View source for JAFP-10-0072, Response to Request for Additional Information Application for Amendment to Modify the Technical Specifications Requirements for Testing of the Shutdown Cooling System

{{Adams
| number = ML101680559
| issue date = 06/17/2010
| title = Response to Request for Additional Information Application for Amendment to Modify the Technical Specifications Requirements for Testing of the Shutdown Cooling System
| author name = Dietrich P
| author affiliation = Entergy Nuclear Northeast, Entergy Nuclear Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000333
| license number = DPR-059
| contact person = 
| case reference number = JAFP-10-0072, TAC ME1819
| document type = Letter type:JAFP
| page count = 24
| project = TAC:ME1819
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:* Entergy
.=::-.=.                                                           Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 Pete Dietrich Site Vice President - JAF JAFP-10-0072 June 17, 2010 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

==SUBJECT:==
Response to Request for Additional Information Re: James A. FitzPatrick Nuclear Power Plant Application for Amendment to Modify the Technical Specifications Requirements for Testing of the Shutdown Cooling System Isolation, Reactor Pressure - High Function (TAC No. ME1819)
James A. FitzPatrick Nuclear Power Plant Docket No.      50-333 License No. DPR-59

==References:==
: 1. Entergy Letter, JAFP-09-0086, Application for Amendment to Modify the Technical Specifications Requirements for Testing of the Shutdown Cooling System Isolation, Reactor Pressure - High Function (TAC No. ME1819),
dated July 31, 2009
: 2. NRC Request For Additional Information Regarding James A. FitzPatrick Nuclear Power Plant Application for Amendment to Modify the Technical Specifications Requirements for Testing of the Shutdown Cooling System Isolation, Reactor Pressure - High Function (TAC No. ME1819), dated February 2, 2010
: 3. Entergy Letter, JAFP-1 0-0033, Response to Request for Additional Information Re: James A. FitzPatrick Nuclear Power Plant Application for Amendment to Modify the Technical Specifications Requirements for Testing of the Shutdown Cooling System Isolation, Reactor Pressure - High Function (TAC No. ME1819), dated March 5, 2010
: 4. NRC Request For Additional Information Regarding James A. FitzPatrick Nuclear Power Plant Application for Amendment to Modify the Technical Specifications Requirements for Testing of the Shutdown Cooling System Isolation, Reactor Pressure - High Function (TAC No. ME1819), dated April 13,2010

==Dear Sir or Madam:==

On July 31,2009, Entergy Nuclear Operations, Inc. (ENO) submitted an application for amendment to the Technical Specifications (TS) for the James A. FitzPatrick Nuclear Power

JAFP-10-0072 Page 2 of 3 Plant (JAF), to revise the surveillance testing requirements for the Shutdown Cooling System Isolation, Reactor High Pressure Function (Reference 1). On February 2, 2010, JAF received a request for additional information from the Nuclear Regulatory Commission (NRC) staff (Reference 2). The request was subsequently clarified in a conference call with the staff on February 17, 2010. Based on the clarifying discussions with the staff, ENG provided the responses documented in Reference 3. After revieWing Reference 3, the NRC staff advised ENO of additional questions. The additional questions were discussed in a teleconference on AprilS, 2010, with written questions being provided on April 13, 2010 (Reference 4).
Based on the discussions with the staff, ENO is providing this response to the request for additional information. Attachment 1 provides a response to each RAI question.
The attached response does not affect the No Significant Hazards Determination submitted with the amendment application, dated July 31,2010.
There are no new commitments made in this letter.
Questions concerning this report may be addressed to Mr. Joseph Pechacek, Licensing Manager, at (315) 349-6766.
I declare under penalty of perjury that the foregoing is true and correct.
Pete Dietrich Site Vice President - JAF PD/JP/ed Attachments: 1. Responses to Request for Additional Information Questions
: 2. Source Document Excerpts cc: next page

JAFP-10-0072 Page 3 of 3 Mr. Samuel Collins                         Resident Inspector's Office Regional Administrator, Region I           U.S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission        James A. FitzPatrick Nuclear Power Plant 475 Allendale Road                         P.O. Box 136 King of Prussia, PA 19406-1415             Lycoming, NY 13093 Mr. Bhalchandra Vaidya, Project Manager    Mr. Paul Eddy Plant Licensing Branch 1-1                 New York State Department of Public Service Division of Operating Reactor Licensing    3 Empire State Plaza, 10th Floor Office of Nuclear Reactor Regulation       Albany, NY 12223 U.S. Nuclear Regulatory Commission Mail Stop 0-8-C2A Washington, DC 20555-0001 Mr. Francis J. Murray Jr., President New York State Energy and Research Development Authority 17 Columbia Circle Albany, NY 12203-6399 Document Component(s):
001 Transmittal letter JAFP-1 0-0072 with Attachments

JAFP-10-0072 Attachment 1 Responses to Request for Additional Information Questions (7 Pages)

JAFP-10-0072 Attachment 1 Responses to Request for Additional Information Questions Question 1:
"In Attachment 1, on page 3, of the licensee's RAI response letter dated March 5, 2010 (JAFP-10-0033), it states in the first paragraph:
    "Following performance ... to review the completed procedure to ensure acceptance criteria was satisfied and that "As-Found" and "As-left" values are within their required tolerance."
And in the second paragraph, it states:
    "If a Condition Report documents that an instrument failed to meet Level 1 acceptance criteria and could not be restored to within tolerance, the instrument would be declared INOPERABLE and the appropriate Condition associated with the applicable Limiting Condition for Operation (LCO) would be entered."
: a. Clarify the use of the terms and provide the values of the "acceptance criteria", the "required tolerances", and "Level 1 acceptance criteria."
: b. Provide the detailed calculation and/or determination of "As-Found" and "As-Left" setting tolerances of the entire instrument loop."

