View source for B13056, Proposed Tech Specs Amending Sections 1.0,3/4.2,3/4.9, 3/4.10,3/4.11,5.0 & 6.0

{{Adams
| number = ML20205N004
| issue date = 10/26/1988
| title = Proposed Tech Specs Amending Sections 1.0,3/4.2,3/4.9, 3/4.10,3/4.11,5.0 & 6.0
| author name = 
| author affiliation = NORTHEAST NUCLEAR ENERGY CO.
| addressee name = 
| addressee affiliation = 
| docket = 05000213
| license number = 
| contact person = 
| document report number = B13056, NUDOCS 8811030404
| package number = ML20205M993
| document type = TECHNICAL SPECIFICATIONS, TECHNICAL SPECIFICATIONS &amp; TEST REPORTS
| page count = 154
}}

=Text=
{{#Wiki_filter:Docket No. 50-213 813056 Attachment 1 Haddam Neck Plant Revised Technical Specifications t
t October 1988 h$kIO$DOOk N3 P

SECTION 1.0 DEFINITIONS

l.0    DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.
ACTION 1.1    ACTION shall be that part of a Technical Specification which prescribes remedial acasures required under designated conditions.
ANALOG CHANNEL OPERATIONAL TEST 1.2    An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, interlock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.
AXIAL OFFS 11 1.3    AXIAL OFFSET shall be the ratio of the difference in upper and lower detector signals, to the sum of upper and lower detector signals, as measured by the excore instruments.
CHANNEL CALIBRATION 1.4    A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions, and may be serformed by any series of sequential, overlapping or total c1annel steps such that the entire channel is calibrated.
CHANNEL CHECK 1.5    A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
HADDAM NECK                            1-1

DEFINITIONS CONTAINNENT INTEGRITY 1.6    CONTAINMENT INTEGRITY shall exist when:
: a. All penetrations required to be closed during accident conditions are either:
: 1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or
: 2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.
: b. The equipment hatch is closed and sealed,
: c. The air lock is in compliance with the requirements of Specification 3.6.1.3,
: d. The containment leakage rates are within the limits of Specification 3.6.1.2, and
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: e. The sealing mechanism associated with each penetration (e.g.,
welds, bellows, or 0-rings) is OPERABLE, CONTROLLED LEAKAGE 4
1.7    CONTROLLED LEAKAGE shall be that seal water flow returned from the i  reactor coolant pump number 2 seals.
1 CORE ALTERATION 1.8    CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe position.
!  DOSE E0VIVALENT I-131 1.9    DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurle/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, 1-133, 1-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table E 7 of U. S. Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor
. Effluent for the Purpose of Evaluating Compliance with 10CFR50, Appendix I,"
!  Revision 1, October 1977.
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l  HADOAM NECK                               1-2
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OEFINITIONS E - AVERAGE DISINTEGRATION ENERL' l.10 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (HeV/d) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity.
END-OF-CORE LIFE 1.11     END-0F-CORE LIFE shall correspond to a reactor operating condition with all control rod banks fully withdrawn, essentially 0 ppm boron concentration in the reactor coolant, and the Reactor Coolant System average temperature (Tavg) no longer maintained at normal operating temperature or the normal rated thennal power (RTP) no longer maintained.
FRE0VENCY NOTATION 1.12     The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
IDENTIFIED LEAKAGE 1.13     IDENTIFIED LEAXAGE 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, or
: 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 B0UNDARY LEAKAGE, or
: c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.
LINEAR HEAT GENERATION RATE 1.14     LINEAR HEAT GENERATION RATE (LHGR) < hall be the heat generation per unit length of fuel rod.      It is the integral of the heat flux over the heat transfer area associated with the unit length.
HEMBER(S) 0F THE PUBLIC 1.15     HEMBER(S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.
HADOAM NECK                               1-3

L DEFINITIONS NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR N
1.16    NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, F      shall be the ratio of the integral of linear power along the rod with th$H,ighest h      integrated power to the average rod power.
OPERABLE - OPERABILITY 1.17    A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),
and when all necessary attendant instrumentation, controls, el u trical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).
OPERATIONAL MODE - MODE 1.18    An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclu-sive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.
PHYSICS TESIS 1.19     PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core, (1) described in the FSAR, (2) authorized under the provisions of 10CFR50.59, or (3) otherwise approved by the Commission.
PRESSURE BOUNDARY LEAKA3E 1.20    PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube lirakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.
PURGE PURGING 1.21    PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating coadition, in such a manner that replacement air or gas is required to purify the confinement.
QUADRANT POWER TILT RATIO 1.22    QUADRANT POWER TILT RATIO shall be the ratio of the maximum quadrant power to the average quadrant power as determined by the excore detectors.
RADI0 ACTIVE WASTE TREATMENTS SYSTEMS 1.23    RADI0 ACTIVE WASTE TREATMENT SYSTEMS are those liquid, gaseous and solid waste systems which are required to maintain control over radioactive material in order to meet the LCO's set forth in those specifications.
HADDAM NECK                           1-4

