposes to delete this surveillance requirement since its provisions have been modified to incorporate the guidance provided by the B&W ISTS and IEEE Std. 450-1995, and have been included in the proposed modified SR 4.8.2.3.2.e.
On the basis of its review, the staff agrees with the licensee that the preceding requirements have been modified and included in the proposed modified SR 4.8.2.3.2.e and, hence, this surveillance requirement deletion is acceptable.
2.4 Proposed Change to TS Table 4.8-1 The licensee proposes to change TS Table 4.8-1., The current table is shown on the next page, and the proposed table is shown on the following page.The licensee stated that TS Table 4.8-1 currently requires battery electrolyte specific gravity measurement correction for temperature and level.Temperature and level correction is performed to permit trending of specific gravity to ensure adequately charged battery cells with adequate capacity.The cell's specific gravity is based on a temperature of 77 degrees E and a specified electrolyte level. The proposed change to TS Table 4.8-1 footnote (a) will not require level correction of the electrolyte specific gravity measurement provided that the electrolyte level is within the specified band and the temperature corrected specific gravity is within the specified range.This proposed change adopts the guidance of IEEE Std. 450-1995.  -[CURRENT]
TABLE 4.8-1 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(l)CATEGORY B(2)Limits for each Limits for each Allowable(3) designated pilot connected cell value for each cell I connected cell Electrolyte Level>Minimum level indication mark, and " above maximum level indication mark>Minimum level indication mark, and < 1/4" above maximum level indication mark Above top of plates, andnot overflowing Float >2.13 volts >2.13 volts(b) >2.07 volts Voltage Not more than Specific >1.200(c)
>1.195 .020 below the Gravity(a) average of all connected cells Average of all Average of all connected cells connected cells>1.205 1.195(c)(a)(b)(c)(1)(2)(3)Corrected for electrolyte temperature and level.Corrected for average electrolyte temperature.
Or battery charging current, following a service or performance discharge test, is less than two amps, when on a float charge.For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all parameter(s) are restored to within limits within the next 6 days.For any Category B parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits within 7 days.Any Category B parameter not within its allowable value indicates an inoperable battery. rPROPOSED1 TABLE 4.8-1 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(1) CATEGORY B(2)Parameter Limits for each Limits for each Allowable(3) designated pilot connected cell value for each cell connected cell Electrolyte
>Minimum level >Mininium level Above top of Level indication mark, indication mark, plates, and not and  k" above and < V" above overflowing maximum level maximum level indication mark{d) indication markd_Float >2.13 volts >2.13 volts (b) >2.07 volts Voltage Not more than Specific >1.200(c)
&#xfd;1.195 .020 below'the Gravity(a)-
average of all connected cells Average of all Average of all connected cells connected cells>1.205 >1.195(c)a) Corrected for electrolyte temperature and level. If the level is between the high and low marks and the temperature corrected sDecific aravitv is within the manufacturer's nominal specific gravity range, it is not necessary to correct for level.(b) Corrected for average electrolyte temperature.(c) Or battery charging current, following a service performance discharge, or modified performance discharge test, is less than two amps, when on a float'charge.(d) It is acceptable for the electrolyte level to temporarily increase above the specified maximum during equalizing charoes provided it is not overflowino.
(1) For any Category A parameter(s) outside the limit(s) shown, the.battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all parameter(s) are restored to within limits within the next 6 days.(2) For any Category B parameter(s) outside the limit(s) shown, the battery may'be considered OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits within 7 days.(3) Any Category B parameter not within its allowable value indicates an inoperable battery. Footnote (c) will be modified in accordance with the proposed modifications to TS SR 4.8.2.3.2.e, which provides for modified performance discharge testing.The proposed addition to TS Table 4.8-1 of the new footnote (d) will allow the electrolyte level to temporarily increase above the specified maximum during equalizing charges provided that it is not overflowing.
The electrolyte level limits ensure that the plates suffer no physical damage, that adequate electron transfer capability is maintained, and that electrolyte does not overflow causing damage to the battery, battery connections, cell covers, or battery racks. This proposed change adopts the guidance of the B&W ISTS and IEEE Std. 450-1995.On the basis of its review, the staff finds that this change is consistent with B&W ISTS and IEEE Std. 450-1995 and, hence, is acceptable.
2.5 Proposed Change to TS Bases 3/4.8 The licensee proposed modifications to the existing TS Bases 3/4.8 to provide a basis for each surveillance requirement.
The surveillance requirements for demonstrating the OPERABILITY of the station batteries are based on the recommendations of Regulatory Guide 1.129, "Maintenance, Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants" (February 1978), and IEEE Std. 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," except that certain tests will be performed at least once each REFUELING INTERVAL.
Battery degradation is indicated when the battery capacity drops more than 10% from its capacity on the previous performance discharge or modified performance discharge test, or is below 90%of the manufacturer's rated capacity.
The bases will also be modified to incorporate the commitment to IEEE Std. 450-1995 for battery cell electrolyte level, float voltage and specific gravity and to reflect new provisions for modified performance discharge testing.On the basis of its review, the staff finds that the licensee has revised the bases to reflect the proposed changes to TS SR 4.8.2.3.2.d, 4.8.2.3.2.e, and 4.8.2.3.2.f, and TS Table 4.8-1, and finds this change acceptable.
 