===Response===
1.a       In accordance with our Surveillance Testing Program procedure AP-19.01 , acceptance criteria is defined as "The measure against which Surveillance Test Procedure (STP) performance results are evaluated to determine if SRs [Surveillance Requirements] are met." In paragraph one of the RAI response letter dated March 5, 2010 (JAFP           0033), the use of the term "acceptance criteria" is in reference to the Technical Specification Allowable Values (TSAVs). The surveillance review, discussed in the RAI response, compares the calibration results to the TSAVs. The acceptance criteria that would be used for this comparison would be the TSAV for the Shutdown Cooling System High Reactor pressure isolation function which is ~ 74 psig.
In paragraph two of the RAI response letter dated March 5, 2010 (JAFP-10-0033), the use of the term "Level 1 acceptance criteria" is in reference to the Allowable Values (AV) as provided in our Technical Specifications (TS). This term is defined in our Surveillance Testing Program procedure AP-19.01 as "The measure that defines characteristics of a system or component that, if not met, result in a violation of TS
[Technical Specifications], TRM [Technical Requirements Manual], ODCM [Offsite Dose Calculation Manual], or plant safety design bases as stated in the UFSAR [Updated Final Safety Analysis Report] or EN-DC-167." [Detail Added] The TSAV for the Shutdown Cooling System High Reactor pressure isolation function is ~ 74 psig.
Therefore, this value is considered a Level 1 acceptance criterion.
The use of the term "required tolerances" is in reference to the criteria provided in the procedures for the calibration of the equipment. The required tolerances are referred to as "As-Found Zones" and "As-Left Tolerances". The 4-20mA transmitter output represents a 1200 psig range. The instrument calibration criteria or required tolerances for the Slave Trip Units (STUs), [02-3STU-251A and 02-3STU-251 D] are +/- 3.39 psig
(+/-0.04 mA, As-Found Zone) and +/- 2.4 psig (+/- 0.03 mA, As-Left Tolerance).
Page 1 of 7

JAFP-l0-0072 Attachment 1 Responses to Request for Additional Information Questions 1.b The detailed calculation was previously provided with the RAI response letter dated March 5, 2010 [Entergy calculation number JAF-CALC-NBS-02052]. The shutdown cooling function is not required to mitigate the consequences of an accident and thus the instrument errors due to harsh environmental effects including process measurement and Insulation Resistance Effect (IRE) do not need to be considered. Per the setpoint calculation methodology presented in ENN-IC-G-003, the As-Found setting tolerance (AFZ) is set the same as the Allowable Value to Trip Setpoint Margin (AVTSM) and the values are scaled as necessary for limitations of test equipment (refer to Section 7.1 in JAF-CALC-NBS-02052 for scaling). The AVTSM is specified for each individual instrument in sections 6.5.1, 6.5.2, and 6.5.3 of JAF-CALC-NBS-02052 as follows: Transmitter AVTSM (AFZ) = +/-13.83 psig, Master Trip Unit (MTU) AVTSM (AFZ)
    =+/-3.53 psig and STU AVTSM (AFZ) =+/-3.39 psig.
The combination of these individual instrument As-Found setting tolerances equate to the As-Found setting tolerance for the entire instrument loop. This value is presented in section 6.5.4 and is specified as +/-14.67 psig (AVTSMchannel).
The As-left Tolerance (ALT) for each individual instrument is provided in JAF-CALC-NBS-02052 sections 6.2.1.10, 6.2.2.10, and 6.2.3.10 as follows: Transmitter ALT =
    +/-3.00 psig, MTU ALT =+/-2.25 psig, and STU ALT =+/-2.4 psig.
Due to the manner in which the instruments are calibrated, the ALT for the entire instrument loop is not necessary and thus is not computed in the setpoint calculation JAF-CALC-NBS-02052. If it was necessary, it would be calculated by combining the ALT of each individual instrument using a Square Root Sum of the Squares method since each of the ALT errors are random in nature. For your reviewing convenience, the ALT for the entire instrument loop is presented below:
ALT Loop =+/- ~ PT  ALT 2
                            + MTU Al.T 2 + STU ALT 2 2       2       2 ALT Loop = +/- .J3.00 + 2.25 + 2.4 (all units are psig)
ALT Loop = +/- 4.45 psig Page 2 of 7

JAFP-10-0072 Attachment 1 Responses to Request for Additional Information Questions Question 2:
  "In attachment 2, section 8.2.3, of the licensee's RAI response, it states:
     "Use of the reactor steam dome pressure measurement (ATTS) to provide the SDC isolation function will result in the utilization of an instrument loop that has a larger instrument uncertainty than originally provided by the pressure switches."
The licensee calculated the allowable value (Section 6.5, Attachment 2 of the RAI response) by applying the "method 3" of ISA-RP 67.04.02-2000. The licensee also calculated the limiting trip setpoint LTS = EAL - CU (section 6.4) with a round up margin of 0.72 psig (section 6.3.1).
However, the round up of AVTSMchannel from +/- 14.67 psig to +/- 15 psig increases the allowance for instrument channel deviation (section 6.3.1); thus is less conservative and cannot be counted as an addition margin.
Explain how you can ensure that the field trip setpoint is conservative with respect to the analytical value (AL) with "method 3" and with such a small margin in the LTS calculation."