DEFINITIONS RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REMODCM) 1.24    A RADIOLOGICAL EFFLUEtiT HONITORING MANUAL shall be a manual containing the site and environmental sampling and analysis programs for measurements # radiation and radioactive materials in those exposure pathways and for t50se radionuclides which lead to the highest potential radiation ecosures to individuals from station operation.
An OFFSITE DOSE CALCULAi10N NJiVAL (ODCH) shall be a manual containing the methodology and parameters to be used in the calculation of offsite
,  doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation Alarm / Trip Setpoints. Requirements of the REMODCH are provided in Specification 6.13.
RATED THERMAL PQWEB 1.25    RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 1825 MWt.
REPORTABLE EVENT 1.26    A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.
RHR LOOP 1.27    An RHR LOOP consists of a Residual Heat Removal (RHR) pump, a dedicated RHR heat exchanger either from the same train or from the opposite train and all other necessary piping and components required to receive and cool reactor coolant.
SHUTDOWN MARGIN                                                                ,
1.28    SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical, or would be subcritical from its present condition assuming all rod cluster assemblies are fully inserted except for the single rod cluster of highest reactivity worth which is assumed to be fully withdrawn.
SITE B0UNDARY 1.29    The SITE BOUNDARY shall be that line beyond which the land is neithe owned, nor leased, nor otherwise controlled by the licensee.
SOURCE CHECK 1.30   A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.
l HADDAM NECK                              1-5                                     l l

DEFINITIONS STAGGERED TEST BASIS 2
1.31                                         A STAGGERED TEST BASIS shall consist of:
: a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and
: b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
THERMAL POWER 1.32                                         THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TRIP ACTUATING DEVICE OPERATIONAL TEST 1.33                                         A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.
UNIDENTIFIED LEAKAGE 1.34                                         UNIDENTIFIED LEAKAGE shall be all leakage which is neither IDENTIFIED LEAKAGE nor CONTROLLED LEAKAGE.
I VENTING 1.35                                         VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that J,                                         replacement air or gas is not provided or required during VENTING, a                                         Vent, used in system names, does nt,t imply a VENTING process.
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i HADDAM NECK                                                                   16 j

TABLE 1.1 FREQUENCY NOTATION NOTATION                        FREQUENCY S                 At least once per 12 hours.
D                 At least once per 24 hours.
W                 At least once per 7 days.
B                 At least once per 14 days.
M                 At least once per 31 days.
SW                At least once per 42 days.
t Q                 At least once per 92 days.    ,
SA                At least once per 184 days.
R                 At least once per 18 months.
S/U               Prior to each reactor         '
criticality.
P                 Prior to each release.        I N.A.              Not applicable.
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c HADDAM NECK                1-7

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TABLE 1.2 N
9                                                      OPERATIONAL MODES REACTIVITY                 % RATED                    AVERAGE COOLANT MODE                    CONDITION. K             THERMAL POWER
* TEMPERATURE 77
: 1. POWER OPERATION           greater than or          greater than 5                 gregterthanorequalto equal to 0.99                                           350 F
: 2. STARTUP                   greater th n or equal    less than or equal             gregter than or equal to to 0.99                  to 5                           350 F
: 3. HOT STANOBY               less than                0                              gregter than or equal to 0.99                                                    350 F
 -  4. HOT SHUTDOWN              less than                0                              T     less than 350'F S                              0.99                                                    b8Egreaterthan200'F
: 5. COLD SHUTDOWN             less than                0                              lesg than or equal to 0.99                                                    200 F
: 6. REFUELING **              less than or equal       0                              lesgthanorequalto to 0.94                                                 140 F
* Excluding decay heat.
    **  Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
2