==3.0 STATE CONSULTATION==
 
In accordance with the Commission's regulations, the Ohio State official was notified of the proposed issuance of the amendment.
The State official had no comments.4.0 ENVIRONMENTAL CONSIDERATION This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes a surveillance requirement.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards  consideration, and there has been no public comment on such finding (63 FR 64125). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
 
==5.0 CONCLUSION==
 
The staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance-of this amendment will not be inimical to the common defense and security or to the health and safety of the public.Principal Contributor:
A. Pal Date: February 9, 1999 NRC ITS Tracking Page I of 2 Assignil 100Return to View Menu Print Doc en FAI Screening Required:
Yes This Document will be approved by: Tim Kobetz This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC Reviewer ID 200711161516 Conference Call Requested?
No.a~tegory~
In Scope ITSSection:
TB-POC: JFD Number: Page Numlber(s):
ITS 5.0 Gerald Weig Gerald Waig 23 112 Information ITS Number: OSI: DOC Number.: Bases JFD Number: 5.5 None None None NRC Autho] Gerald Waig Question: Please identify the reference document and explain why ITS 5.5.15.d.1 identifies the containment penetration and valve Type B and C tests leakage rate acceptance criteria as less than or equal to 0.60 La. This value is contrary to the value of less than 0.60 La identified in the Davis Besse CTS, 10 CFR 50 App. J, and ISTS.Discussion:
Comment JFD #23 provides an explanation for the containment penetration and valve Type B and C tests leakage rate acceptance criteria deviation from ISTS and states, in part, that the value is "...consistent with current licensing basis." The reviewer was unable to locate, in the current licensing basis, the ITS value (less than or equal to 0.60 La) acceptance criteria for the Type B and C containment penetration and valve tests to support the deviation.
 