===Response===
As defined in our instrument loop accuracy and setpoint methodology procedure ENN-IC-G-003, the "Allowable Value to Trip Setpoint Margin" (AVTSM) is the difference in the parameter of interest between the Allowable Value (AV) and the Trip Setpoint. At a minimum, the AVTSM consists of all channel As-Found Tolerance (AFT) uncertainties. The AVTSM can be all uncertainties representing normal operation conditions that will be experienced between successive calibrations."
Even though increasing the AVTSMchannel from +/- 14.67 psig to +/- 15 psig increases the allowance for instrument channel deviation, the direction of conservatism is based on perspective. By increasing this allowed channel deviation, it is less likely that the TSAV would be exceeded.
This is because the resultant the field trip setpoint would be set to a lower value. This is considered conservative since the SDC isolation function occurs on an increasing pressure and a lower setpoint would actuate earlier. The potential for exceeding the TSAV is further minimized by the application of the Surveillance Testing Program, Corrective Action Program, and instrument calibration procedures. The details of these processes were discussed in the RAI response letter dated March 5, 2010 (JAFP-1O-0033). These processes are in place to monitor and identify trends of instrument behavior relative to the TSAVs. Therefore, the combination of procedures, the TSAV, and the additional allowance for instrument channel deviation between the field trip setpoint and the AV will adequately protect the Analytical Limit (AL).
Increasing the AVTSM is only acceptable as long as the AL is adequately protected. As requested, the margin in the field trip setpoint calculation and assurance of a conservative field trip setpoint with respect to the AL using the ISA-RP 67.04 "method 3" are evaluated below.
Calculation JAF-CALC-NBS-02052 determines both a total loop uncertainty (CU STU in Section 6.3.1) and the Allowable Value to Trip Setpoint Margin (AVTSMchannel in Section 6.5.4). Per our setpoint calculation methodology, one main difference between the total loop uncertainty and Page 3 of 7

JAFP-10-0072 Attachment 1 Responses to Request for Additional Information Questions the AVTSM is the instrument errors associated with harsh environments or abnormal conditions. The difference between the calculated total loop uncertainty of +/-15.28 psig and the calculated AVTSMchannel of +/-14.67 psig is 0.58 psig. Considering this and the difference of 1 psig between the AL of ~75 psig and the TSAV of ~74 psig, there is additional margin of 1.0 psig - 0.58 psig = 0.42 psig between the AL and the AV. Though this may not be considered a large amount of margin, it is still margin. The amount of applied margin is not a specified value; thus, any amount of margin adds conservatism. From an operational perspective, application of too much margin can have negative impacts such as pushing the setpoint into the normal operating region of the process.
Based on the preceding information, the allowed instrument channel deviation has been increased to help ensure the TSAV is not exceeded. In addition, Entergy has procedures in place to monitor and identify trends of instrument behavior relative to the TSAVs. Also, a margin of 0.42 psig is provided between the Shutdown Cooling System High Reactor pressure isolation function AV and the associated AL. Furthermore, the Rosemount Analog Transmitter Trip System (ATIS) that will be used for the Shutdown Cooling System High Reactor pressure isolation function has a proven history of reliability and accuracy at JAF. Therefore, the nominal field trip setpoint of 59 psig is considered conservative and will adequately protect the AL.
Page 4 of 7

JAFP-1 0-0072 Attachment 1 Responses to Request for Additional Information Questions Question 3:
In Attachment 2, Table 1, of the licensee's RAI response, it states that the humidity, radiation, and temperature effects for the total loop uncertainty calculation of pressure transmitters and ATTS under LOCA and HELB conditions are same as those under normal condition. This statement is based on assumption 1.2.6. Assumption 1.2.6 states that "OBO-01 0 (Ref 4.2.11) identifies the SOC Isolation function as not having any accident mitigating functions." However, according to TS Table 3.3.6.1-1 (page 5 of 6), Function 6a "SOC Isolation, Reactor Pressure -
Hi" is used to initiate a primary containment isolation function. Assumption 1.2.6 seems to contradict with this TS requirement. Please provide the following:
: a. Where the pressure transmitter 02-3PT-55A/O is located (inside or outside the primary containment)?
: b. What are the worst-case environmental conditions at the locations for the transmitters and ATTS?
: c. What are the worst-case environmental conditions under which the SOC isolation valves are required to be actuated?
: d. Provide a clarifying description regarding the conditions under which the function of this instrument channel is required when using the SOC line that resolves the apparent contradiction between assumption 1.2.6 and the TS requirement of Function 6a in the TS Table 3.3.6.1-1 (page 5 of 6).
: e. Provide copies of the source documents that justify the values for seismic, radiation, humidity, and temperature values used in the calculation, JAF-CALC-NBS-02052.

===Response===
3.a      The pressure transmitters, 02-3PT-55A and 02-3PT-550, are located outside of primary containment on the 300 foot elevation of the reactor building.
3.b      The worst case environmental conditions for the transmitters 02-3PT-55A, B (RB EL.
5 EO Node RB300Switch) are 100% relative humidity, 3.19 x 10 5 Gamma and 3.04 X 10 Beta and 215&deg;F.
The worst case environmental conditions for the transmitters [02-3PT-55C and 02-3PT-550] (RB EL. EO Node RB300North) are 100% relative humidity, 1.07 x 105 Gamma 5
and 3.04 X 10 Beta and 158.38&deg;F.
The worst case environmental conditions (Ref.OBO-70) for the ATTS (RR EL. 284) are 2
50% relative humidity, 1.75 x 10 R TIO (40 years) and 104&deg;F.
As discussed in section 1.2.6 of JAF-CALC-NBS-02052, the Shutdown Cooling System High Reactor pressure isolation function is not credited in any accident analysis and is not required to mitigate the consequences of an accident. Therefore, the worst case environmental conditions that the transmitters 02-3PT-55A and 02-3PT-550 will be exposed to when the instruments are required to operate is the same as normal Page 5 of 7