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Section3/4.2 POWER DISTRIBUTION LIMITS l
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3/4.2 POWER DISTRIBUTION llMITS 3/4.2.1 AXIAL OFFSET FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.2.1.1     The AXIAL OFFSET shall be maintained within the limits of Figures 3.2-1 a or b.
APPLICABILITY: MODE 1, above 40% of RATED THERMAL POWER.                                             .
ACIl0N:
With the AXIAL OFFSET outside the Acceptable Operation Limits specified in the above figures, within 15 minutes initiate corrective action and continue the corrective action so that the AXIAL OFFSET is within limits within 2                             ,
hours or reduce THERMAL POWER to less than 40% of RATED THERMAL POWER within                        r the next 4 hours.
SURVEILLANCE REQUIREMENTS 4.2.1.1.1 1he AXIAL OFFSET shall be determined to be within the Acceptable Operation Limits of Figures 3.2-1 a or b by monitoring the AXIAL OFFSET
;       using at least two OPERABLE excore Power Range channels and applying the excore/incore correlation or. a continuous basis.
4.2.1.1.2     The excore/incore correlation shall be verified at least once                         !
per 31 EFPD and adjustad at least once per 92 EFPD using the results of the j       measurements obtained in accordance witi Specification 4.2.2.1.1.
4.2.1.1.3 The excore/incore correlation shall be determined after each                              !'
fuel loading or major change in excore Power Range instrumentation prior to exceeding 80% of RATED THERMAL POWER.
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4.2.1.1.4 The excore Power Range detectors shall be calibrated / correlated                         ,
relative to the Movable Incore Detector System measurements within 7 days after completion of incore measurements.                                                            ;
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t HADOAM NECK                                                               3/4 2-1                   !

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          . ~Y1tuME                    3.2-1 a             POWER LEVEL VS. RX1 AL OFF5ET LIM 115.                       -
0 250 EFPD. FOUR LOOP OPERAfl0N                            tt 120
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                     -40                               -20                     0                   20           40 RXIAL OFF5El 1%1 3/4 -

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                 .               .F J CUME 3.2-1 b .                            PONIM. LEVEL V5. Axl AL OFFSIT LIMJ15, 250 IFPD-IOL, FOUR LOOP SPERATION 170
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                                     -40                                 -20                                           0                          70                 40 Ax!AL OFr$ET t%I 3 / t. 2-3

3/4.2 POWER DISTRIBUTION llMITS THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.2.1.2     The AXIAL OFFSET shall be maintained within the limits of Figures 3.2-2 a or b.
APPLICABillTY: MODE 1, above 40% of RATED THERMAL POWER.
ACTIQB:
With the AXIAL OFFSET outside the Acceptable Operation Limits specified in the above figures, within 15 minutes initiate corrective action and continue the corrective action so that the AXIAL OFFSET is within limits within 2 hours or reduce THERMAL POWER to less than 40% of RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE RE0VIREMENTS 4.2.1.2.1     The AXIAL OFFSET shall be determined to be within the Acceptable Operation Limits of Figures 3.2-2 a or b by monitoring the AXIAL OFFSET using at least two OPERABLE excore Power Range channels and applying the excore/incore correlation on a continuous basis.
4.2.1.2.2 The excore/incore correlation shall be verified at least once per 31 EFPD and adjusted at llast once ser 92 EFPD using the results of the measurements obtained in acerrdance witi Specification 4.2.2.1.1.
4.2.1.2.3 The excore/ince.re correlation shall be determined after each fuel loading or major chrige in excore Power Range instrumentation prior to exceeding 50% of RATED TriERMAL POWER.
4.2.1.2.4     The excure Power Range detectors shall be calibrated / correlated relative to the Movable incore Detector System measurements within 7 days after completion of incore measurements.
HADOAM NECK                              3/4 2-4

n 100<
90 80 70 <
(-12,65)                   (10,65) 60<
50 <
(-30.45.5)                                         (20,45.5)
(-30,40)                     40                 (20,40) 30 20 10
          -5'O    -40   -3'O    -20   -10       0   10     20     30     40 5'O
                                         $ Offset FIGURE 3 2-2 a POWER vs. OFFSET, 0-250 EFPD, THREE LOOP HADDAM NECK                            3/4 2-5                     ' -

1 100    '
90 <
80 70   -
(-20,65'                     (10,65) 60   -
50
(-30,45.5)                                                          (20,45.5)
(-30,40)                                         40 <               (20,40) 30 <
20 10
                                -50                        -4'O  -fo   -2'O  -1b       0   10     2'O     36    46 50 5 Offset FIGUE 3 2-2 b POWER vs. OFFSET, 250-EOC, THREE LOOP HADDAM NECK                                             3/4 2-6

