ML20099E380

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Proposed Tech Specs Re Fire Protection
ML20099E380
Person / Time
Site: Millstone Dominion icon.png
Issue date: 07/31/1992
From:
NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20099E319 List:
References
NUDOCS 9208100148
Download: ML20099E380 (35)
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Text

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Docket No. 50-423 B14104 l

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4 Attachment 1 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications i

July 1992

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INDEX DEFINITIONS .

SECTION- PAGE 1.0 DEFINITIONS 1 .- 1. ACTI0N....................................................... 1-1 1.2 ACTUATION LOGIC TEST......................................... 1-1 1.3 ANALOG CHANNEL OPEP.ATIONAL TEST.............................. ~'

l.4 AXIAL FLUX DIFFERENCE........................................ 1-1 1.5 CHANNEL-CALIBRATION.......... ............................... 1-1 1.6 CHANNEL CHECK................................................ 1-1 1.7 CONTAINMENT INTEGRITY.............. ..................... 1-2 1.8 CONTROLLED LEAKAGE............... . ................... 1-2 1.9 CORE ALTERATIONS...... ., ............................. 1-2 1.10 DOSE EQUIVALENT I-131........................................ 1-2 1.11 E - AVERAGE DISINTEGRATION ENERGY............................ 1-2 1.12 ENCLOSURE BUILDING INTEGRITY................................. 1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 1.14 DELETED 1.15 FREQUENCY N0TATION........................................... 1-3 1.16 IDENTIFIED' LEAKAGE........................................... 1-3 1.17 MASTER RELAY TEST............... ....................... ... 1-3

-1.18 MEMBER (S) OF IHE PUBLIC...................................... 1-4 1.19 OPERABLE - OPERABILITY....................................... 1-4

1.20' OPERATIONAL MODE - M0DE...................................... 1-4 1.21 PHYSICS TESTS................................................ 1-4 1.22 PRESSURE BOUNDARY LEAKAGE................................. .. 1-4 1.23 PURGE --PURGING....................................... . .... 1-4 1.24- QUADRANT POWER-TILT RATI0.............................. ... 1-5

~ 1.25 RAD!DACTIVE WASTE TREATMENT SYSTEMS.......................... 1-5 1.26 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REM 00CM)................................. 1-5 1.27 RATED. THERMAL P0WER.......................................... 1-5 1.28' REACTOR TRIP S"C'EM RESPONSE TIME............................ 1-5 1.29 REPORTABLE EVEni...................... ...................... 1-5 1.30 SHUTDOWN MARGIN. .................................... ....... 1-5 1.31 SITE B0VNDARY..............................................., 1-6 MILLSTONE - UNIT 3 1 0082

INDEX LIMITING CONDITIONS FOR QPERATION AND SURVEILLANCE REW IREMENTS SECTION ILA.GE

TABLE 3.3 5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............ 3/4 3-32

-TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............... 3/4 3-36 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Pl ant Operations. . . . . . . . . . . . . . . 3/4 3-42 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR_ PLANT'0PERATIONS.................................... 3/4 3-43 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS....... ............ 3/4 3-45 Movabl e Incore Detectors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-46

. Seismic Instrumentation................................. 3/4 3-47 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION. ................. 3/4 3-48 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENiATION SURVEILLANCE REQUIREMENTS............................................ 3/4 3-49 Meteorological Instrumentation.......................... 3/4 3-50

-TABLE 3.3-8 METEOR 9 LOGICAL HONITORING INSTRUMENTATION............ 3/4 3 51

-TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................ 3/4 3-52 Romote Shutdown Instrumentatien......................... 3/4 3-53

. TABLE 3.3-9 REMOTE SHUTDOWN INSTRUMENTATION......................