JAFP-10-0072 Attachment 1 Responses to Request for Additional Information Questions environmental conditions. For further information, see response to 3.c and 3.d below.
The MTUs and STUs that make up the ATTS are physically located in the Relay Room.
This room is below the Control Room and has an HVAC system that controls the environmental conditions. Therefore, the worst case environmental conditions that the MTUs and STUs will be exposed to when the instruments are required to operate are the same as normal environmental conditions. The normal environmental conditions are presented in Section 3.2 Table 1 of JAF-CALC-NBS-02052.
3.c To satisfy the TS requirement Function 6a in TS Table 3.3.6.1-1, the SOC isolation valves, 10MOV-17 and 10MOV-18, must close when a reactor high pressure condition exists. When this condition occurs, if the valves are not already closed, they will automatically close with a trip of the SOC System High Reactor pressure isolation logic.
Since the plant is normally operating and a reactor high pressure condition exists, these valves are normally closed.
When plant power/pressure is reduced, the SOC isolation valves, 10MOV-17 and 10MOV-18, can only be opened when: 1) the Shutdown Cooling System High Reactor pressure isolation logic is reset; and 2) the Plant Operator manually initiates SOC by opening the valves. The logic of the circuit is designed such that a low reactor pressure condition (plant must be shutdown) is required in order for the isolation logic to be reset.
Therefore, the plant condition that would exist when the SOC isolation valves are required to actuate, per Function 6a of TS Table 3.3.6.1-1 (from open to closed), is plant heatup from a shutdown condition. If the plant is in a state of heatup, normal environmental conditions exist. Thus the worst case environmental conditions under which the SOC isolation valves are required to actuate (from an open to a closed state) are normal environmental conditions.
3.d The conditions under which the function of this instrument channel is required will be discussed in reference to reactor low and high pressure conditions:
Low Reactor pressure - Plant shutdown:
When reactor pressure is equal to or less than the SOC logic reset value, a reactor low pressure logic condition is generated which interfaces with two separate logic circuits via relay contacts. This will allow two functions to occur: 1) the manual opening of the SOC isolation valves, 10MOV-17 and 10MOV-18, will be permitted to support the operation of the SOC mode of RHR; and 2) a portion of the RHR injection valve (10MOV-25A, B) control circuit will be satisfied. This separate RHR injection valve control circuit will automatically close the RHR (LPCI) injection valves 1OMOV-25A and 10MOV-25B if 10MOV-17 and 10MOV-18 are both open and a reactor low level or a drywell high pressure signal is received. The purpose of this second portion of the logic is to isolate a possible reactor drain path which could have been created during the SOC mode of operation.
High Reactor pressure - Plant operating:
Under this condition, an isolation of the SOC mode of RHR is required per TS Table 3.3.6.1-1 Function 6a. This requirement is satisfied as follows: When reactor pressure is greater than the SOC logic trip value, the isolation logic will trip generating a reactor high-pressure logic condition to automatically close the SOC isolation valves 10MOV-17 and 10MOV-18. The purpose of closing 10MOV-17 and 10MOV-18 is to prevent the Page 6 of 7

JAFP-10-0072 Attachment 1 Responses to Request for Additional Information Questions over pressurization of the RHR pump suction piping. In addition, the logic associated with the RHR injection valve control circuit (1 OMOV-25A and 10MOV-25B) that is in place only during SOC mode of RHR will be tripped (disabled) since 10MOV-25A and 10 MOV-25B are closed during normal plant operation.
The discussion above supports the association between Section 1.2.6 of JAF-CALC-NBS-02052 and the TS requirement of Function 6a in TS Table 3.3.6.1-1.
3.e Due to the size of the source documents, only the pertinent excerpts that support the environmental conditions as presented in Section 3.2 of JAF-CAL-NBS-02052, Table 1 are attached. As discussed in Section 1.2.5 of JAF-CALC-NBS-02052, seismic uncertainty effects are not included, because the instrument loop's function is not required during and following a seismic event. Should a seismic event occur, operability of the instruments will be evaluated as part of an abnormal operating procedure.
Page 7 of 7

JAFP-10-0072 Attachment 2 Source Documents Excerpts (12 Pages)

                                                                                                                 * "
* I ENVIRONMENTAL QUALIFICATION SERVICE CONDITIONS ENTERGY NUCLEAR NORTHEAST JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MISC-04046 REVISION 0 CAUTION This document is intended specifically for the EnviroDmenul QU2lificatioD (EQ) Program.
Other users are cautioned that the information presented herein may represent very conservative, usually worst-<:ase, information for Environmental Qualification of safety-related electrical equipment. It is based on very specific assumptions and mayor may not be valid for a particular design purpose.
Prepared bY:~/ /t:tu!f(~L                                                         Date:
J. K. Seill R. R. Burdett I C. W. Himmelberger I K. Weise Parson. E&C ***                                                               I Reviewed by:         ifo/~
Paul L. Bunker I Sura   Dasgupt 1M. K. Reed / P. M Witman Date:                                 I t
I!
Parsons E&C ***
Appmvedby,       ~~                                                              Date:
Paul L. EillI1kefItIf1Ithoasgu Parsons E&C ...
            **. Report JAF-RPT-MISC-04046, ReviJion 0, was prepared, reviewed and approved                  by a project tum. Refer to the attached Panons E&C Design VerificatlOD Record.
This (S a Quality Reco~:                       O~/C'5
* QACAT:                                 _
PrLBI:                                   _
COMP. PR1N"mrrr IfY:.: .