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POWER DISTRIBUTION LIMITS 3/4.2.2 LINEAR HEAT GENERATION RATE FOUR LOOPS OPERATING J,IMITING CONDITION FOR OPERATION                                                                       ,
r 3.2.2.1     All LINEAR HEAT GENERATION RATES (LHGRs) shall not exceed the following kilowatt per foot limits for cycle residency time:                                            !
I
: a. Less Than 125 EFP0              13.3 kW/ft i
: b. 125 To 250 EFPD                 13.3 kW/ft                                                  !
: c. Greater Than 250 EFPD But Litss Than                                                                              i END OF CORE LIFE                14.6 kW/ft                                                  [
APPLICABILITY: MODE 1, above 40% of RATED THERMAL POWER.                                                ;
ACTION:                                                                                                 !
With the LHGR of any fuel rod exceeding the limits specified above, initiate I
corrective action within 15 minutes and continue corrective action so that                              '
the LHGR is within the limits within 2 hours or reduce THERMAL POWER to less than 40% of RATED THERMAL POWER within the next 4 hours.                                                ;
SURVEILLANCE RE0VIREMENTS 4.2.2.1.1 LHGR's shall be determined to be within the above limits by a core power distribution measurement using the Moveable Incore Detector System in accordance with Specification 3.3.3.2 and in consideration of the factors listed in Specification 4.2.2.1.2:
: a. At least once per 31 EFPD,
: b. Prior to THERMAL POWER exceeding 80% of RATED THERMAL POWER after each fuel loading, and
: c. After reaching 100% of PATED THERMAL POWER and achieving equilibrium xenon conditions after each refueling.
HADOAM NECK                           3/4 2-7

POWER DISTRIBUTION LIMITS SVRVElllANCE RE0VIREMENTS (Continued) 4.2.2.1.2 Measured values of core power peaking factors used in determining LliGRs shall include the following allowances:
: a. Normal power peaking * **,
: b. Flux peaking augmentation factors (Power Spike)*, Figure 4.2-1
: c. Measurement uncertainty of 1.05,
: d. Statistical density factor of 1.012,
: e. Engineering factor of 1.02,
: f. Stack shortening / thermal expansion factor of 1.007, and
: g. Power level uncertaincy of 1.02.
Items a. and b. are chosen at a core height to maximize the prrduct.
Determined in accordance with Specification 4.2.2.1.1, using the thimble location which yields the higher total core peaking factor.
l HADOAM NECK                          3/4 2 8

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==Reference:==
J. Hellman and P. Sipush, Power Spike Model for Steel-Clad Cores, WCAP-8095, April 1973.
1.08'"'
1.06 '
2-E e   1.04  -
               ' n.
S Y
vt 1.02   -
                    '1.00    I 0       24          48            (2      96         120 Axial Position, Inches FIGURE 4.2-1 SPIKE PENALTY FOR STAINLESS SIEEL RCC CORES MADDAM NECK                             3/4 2-9                 *~ '

7; EDRER DISTRIBUTION LIMITS THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.2.2.2 All LINEAR HEAT GENERATION RATES (LHGRs) shall not exceed the following kilowatt per foot limits for cycle residency time:
: a. Less Than 125 EFPD                 9.295 kW/ft
: b. 125 To 250 EFPD                    9.425 kW/ft
: c. Greater Than 250 EFPD But less Than END-0F-CORE LIFE                  10.075 kW/ft APPLICABILITY: MODE 1, above 40% of RATED THERMAL PGWER.                           .
ACTION-With ae LHGR of any fuel rod excoeding the limits specified above, initiate corrective action within 15 minutes and continue corrective action so that the LHGR is within the limit withitt 2 hours or reduce THERMAL POWER to less than 40% of RATED THERMAL POWER within the next 4 hours.
SVRVEILLAt1CE RE0VIREMENTS                                                     ___
4.2.2.2.1     LHGR's shall be determined to be within the above limits by a core power distribution measurement using the Hoveable incore Detector System in accordance with Specification 3.3.3.2 and in consideration of the factors listed in Specification 4.2.2.2.2:
: a. At least once per 31 EFPD,
: b. Prior to THERMA 8. POWER exceeding 50% of RATED THERMAL POWER after each fuel loading, if the initial r.ower ascensicn is performed with 3 loops operating.
: c. After achieving 65% of RATED THERMAL POWER and w hieving equilibrium xenon conditions after each refueli;u If the        initial power ascension is performed with three loops operating, i
HADDAM NECK                            3/4 2 10

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POWER DISTRIBUTION LIMITS                                                                     l SURVEILLANCE RE0VIREMENTS (Continued)                                        __               l
      /. 2.2.2.2 Heasured values of core power peaking factors used in determining LHGRs shall include the following allowances:
: a. Normal power peaking * **,
: b. Flux peaking augmentation factors (Power Spike)*, Figure 4.2-1
: c. Heasurement uncertainty of 1.05,
: d. Statistical density factor of 1.012,
: e. Engineering factor of 1.02,
: f. Stack shortening / thermal expansion factor of 1.007, and
: g. Power level uncertainty of 1.02.
Items a. and b. are chosen at a core height to maximize the product.
      **    Determined in accordance with Specification 4.2.2.2.1, using the thimble location which yields '     higher total core peaking factor.
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HADDAM NECK                         3/4 2-11