3/4 3-54 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SU_RVEILLANCE REQUIREMENTS..................... ......... 3/4 3-58 Accident Monitoring Instrumentation.................... 3/4 3-59 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION................. 4 3-60 TABLE'4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................ 3/4 3-62 T'ABLE:3.3-11 DELETED- l Loose-Part Detection System............................. 3/4 3-68 Radioactive Liquid Effluent' Monitoring Instrumentation.. 3/4 3-69 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ..................................... 3/4 3-70 TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............... 3/4 3-72 Radioactive Gaseous Effluent Monitoring Instrumentation. 3/4 3-74 HIU STONE - UNIT 3 vi 008)

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INDEX LlIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS ,

1 SECTION PAGE l

--TABLE 3.7-3 STEAM LINE SAFETY VALVES PER L00P...................... 3f 4 7-3 Auxiliary Feodwater System.................. ............ 3/4 7-4 Demineralized Water Storage Tank......................... 3/4 7-6 Specific Activity........................................ 3/4 7-7 TABLE 4.7-1 SECONDARY C00LANI SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM...................................... 3/4 7-8 Mair. Steam Line Isol ation Valves. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-9

'3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-10 3/4.7.3- REACTOR PLANT-COMP 0NENT COOLING WATER SYSTEM............. 3/4 7-11 3/4.7.4 SERVICE-WATER SYSTEM..................................... 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK.......................................  ;/4 7-13 3/4.7.6 FLOOD PROTECTION......................................... 3/4 7-14 3/4.7.7 _ CONTROL ROOM EMERGENLY VENTILATION SYSTEM................ 3/4 7-15 3/4.7.8 _ CONTROL K00M ENVELOPE PRESSURIZATION SYSTEM.............. 3/4 7-18 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM......................... 3/4 7-20' 3/4.7.10 SNUBBERS................................................. 3/4 7-22 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST.......... 3/4 7-27 3/4,7.11 SEALED SOURCE CONTAMINATION,............................. 3/4 7-28

~3/4.7.12 DELETED TABLE 3.7-4 DELETED TABLE 3.7-51 DELT.TED

.3/4.7.13 DELETED

, 3/4.7.14 AREA TEMPERATURE MONITORING.............................. 3/4 7-44 TABLE 3.76 -AREA TEMPERATURE MONITORING........................... 3/4 7-45 i

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SECTION pag A

3/4.7.11 SEALED SOURCE CONTAMINATION......................... B 3/4 7-6 3/4.7.12 DELETED 3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONITORING................... ..... B 3/4 7-8 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION........................ ... B 3/4 8-1 3.4.8.4 ELECTRICAL EQUIPMENT-PROTECTIVE DEVICES.............. B 3/4 8-3

-3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.................................. B 3/4 9-1 3/4.9.2 INSTRUMENTATION...................................... B 3/4 9-1 3/4.9.3 DECAY TIME........................................... B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.................... B 3/4 9-1 3/4.9.5 COMMUNICATIONS....................................... B 3/4 9-1 3/4.9.6 REFUELING MACHINF.................................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS.............. B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAAND COOLANT CIRCULATION........ B 3/4 9-2 3/4.9.9- CONTAINhENT' PURGE AND EXHAUST ISOLATION SYSTEM.......

B 3/4 9-2 3/4.9.10 and'3/4.9.11 WATER. LEVEL.- REACTOR VESSEL AND STORAGE-P00L......................................... B 3/4 9-3 3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM.................. B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN....................... .............. B 3/4-10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRI30 TION LIMITS. B 3/4 10-1 3/4.10.3 PHYSICS TESTS......................................... B 3/4 10-1

- 3/4.10.4 REACTOR COOLANT L00PS................................. B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN................. B 3/4 10-1 MILLSTONE - UNIT 3 xv 0085-

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1 DEFINITIONS ENCLOSURE BUILDING INTEGRITY 1.12 ENCLOSURE BUILDING INTEGRITY shall exist when:

a. Each door in each access opening is closed except when the access opening is being used for normal transit entry and exit,
b. The Supplementary Leak Collection and Release System is OPERABLE, and
c. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

ENGINEERED SAFETY FEATURES RESPONSE llM1 1.13 - The ENGINEERED SAFETY FEATULc5 (E5F) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety' function (i.e., the valves travel to their required positions, pump

discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.
  • 1.14 Deleted fRE00ENCY NOTATION 1.15 The . FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED' LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

a. -Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, cr

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b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or
c. . Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TESI 1.17 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each rel ay. The MASTER RELAY TFST shall include a continuity chock oi each associated slave relay.

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l-JNSTRUMENTATION BASES REMOTE SHUTDOWN INSTRUMENTATION (Continued) instrumentation, control, and power circuits and transfer switches necessary to eliminate effects of the fire and allow operation of instrumentation, con-trol and power circuits required to achieve and maintain a safe shutdown con-dition are independent of areas where a fire could damage systems normally used to shut down the reactor. This .apability is consistent with General Design Criterion 3 and Appendix R to 10 CFR Part 50, 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. The instrumentation in-cluded in this specification are those instruments provided to monitor key variables, designated as Category 1 instruments following the guidance for classification contained in Regulatory Guide 1.97, Revision 2, "Instrumenta-tion for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and following an Accident."