                       " ..                                                                                                              JAP-RPT*MISC*04046 Revision 0 Page 42 Table 4.2.1.J.l                 Nonnal Reactor BuDding Temperatures EQ Node    Old HELB            Temp.        Percent (%j          Reference                                  Comment Node                (OF)        at Temp.
RR300RWCUA         300-2                120            100          [5.4.2.1.1.5}         Normal ambient temperatures in these (5.4.2.1.1.2)         rooms havc generally increased from the' original design values provided in GE.
Specification No. 22A2928 [5A.13.1}.
Therefore, lhe EQ nonnal ambient temperature has been revised upwards periodically.
RO.1()OSW ITCH     .10()-3               llS            90           /5.4.2.1.1.2) 100              10 RB300MGSET         300-4                 85             90           [5.4.2.1.1.2 J 100             10 RB300NORTH         300-5                 85             90           [5.4.2.1.1.2]
100             10 RIJ300STAIR        300-6                1!5            90           [5.4.2.1.1.2J 100              10 RB300HTEXCH        300-7               105             100          [5.4.2.1.1.21 RB300TANK          300-8                85             90           [5.4.2.1.1.2]
100              10 RB.100SAMPLE       300-9                85             90           [5.4.2.1.1.2]
100               10 RB300SLUDGE      .100-10              104             100             (5.4.13.IJ          No temperature data identifred. No EQ equipment in enclosed room location.
Temperature assumed equivalent to worst-case RB general area normal, from RB326              326-1 326-2 85 100 90 10
[5.4.2.1.1.2J GE Report 22A2928 (5.4. I3.1].
Per Reference [5.4.2.1.1.2. page 91. The Fuel Pool Cooling Heat Exchanger Room 326-3                                                                  equipment does not provide a significant 326-4                                                                  source of heat to the room due to the                           ,
326-5                                                                  following:                                                     ,, .
326-6 I. The maximum design temperature of 326-7 326-8 2.
the fuel pool (l35&deg;F) is greater than the fuel pool inletlemperacure to the heat exchangers (I 2.5 oF). and The heat exchanger:. are operated on h
                                                                                                                    .m intermittent basis to offset the heat load input from the stored spent                    :
I
                                                                                                                                                                                 ,I fuel.
Therefore, this room is considered 10 have nn equivalent normal ambient lemperature as the RB 326 n. General I.
* i 1                                                                       Area
       . *
* _4 *
* _.~nt'll'~-......~       p............... il:. .. ri _.. -~ - ._-. **-*l-~*-**"'----:''''::''~'''''''-''-~-~''''''''';'''--~*'''_*''''_-- --~_  --.

JAF*RPT*MISC-04046 Revblon 0 Page 4-'
I?lant and Equipment Operation Nonnal ambient tempcrnture~ may nuctuale ~ig"iticllntly belween times of plant operation and shutdown periods. Note that the average temperature for Ihe Reactor Building over the life of the plMt is currently based on plant operational temperatures, which is conservative. FUlure use of plant capacity factors and temperature monitoring of shutdown periods can provide more realistic normal ambient temperatures. Note that certain rooms. such as the RHR Heat Exchanger Rooms. may be at higher temperatures when the plant i~ not in operation.
Similar effects can be obf.erved for certain equipment during equipment operational periods. which is closely related to localized area temperatures. For example. Ihe rooms for pumps and heat exchnngers experience higher temperatures during periods of equipment operation due to the additional heat input (5.4.2.1.1.2, pg. 2}.
Note that this report (JAF-RPT-MISC-04046) does not examine internal equipment henting (e.g.. electrical MCC and switchgear enclosure, energized Solenoid Operated Valve (SOV). motor. or Motor Operated Valve (MOV) internal temperatures).
Rather. Ihis information is provided in the individual EQ Qualification Documentation Rcports (QDRs).
4.2.1.2   Pressure The normal Reactor Building pressure is maintained between (-) 0.10 inches of water gage (abbreviated in w.g.) to (-) 1.0 inches w.g. [5.4.13.1. Section A-I].
%~ 4.2.1.3    Relative Humidity Reactor Building nonnal relative humidity (RH) varies between 20% and 90% with a nominal value of 40% RH. per GE Document No. 22A2928 [5.4.13.1. Section A-I].
The power uprale effect on these values is negligible. in Ihal slightly warmer fluid in the steam lines mny increase the overall Reactor Building temperatures. This slight increase in temperature increases the capability of the air to hold moisture and result in a slight decrease in the Reactor Building nonnal relative humidity. DBD-066 "Inside Design Conditions" Table confinns the above nonnal RH values
[5.4.2.1.1.4}.
4.2.1.4  Radiation Gamma radiation is the primary contributor to nonnal service doses. There is no neutron dose outside the DrywelJ and beta dose from fission or corrosion products arl': contained within the reactor coolant and auxiliary systems.
Appendix B provides the normal radiation dose and dose rate for each Reactor Building EQ Node based on the values presented in Table 4.2.1.4.
                                                                                  ._-....I-.-.... __ .....