==Reference:==
 
10 CFR 50 App. J I ssu&#xfd;e Date I 11/16/2007 Close Date [01/11/2008 Logged in User: Jerry Jones'Responses
].1 Licensee Response by Jerry Jones on 12/05/2007 ITS 5.5.15.d.1 (Volume 16, Page 105) incorrectly changed the Types B and C acceptance criteria value. The current licensing http://www.excelservices.com/exceldbs/itstrack-davisbesse.nsf/1fddcea1Od3bdbb585256e8...
-6/2/2008 NRC ITS Tracking Page 2 of 2 basis value is less than 0.60 La, as shown in CTS 6.16.d. 1) (Page, 60), not less than or equal to 0.60 La. A draft markup regarding this change is attached.
This change will be reflected in the supplement to this section of the ITS Conversion Amendment.
Date Created: 11/16/2007 03:16 PM by Gerald Waig Last Modified:
01/11/2008 04:03 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/l fddcea lOd3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 1 of 2 FK ;A',&#xfd;eReturn to View Menu] Prnt Docunment RAI Screening Required:
Yes This Document will be approved by: Gerald Waig; Carl Schulten; Tim Kobetz This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC ITS TRACKING N1RC Roriowwr ID 200801111544 Conference Call Requested?
No Cater ESI -Emergent Staff Issue ITSS.ec.tion:.
TB POC: JFD Number.:.
PageNumber(s):.
5.0 Gerald Robert Clark None 74 ITS Weig/Ravinder OSI: DOC.. Number: Bases.JFD.Number:
Information Grover None M.2 None ITS. Numb!.er.:
5.5 INRC A o Robert Clark Regulatory Bases: 10CFR50.36(c)(3) requires Surveillance Requirements to verify that the LCO's are met.Required Action: The current TS require that the stored fuel oil properties for water and Comment sediment content be verified to be within limits every 92 days. This surveillance requirement is used to ensure that the quality of the stored fuel oil is satisfactory for long term operation of the EDGs. However, the proposed change to ITS Section 5.5.12, Diesel Fuel Oil Testing Program, does not retain this surveillance requirement for stored fuel oil. Please provide justification for not including this surveillance requirement in the ITS.Issue Date 01/11/2008 Close Date 02/19/2008 Logged in User: Jerry Jones-'Responses Licensee Response by Bryan Kays on 01/27/2008 ITS 5.5.12 (Volume 16, Pages 100 and 101) specifically details the Diesel Fuel Oil Testing Program. Discussion of Change (DOC)M02 (Page 74) was written to add the requirements for testing of new fuel prior to addition to the fuel oil storage tank. DOC LO 1 http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/1 fddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page 2 of 2 (Page 78) must be read in addition to DOC M02. DOC LO1 was written for the deletion of the quarterly viscosity, water, and sediment checks of stored fuel oil. DOC LO 1 states that these changes are acceptable because DOC M02 provides an acceptable level of equipment reliability.
It also states that fuel degradation during long term storage shows up as an increase in particulate and that the total particulate is determined and compared every 31 days, as required by ITS 5.5.12.c.
Additionally, DOC LO1 states that ITS SR 3.8.3.5 has been added by ITS 3.8.3 DOC M04 (Volume 13, Pages 138 and 139) to ensure that microbiological fouling does not occur. ITS 3.8.3.5 is performed every 31 days.Based on the changes described, the quality of stored fuel is satisfactory for long term operation of the EDGs and the requirement for not performing water and sediments content, of stored fuel oil, on a 92 day frequency is acceptable.
Additionally, the proposed ITS is consistent with the ISTS, i.e., Davis-Besse deleted the current CTS 92 day surveillance and added the new ISTS 31 day surveillance.
Furthermore, Davis-Besse completely
,,adopted the ISTS Diesel Fuel Oil Program.NRC Response by Robert Clark No further questions at this time. Item closed.on 02/19/2008
_.Date Created: 01/11/2008 03:44 PM by Robert Clark Last Modified:
02/19/2008 05:05 PM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/lfddcea 1Od3bdbb585256e8...
6/2/2008 NRC ITS Tracking Page I of 2 Return to View Menuj Prit~ounn RAI Screening Required:
Yes This Document will be approved by: Carl Schulten This document has been reviewed and information in this question contains NO SUNSI sensitive material (the checkbox to the right must be selected before this question can be submitted)
NRC ITS TRACKING Status: Closed Regulatory Basis must be included in Comments section of this Form Yes NRC Re~viewer ID 200806260901 Conference Call Requested'?
NoOther Technical Challenge ITS Section: TB POC: JFD Number: Page Number(s):
5.0 Gerald None ITS Weig/Ravinder OS:. DOC Number: Bases Nunmber.:.
Information Grover None None None IT.S.-Number:
5.5 On 6/26/08 licensee requested:
Com "Please start a question thread related to questions that have come up related............
to the conversion from CTS to ITS 5.5.11." This thread is posted as requested.
IssueDate 1] 06/26/2008 Close Date [07/02/2008 Logged in User: Anonymous-'Responses Licensee Response by Jerry Jones on 06/26/2008 Davis-Besse has recently discovered that there are two differences between CTS 3.11.1 and ITS 5.5.1 .b that are not discussed in either the Discussion of Changes or the Justification for Deviations.
First difference:
Note
* to CTS 3.11.1 essentially defines unprotected outdoor tanks as those tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents "or" that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. That is, if either of the options is provided for a tank, the CTS 3.11.1 requirements are not applicable to the tank, since it is not unprotected.
ITS 5.5.11 includes similar requires on unprotected outdoor tanks, but changes the "or" in the CTS Note
* description of unprotected to an "and." That is, for a tank to be considered as http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/lfddcealOd3bdbb585256e...
7/17/2008 NRC ITS Tracking Page 2 of 2 protected it must be surrounded by liners, dikes, or walls capable of holding the tank contents "and" it must have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. This change from the CTS to the ITS has not been described.
A new Discussion of Change (DOC M04) is being provided to justify this more restrictive change. Second difference:
The Davis-Besse CTS 3.11.1 requirements specify that "each" unprotected outdoor tank is allowed to contain the LCO limit of radioactivity.
ITS 5.5.1 .b can be interpreted as requiring the total amount of radioactivity in all the unprotected outdoor tanks to be within the same limit. Essentially this means that the assumption is that all the unprotected outdoor tanks rupture simultaneously.