3/4.3.3.7 Deleted. I r

MILLSTONE - UNIT 3 B 3/4 3-5 0094

PLANT SYSTEMS BASES 3/4.7.11 SEALED SOURCE CONTAMINATION (Continued) plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.

Saaled sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required tre be tested more often than those which are ne'. Scaled sources which are continuously enclosed within a shielded mechan ,m (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism. ,

3/4.7.12 Deleted MILLSTONE - UNIT 3 8 3/4 7-7 Amendment No. EE

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I PLANT SYSTEMS EASES 3/4.7.11 Deleted 3/4.7.14 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of i2.2*F. _

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ADMINISTRATIVE CONTROLS

. FACILITY STAFF (Continued)

b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be. in the control room;
c. At least two licensed Operators shall be present in the control room during reactor startup, scheduled reactor shutdown and during recovery from reactor trips,
d. A Health Physics Technician
  • shall be on site when fuel is in the reactor;
e. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Reactor Operator or licensed Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;
f. Deleted
g. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions. These procedures should follow the general guidance of the NRC Policy Statement on working hours (Generic Letter No.

82-12).

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  • The Health Physics Technician may be absent for a_ period of time qpt to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected absence, provided immediate action is taken to fill the required position.

l MILLSTONE - UNIT 3 6-2 Amendment No. JE 0090 I

ADlilllllilRATIVE CONTROLS 6.3 Utili staff OVAllFICATIQM o.3.1 Each member of the facil~.ty staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Radiatian Protection Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, Revision 1, May 1977. The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees.

6.4 TPAlElHQ 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Station Superintendent and

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shall meet or exceed the requirements and recommendations of Section 5.5 of  ;

ANS1 N18.1-1971 and Appendix A of 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees.

6.4.2 Deleted.

6.5 REVIEW AND AURil 6.5.1 Pl#1T OPERATIONS REVIEW COMMITTEE (PORC) h fAllClLQtl 6.5.1.1 The PORC shall function to advise the Unit Yiperintendent on all matters related to nuclear safety.

COMPOSITIOM b 6.5.1.2 The PORC shall be composed of the: _

Chairman: Unit Superintendent Vice Chairman and Member: Operations Supervisor i Member: Maintenance Superv Uor Member: Instrument and Cor trol Supervisor Member: Reactor Engineer '

Member: Engineering Supervisor or Startup Supervisw*

Member: Station Services Superintendent or Quality Services Supervisor or Radiological Services Supervisor Member: Staff Engineer **

  • Wen positich is staffed.
    • The naff Engineer member of the PORC shall have an academic degree in engiotering or physical science field; and, in addition, shall have a five years technical experience, of which a minimum of three years shall be in the nuclear power plant industry.

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ADMINISTRATIVE CONTROLS ALTERNATES S  :

6.5.1.3 All alternate members shall t,e appointed in writing by the PORC i Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PORC activities at any one time.

MEETING FRE0VENG.1 j 6.5,1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Lhairman.

t- 000 RUM 6.5.1.5 The quorum of tue PORC shall consist of the Chairman or Vice Chairman or Station Superhtendent and four nambers incluJing alternates.

FESPONSIBILITIES  ;

6.5.1.6 The PORC shall be renponsible for:

a. Review of: (1) all procedurt:s, except wmon site procedures, required by Specif'. cation 6.8 and changes thereto, and (2) any other proposed procedres or changes thereto as determined by the Unit Superintendent io affect nuclear safety;
b. Review of all proposed tests and experiments that affect nuclea.- ,

safety;

c. Review of all proposed changes to Sections 1.0-5.0 of these Technical Specifications;
d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety;
e. . Investigation of all violations of the Technical Specifications, including the preparation end forw:rding of reports covering evalua- -

tion and recommendations to prevent itcurrence, to the Vice .