J AF*RPT*M1SC..Q4046 Revision 0 Page 46 Design normal radiation values in the Reactor Building were provided in GE Document No. 22A2928 (5.4.13.1. Se1:tion BJ. Subsequently. actual radiation measure~nts made          during power operation hnve been incorporated in to the radiation service conditions. Table 4.2.1.4 combines the data from the various radiation monitoring surveys/references for the EQ normal service conclitions.
The effects of power uprate are included in Table 4.2.1.4. Since primary coolant activities are controlled (for example. RWCU flow maximization, zinc injection and piping decontamination). power upra~~ conditions were not expected to significantly affect Reactor Building normal service conditions [5.4.12.6, Section 4.2.1].
Subsequent reviews of normal radiation values after power uprate identified tl1at Elev. 326' experienced incre.uscd radiation values [5.4.1.1.4.2]. A radiological survey wa<; also performed in 1999 of all Reactor Building areas. An assessment of the data in Memorandum JRES-99*200 [5.4.2.1.4.4} identified elevated radiation
'\:     doses at Elev. 326' and 344'. These effects have been incorporated into the normal service dose rates in Table 4.2.1.4.
Table 4.2.1.4       General Area Reactor Building Normal Radiation Reactor Building               Dose Rate               Dose
           . Floor Elevation ( I)              (mrlHr)               (Rad)                Reference 227' (2)                      200                 7.0E4            t  [5.4.2.1.4.1]
[5.4.2.1.4.2]
212' 0)                        50                 1.8E4               [5.4.2.1.4.1]
[5.4.2.1.4.2]
300' (5)                       50                 !.8E4              (5.4.2.1.4.1]
[5.4.2.1.4.2]
326'                         200                7.0E4               [H.) .1.4.2) 344'                          15               5.26E3               [5.4.1.1.4.2]
[5.4.2.1.4.4J (5.4.2.1.4.3J 369' (4)                        15               5.26E3                     nla
      ~:
I. For specific EQ Node doses, refer to rhe Appendices of this report.
    . 2. The Reactor Building at Elev. 227' is representative of the Torus Room, as confirmed by historical field verifications [5.4.2. I .4. I].
: 3. The Re;v;lor Building at EI~v. 272' is considered representative of the SBGT Room, which is in a separate enclosed room adjacent !o the Reactor Building general area.
SBOT equipment is considered 10 have minimal internal normal dose.
: 4. Radialion monitoring has nol been performed for Elev. 369'. The dose rale at Elev. 369' is considered the same as Ihal of Elev. 344'.
: 5. RWCU heat exchanger room dose rale is 15 RadJhr wilh a 40-yc.ll' inlegralM normal service dose of S.4E6 Rad. RWCU Iilrer nnd tank dose rotc is 10 Radlhr with a 4()..year normal service dose of 3.6E6 Rad. Values arc hased on Reference 5.4.13.1.

Document No. DBO-o70 ENTERGY NUCLEAR NORTHEAST JAMES A. FITZPATRICK NUCLEAR POWER PLANT DESIGN BASIS DOCUMENT FOR THE CONTROL ROOM RELAY ROOM VENTILATION AND COOLING SYSTEMS Print/Sian/Date                        Print/Sign/Date Prepared by:                         ENN DBD Owner:
Reviewed by:                         ENN PrOQ. Enar.:
Approved by:                         ENN Mgr. Design Ena.:

ENTERGY NUCLEAR NORTHEAST                                                           DBO-070 JAMES A. FITZPATRICK NUCLEAR POWER PLANT                                        REVISION 10 CONTROL ROOM AND RELAY ROOM VENTILATION AND COOLING SYSTEMS                       PAGE 2-18 When an air handling unit, 70AHU-3A or B, is operating, the corresponding chilled water pump, 70P-9A or B, starts automatically when the outside air temperature is above nominal 50&deg;F as sensed by temperature switch 70TS-106A or B. The respective flow switch, 70FIS-100A or B, sensing the flow of water automatically starts the corresponding chiller, 70RWC-2A or B. (Ref.
6.7.6,6.7.7,6.8.1.15,6.8.1.17,6.13.17, Subsect. 7.1.42)
The operating chiller, 70RWC-2A or B, supplies chilled water to the air handling unit cooling coils. The relative humidity of the air discharged by the air handling unit is controlled by 70TIC-1 05 and 70TIC-107 by regulating chilled water flow through control valves 70TCV-121A, B (Subsect. 2.2.3).
70TIC-105 and 70TIC-107 also control electric heaters E-8 and E-7 respectively, which are installed in the ducts, to raise the discharge air temperature and effectively reduce relative humidity as required in response to room thermostats 70TIC-106 and 70TIC-108 (Ref. 6.8.3.2, 6.8.1.1, 6.8.1.18, 6.8.2.43).
When the Control Room temperature rises above 98&deg;F, temperature switch 70TS-109A or B de-energizes the actuators of both return air dampers to open both dampers, and closes the outside air and exhaust air dampers.
(Subsect. 7.1.28, 8.1.6) It also starts the redundant standby air handling unit, exhaust fan and chiller, and interrupts the power supply to electric heaters 70E-7 and 70E-8. (Ref. 6.8.1.1, 6.8.1.6, 6.8.1.12, 6.8.1.13, 6.8.1.18, 6.8.2.43)
Water is not directly added to the control room. This lack of humidification control means that lower relative humidities cannot be controlled. Static discharges may occur due to low relative humidities. Static discharges generally can occur when humidity falls below 45%. Electrostatic charges are generated when materials of high electrical resistance move against each other. The accumulation of such charges can create static discharges and can potentially destroy data stored on magnetic disks and tapes or possibly ignite an explosive gas mixture. However, no explosive gases or safety-related equipment relying on magnetic disks or tapes exists in the control room. (Ref. 6.3.8, 6.8.3.1,6.8.3.2) (ACTS 99-45457, 99-45458, 99-45656,99-45391,99-45392) 2.2.1.2    RRHV System Normal Operating Functions Requirements: The Relay Room shall be supplied with outside filtered air (Ref. 6.7.1).
The Relay Room shall be maintained at a slight positive pressure (Ref.
6.7.1.).
l'         ~ The Relay Room shall be maintained at a temperature between 60&deg;F and 90&deg;F and a relative humidity between 40*/. and 50*/. for equipment operability (Ref. 6.10.3).