Davis-Besse believes the intent is that each tank is allowed to contain the radioactivity limit; the assumption is that only one tank ruptures.
Therefore, the ITS is being changed to require "each" tank to be within the limit. This is consistent with NUREG-0472, Revision 3, which is the version of the Radiological Effluents Technical Specifications for PWRs to which Davis-Besse is licensed, as documented in the NRC Safety Evaluation for Amendment 86, dated July 2, 1985. A draft markup regarding these changes is attached.
These changes will be reflected in the ,,supplement to this section of the ITS Conversion Amendment.
NRC Response by Aron Lewin No further questions at this time.on 07/02/2008
[Date Created: 06/26/2008 09:01 AM by Aron Lewin Last Modified:
07/02/2008 09:39 AM http://www.excelservices.com/exceldbs/itstrackdavisbesse.nsf/
1 fddceal Od3bdbb585256e...
7/1'7/2008 Attachment 2 L-08-241 Response to Request for Additional Information Related to Update to B&W Plants Standard Technical Specifications NUREG-1430 Page 1 of 3 To complete their review, the NRC staff has requested additional information regarding the license amendment application for the conversion of Current Technical Specifications to Improved Technical Specifications.
The staff request is provided below in bold type, followed by a reference to the associated website question number where the responses are documented in the section titled "Section 3.3 RAIs" in Attachment 1: Beyond Scope Item -1 "EDG Degraded Voltage & Loss of Voltage Relay Channel Check" 1. The proposed TS change to limiting condition for operation (LCO) 3.3.8 is to delete every 12-hour channel check surveillance requirement for the loss of voltage and degraded voltage emergency diesel generator (EDG) loss of power sequencing (LOPS) instrumentation.
These safety functions are performed by voltage relays. The relays can be electro-mechanical or solid state design.To complete its review of the proposed changes to the TSs, the NRC staff needs to review the details of these relays and, therefore, requests the licensee to provide model and make information, and instruction leaflet references for each of these two types of relays.FENOC's response is provided in the response to conversion website question 200712261047 in Section 3.3 RAIs in Attachment 1.Beyond Scope Item -2 "Steam and Feedwater Rupture Control System Instrumentation Trip Setpoints" (LCO 3.3.11)The proposed change revises the current TS Table 3.3-12 trip setpoint allowable values (AV) for Steam Line pressure -Low, Steam Generator Level -Low, and Steam Generator Feedwater Differential Pressure -'High, functional units instrumentation.
To determine the acceptability of the proposed TSs change involving revision of instrumentation setpoints, the NRC staff generically requests all licensees who propose revision of instrumentation setpoint and/or setpoint AVs (conservative or non-conservative), to provide the following information (I&C Branch Guideline for Setpoint -Related TSs License Amendment Request -Agencywide Documents Access and Management System (ADAMS) Accession No.ML061810132):
Attachment 2 L-08-241 Page 2 of 3 2. Provide documentation (including sample calculations) used for establishing the limiting setpoint (or NSP) and the limiting acceptable values for the As-Found and As-Left setpoints as measured in periodic surveillance testing. Indicate the related analytical limits and other limiting design values (and the sources of these values) for each setpoint.FENOC's response is provided in the response to conversion website question 200712261240 in Section 3.3 RAIs in Attachment 1.3. Provide a statement as to whether or not these three setpoints are limiting safety system settings (LSSSs) for the variables on which a safety limit (SL) has been placed as discussed in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(c)(1)(ii)(A).
These variables provide protection against violating reactor core safety limits, or reactor coolant system pressure boundary safety limits. For each setpoint that you determined not to be SL-related, explain the basis for this determination.
FENOC's response is provided in the response to conversion website question 200712261243 in Section 3.3 RAIs in Attachment 1.4. For setpoints that are determined to be SL-related, the NRC letter to the Nuclear Energy Institute Setpoint Methods Task Force (SMTF) dated September 7, 2005 (ADAMS Accession No. ML 052500004), describes-Setpoint-Related TS (SRTS) that are acceptable to the NRC for instrument settings associated with SL-related stepoints.
Specifically, part "A" of the enclosure to the letter provides LCO notes to be added to the TS, and part"B" includes a check list of the information to be provided in the TS Bases for the proposed TS changes. In this regard, please address items a, b, and c of this section.a. Describe whether and how you plan to implement the SRTS suggested in the September 7, 2005, letter. If you do not plan to adopt the suggested SRTS, then explain how you will ensure compliance with 10 CFR 50.36 by addressing items b and c below.b. Describe how surveillance test results and the associated TS limits are used to establish operability of the safety system. Show that the As-Found Setpoint valuation is consistent with the assumptions and results of the setpoint calculation methodology.
Attachment 2 L-08-241 Page 3 of 3 c. Describe the controls employed to ensure that the instrument setpoint is, upon completion of surveillance testing, consistent with the assumptions of the associated analyses.
If the controls document is other than the TS (e.g. plant test procedure), explain how the requirements of 10 CFR 50.36 are met.FENOC's response is provided in the response to conversion website question 200712261254 in Section 3.3 RAIs in Attachment 1.5. For setpoints that are not determined to be SL-related, describe the measures to be taken to ensure that the associated instrument channel is capable of performing its specified safety functions, in accordance with applicable design requirements and associated analyses.
Include in your discussion the information on controls you employ to ensure that the as-left trip setpoint setting after completion of periodic surveillance is consistent with your setpoint methodology.
FENOC's response is provided in the response to conversion website question 200712261257 in Section 3.3 RAIs in Attachment 1.}}

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