President-Nuclear Operations and to the Chairman of the Nuclear Review Bocrd; L f. Review of all REPORTABLE EVENTS; p

g. Review of facility operations te detect potential safety huards; h.. Performance of special reviews, investigations, or analyses and reports thereon as requested by the Chairman of the Nuclear Review Board or the Station Superintendent; and
i. Render determinations in writing with regard to whether or not each item considered under Specification 6.5.1.6a. through d. above a n- ,

stitutes an unreviewed safety question.

j. Review of Unit Turbine Overspeed Protection Maintenance and Testing Program and revision thereto.
k. Review of the Fire Protection Program and impicmenting procedures l

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6DMINISTRATIVE CONTRQLS l

DM0RW G.5.2.5 A quorum of the 50RC shall consist of the Chairman and four members including alternates.

RESPONSIBILITIES 6.5.2.6 The SORC shall be responsible for:

a. Review of (1) all common site procedures required by Specifica-tion 6.8 and changes thereto, (2) any other proposed procedures or changes thereto as determined by the Station Superintendent to affect site nuclear safety;
b. Review of all aroposed changes to Section 6.0 " Administrative Controls" of taese Technical Specifications;
c. Performance of special reviews and investigations and re p rts as -

requested by the Chairman of the Site Nuclear Review Board;

d. Review of the Plant Security Plan and implementing procedures and submittal of recommended changes to d e Chairman of the Site Nuclear Review Board;
e. Review of the Emergency Plan and implementing procedures, and sub-mittal of reconnended changes to the Chairman of the Site Nuclear '

Review Board;

f. Review of all common site proposed tests and experiments that affect nuclest safety;
g. Review of all common site proposed changes or modifications to systems or equipment that affect nuclear safety; and
h. Render determinations in writing or meeting minutes with regard to whether or not each item considered under Specification 6.5.2.6(a) through (g) above constitutes an unreviewed safety question,

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i. Review of the common site fire Protection Program and implementing procedures.

AUTHORill 6.5.2.7 The 50RC shall:

a. Recommend to the Station Superintendent written approval or disapproval in meeting minutes of items considered under Specification 6.5.2.6(a) through (g) above, and
b. Provide immediate written notification or reeting minutes to the Vice President-Nuclear Operations and the Chairman of the Site Nuclear Review Board of disagreement between the 50RC and the Station Superintendent; however, t!w Station Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

PECORDS 6.5.2.8 The SORC shall maintain written minutes of each meeting and copies shall be provided to the rice President-Nuclear Operations and Chairman of the Site Nrlear Review Bnard.

MILLSTONE - UNIT 3 6-10 C092

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(5) la tenktinitterM f re NLLSssthn 5 , L.4A_ S E R . 6 . 6 .2.Jffd Prior to May 25, 1J36, NNECO shall submit the '.n s e rvi c e inspection program which conforms to the ASME Code in effect on November 25, 1984 in accordance with 50.55(a)(g)(4), for NRC staff review and approsal, (6) Deleted (7) Deleted } _

(8) Deleted (9) Deleted

Docket No. 50-423 814104 1

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Attachment 2  ;

Millstone Nuclear Power Station, Unit No. 3 Proposed Changes to the.0perating License i'

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July 1992 l

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4 (5) Jnservice Inspected Procram (Section 5.2.4.3. SER 6.6.3 SER1 Prior to May 25, 1986, NNEC0 shall submit the inservice inspection program which conforms to the ASME Code in effect on November 25, 1984 in accordance with 50.55(a)(g)(4), for NRC r.taff review and approval.

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follow-up within thirty days in accordance with the procedures described in 10 CFR 50.73(b), (c), and 's:).

G. The licensees shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accor dance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

H. This license is effective as of the date of issuance and shall expire at midnight on November 25, 2025. -

1. Fire Protection (Section 9.5.1. SER. SSER 2. SSER 4. SSER 5)

Northeast Nuclear Energy Company sna11 implemant and maintain in effect all provisions of the approved fire protection program as described in the final Safety Analysis Report for the facility and as approved in the SER (NUREG-1031) issued July 1984 and Supplements Nos. 2, 4, and 5 issued September 1985, November 1985 and January 1986, respectively, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those charges would not adversely affect the ability to achieve and maintain safe shutdown in the event f a fire.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by 51. R. Denton -

Harold R. Denton, Director Office of Nuclecr Reactor Regulation Attachments / Appendices

1. Appendix A - Technical Specifications (NUREG-1176)
2. Appendix B - Environmental Protection Plan Date of Issuance: January 31, 1986 P

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