                                 *   *   ** _ * * * " _ _   . _   ~   _   . _   **   r   ~   *   ------"
INSTRUCTION MANUAL 4247-1 OPERATIONS MANUAL TRIP/CALIBRATION SYSTEM r           MODEL 5100U L' lDRosemount------------
r-

and trip status LED on when trip output is 0 Vdc. Trip TABLE 4 status output 10~lic is indicated by NORM and REV at COPPER WIRE, DC RESISTANCE                                 the trip status output/LED logic switch on the printed (Per ASTM Specification 81*56) circuit board (F igure 91.
DC Resistance at 2QoC AWG Size       1      Maximum Ohms per 1,000 Feet               Accompanying Instrumentation 1-_ _.....;8'-_.---L-....._ _~ . . _0.:.l!~._._------         ..
I----'~..::.~.--t-- -----+~-----.-
1-------=--._-+_. .-------.---. - - - - - - -
RELAYS:
Relays should be selected and supplied by the customer
__..!i....          i                    2.63.       ..       _         in accordance with specific environmental requirements.
_       .16       ....:..-.            ..4.18. _ _      _    _.
TRANSMITTE RS:
18         1__.             .. ~64 _ ...         ._
20'                           10.50                            Two-wire, 4 to 20 mA transmitters are required, and should be purchased by the customer according to par*
NOMINAL POWER CONSUMPTION (exclusive of output                           ticular requirements. Rosemount's line of Model 1151 loads):                                                                 aml 1152 Pressure Transmitters, and the Model 535E Temperature Transmitter are qualified for use in nuclear Maner Trip Unit (with or without Analog Output):
  '.;'.                                                                             power generating stations.
105 mA, Slave Trip Unit: 90 mA                                            CABINET:
calibration Unit: 140 mAo                                             A standard 19*inch rack for housing the 510DU System, power supplies, and relays should be supplied by the Readout Assembly: 475 rnA.
customer. Location of the cabinet at the customer site OUTPUTS:                                                                 should be such that the resistance of the wire used to make connections between the 510DU System and the 24 Vdc for each trip output and gross failure output; transmitters does not exceed 16 ohms. This assures a 12 Vdc for each trip status output;                                   minimum turn*on voltage for the transmitters of 15.0 24 Vdc calibration status si\lnal for remole indication;              Vdc when the power supply is at a minimum value of 22.0 Vdc (adverse operating conditions)
Optional 1 to 5 Vdc analog signals proportional to trans*
miUer input.
Environmental Specifications TRIP OUTPUT LOGIC:
PLANT ENVIRONMENTAL CONDITIONS: See Table 5.
Normal trip output logic provides a 24 Vdc output when SEISMIC VIBRATION:
transmitter current is greater than the trip point. Reversed trip output logic provides a 24 Vdc output when trans*               The Card File, Master and Slave Trip Units, and Calibra*
mitter current is less than the trip point. Trip output logic         tion Unit operate during and after exposure to seismic is indicated by NORM and REV at the trip output logic                 vibration of 11 g's peak in all axes.
switch on the printed circuit board (Figure 9).
ELECTROMAGNETIC SUSCEPTIBILITY:
TRIP STATUS OUTPUT/LED LOGIC:                                            The 510DU operates in EMI conditions normally ex*
Normal trip status output logic prOVides a 12 Vdc output              pected in a power plant control room environment, pro*
and trip status LED on when trip output is 24 Vdc, Re*                vided that shielded, wires are used for transmitter can*
versed trip status output logic provides a 12 Vdc output              nections and the auxiliary analog output.
TABLE 5                                         j'. "'~~-""')
ENVIRONMENTAL CONDITIONS                                                J Plant    I          'II                            1                             Radi~      Exposure             T !Power Operating 11 EnYiron-          Temperature               Relative        Dose Rate     f,Jntegrated Dose (!!~         ! Supply Condition      ment   i     "F     I       "c          Humidity (%1 i (Rads Si/Hourl i 6 Months f40 Years          V      (Vdc) 7

                                                                                       .,  -     -       -, .. .-  ,. _. __.__ .~  ~~~----
INSTRUCTION MANUAL 4471-1 Revi8ion A Instruction Manual
., S;i
   .;~,
Operations Manual Trip/Calibration System Model 710DU
            )
                                                                                          ~~                      Rosemount'
_. --'---~'-'-"- _... _.. __ ._--_._ .... _. . - --- .. - ._.----~-~_ .. _-    --.--.-_.. -     --~

TABLE 4                                         INPUT SENSORS:
COPPER WIRE. DC RESISTANCE                                      Transmitters: TWO-Wire or 4-wire, 4-20mA transmit-(Per ASTM Specification 81-56)                                 ters are required. and should be purchased by the customer according to particular requirements. Rose-DC Resistance at 20&deg;C                           mount's line of Model 1152 and 1153 Pressure Trans-AWG Size        I    MaKimum Ohms per 1,000 Feet                         mitters are qualified lor use in Nuclear Power Gener-8 10        +-----             0.824 1.04 ating Stations.
12         !      ._-----~~~--------                                 Resistance Temperature Detectors (ATO's); Shielded, 14         I                 2.63                                   3*wire, platinum RTO's with Ro = 100 ohms are re-16          j    .-.
4.18                                   quired and should be purchased by the customer ac-18         i                 6.64                                   cording to particular requirements.
20                          10.50 CABINET:
TRIP OUTPUT LOGIC:                                                          A standard 19-inch rack for housing the 710DU Sys-tem, power supplies, and relays should be supplied Normal trip output logic provides a 24 Vdc output                       by the customer. Location of the cabinet at the custom-when input signal is greater than the trip point. Re-                    er site should be such that the resistance 01 the wire versed trip output logic provides a 24 Vdc output                        used to make connections between the 710DU Sys-when input signal is less than the trip point. Trip                      tem and the transmitters does not exceed 16 ohms.
output logic is indicated by NORM and REV at the trip                    This assures a minimum turn-on voltage for the output logic switch on the printed circuit board (Fig-                   transmitters of 15.0 Vdc when the power supply is at a ure B).                                                                  minimum value of 22.0 Vdc.
TRIP STATUS LED LOGIC:                                                   Environmental Specifications Normal trtp status logic provides a 12 Vdc output and                 ENVIRONMENTAL CONDITIONS: See Table 5.
trip status LED on when trip output is 24 Vdc. Re-versed trip status logic provides a 12 Vdc output and                 SEISMIC VIBRATION:
 ~; . trip stalus LED on when trip output is 0 Vdc. Trip
: .                                                                               The Card File, Master Trip Units. and Slave Trip Units status logic is indicated by NORM and REV at the trip status LED logic switch on the printed circuit board                     operate during and after exposure to seismic vibra-(Figure 8).                                                              tion with a ZPA 01 1.17 g OBE and 1.75 g SSE. See Rosemount Report 08200037.
Accompanying Instrumentallon ELECTROMAGNETIC SUSCEPTIBILITY:
RELAVS:
The 710DU operates in EMI conditions normally Relays should be selected and supplied by the cus-                      expected in a power plant control room environment, tomer in accordance with specific environmental                         provided that shielded wires are used for all signal requirements.                                                           conneclions and the auxiliary analog output.
TABLE 5 ENVIRONMENTAL CONDITIONS Operating Condition               Normal                              Tran,lent                  Accident      (Includ" Margin, 160 for 24 hours                      185 &deg; for 6 hours OF              60 to 90                          once per year 150 0 for 8 hours Temperature 71 for 24 hours                       85&deg; for 6 hours "C                151032                           once per year                        65.6&deg; for 8 hours Relative                                                  90% for 24 hours Humidity                   40 to 50%                                                              90% for 14 hours once per year Radiation                      <~ 10 5 Rad (alf) TID over 20 years                        2x  10~ Rad (air) TID in 24 hours r-';;
Power Supply                                   22 to 28 Vdc
                                  \
                              ,~
7

Do cum e!\ t: No. -..:2:..:;2:.:..A~2.:..9;;;..2R..:.-._ _Re v. 2 General Electric Class TRANS:.4ITTAL PROJECT(S) SV..'1IlARD             PLA.~TS     236X350A,B.AG4,BG4; CHINSHAN, FITZPATRICK, FUKUSHI~.A-2, COOPER,               SUSQUE~\~A,           NEWBOLD 1&2, HATCH-2.                               Llt~RICK TITLE OF                                                                    "~I>.
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: n.      ~.    .;;"yU RESPONSI8LE           ENGINEER~ __VM:..:;.~D~O~C~H~EZ~        __ ISSUED BY_~R'""_* ....;N"'_"O=R=T=ON:.:.-.                                  DATE    APR 1 6 1:7 REFERENCES MASTER PARTS LIST (HPL) NOS,                          A61-4270, 1-132; A71-4030 (Hatch)
SPECIFICATIONS                                                                                                                                      _
DRAW I NGS                                                                                                                                       _
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Equip"""nt or Area I
Radiation 1) o;>e:-aang Oose ~te IV. Rad-Yaat. Buildinc Accident Ooa.              2)   Integrated Do.e Type            Plant O?er System'Oper              Type       00 ** KAt.          llot'1ll.al        Ac:-:1dent     :
                                                                                                                                                              ~
Control lloo1D            C:au:::oa                  I         0.001                                      I      1.S    :II: 10 2                  i Valve , Pucp RO<X:la     C:acz:sa                              0.020 I
7.0         10)                 . I, I
:II:
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Ca=a
                                                                    . 20.0
                                                                                           ,                                       :II:
                                                                                                                                                                ~
Centri!uge                ~c:a&                                100.0                                             1      :II: 10 '
I V. UACrOlt CO!rrRDL !lOOK II 2              0 Control llDooa           C&#xa3;cu                    O.OOOS                                                        1.7S    lC    10      J :II: 10 I ~
: 1)  C~a    Do.e KAce                               b.d. (c.r~o~ut' I
Neutron flux                                   Neutrons/co     .ac
: 2)  C:.-. Ooaa                                     lLWa (c....-Son)
                   !feecron fluu:e.                               Ncutr=al.:2 (liVI)
No =al Cand 1t i     OtLS                      tnteg:-a:ed over 40 year. - 100 percent load faccor .t rated p~er.                              !r Accidant Cond1tiona                            Integracad avar 6 conch. ~OCA Analyai. vas based 00 the assumpcion that                         l 100 percent of the noble gases. SO percenc of c.,<.: '",lugells, and 1 percent of elle solid fission prcducca were released {rom the core.
 )
APR 1 f) 1971}}