POWER DISTRIBUTION LIMITS 3/4.2.3   NUCLEARENTHALPYRISEHOTCHANNELFACTOR-Fh FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION                                ;
3.2.3.1    The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, FAH, shall be limited by the following relationship:
F AH   1 1.60 (1.0 + 0.3 (1 -P)]
where:       P-   THERMAL & WEB                , and RATED THERMAL POWER N
F     NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, is defined as the ratio of t$, integral of linear power along the rod with the highest integrated power to the average rod power.
APPLICABILITY: MODE 1.
ACTION:    With F    g   outside of the above specified limit:
: a. Within 2 hours either-
: 1. Restore F AH to within the above limit, or
: 2. Reduce THERMAL POWER to less than 65% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to less than or equal to 72% of RATED THERMAL POWER.
: b. Within 24 hours of initially beingNoutside the above ilmit, verify through incore flux mapping that F         is restored to within the abovelimit,orreduceTHERMALPOWNtoler.sthan5%ofRATED THERMAL POWER within the next 2 hours; and
: c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTIONS a.2 and/or b. abgve; subsequent THERMAL POWER operation may proceed provided that FA u is demonstrated, through incore flux mapping, to be within the Wbove limit prior to exceeding the following THERMAL POWER levels:
: 1. A nominal 50% of RATED THERMAL POWER,
: 2. A nominal 75% of RATED THERMAL POWER, and
: 3. Within 72 hours of attaining greater than or equal to 95% of RATED THERMAL POWER.
HADDAM NECK                              3/4 2-12

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POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1.1   The provisions of Specification 4.0.4 are not applicable.
4.2.3.1.2 The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR shall be determined to be within the above limit:
: a. Prior to operation above 80% of RATED THERMAL POWER after each fuel loading, and
: b. At least once per 31 Effective Full Power Days.
RADDAM NECK                          3/4 2-13

l POWER DISTRIBUTION LIMITS IEEE LOOPS OPERATINIa LIMITING CONDITIO:1 FOR OPERATION 3.2.3.2     The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, FAH, shall be limited by the following relationship:
FAH s 1.64 (1.0 + 0.3 (0.65 - P)]
where:     P-    THERMAL POWER                  , and RATED THERMAL POWER N
F     NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, is defined as the ratio of t$, integral of linear power along the rod with the highest integrated power to
* average rod power.
APPLICABILITY: MODE 1.
ACTION:     With F AH outside of the above specified limit:
: a. Within 2 hours either:
: 1. Restore F AH to within the above limit, or
: 2. Reduce THERMAL POWER to less than 20% of RATED THERMAL POWER and reduce the Power Range Neutron Flux High Trip Setpoint to less than or equal to 23% of RATED THERMAL POWER.
: b. Within 24 hours of initially beingN utside the above limit, verify through incore flux mapping that F         is restored to within the above limit, or reduce THERMAL POWN to less than 5% of RATED THERMAL POWER within the next 2 hours; and
: c. Identify and correct the cause of the out-of-limit condition prior
;              to increasing THERMAL POWER above the reduced THERMAL POWER limit requiredbyACTIONSa.2.and/ orb.abgve;subsequentTHERMALPOWER operation may proceed provided that F A is demonstrated, through incore flux mapping, to be within the gbove limit prior to exceeding the following THERMAL POWER level:
: 1. A nominal 50% of RATED THERMAL POWER i
l HADDAM NECK                               3/4 2-14

e-POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS 4.2.3.2.1   The provisions of Specification 4.0.4 are not applicable.
4.2.3.2.2   The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR shall be determined to be within the above limit:
: a. Prior to operation above 50% of RATED THERMAL POWER after each fuel loading if the initial power ascension was performed with three loops operating.
: b. At least once per 31 Effective Full Power Days.
1
(
l HADDAM NECK                           3/4 2-15                                          ,

POWER DISTRIBUTION LIMITS
  .3/4.2.4     00ADRANT POWER TILT RATIO LIMITING CONDITIONS FOR OPERATION 3.2.4    The QUADRANT POWER TILT RATIO shall not exceed 1.02.
APPLICABlul.1: MODE 1, above 50% of RATED THERMAL POWER.
ACTION:
: a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:
: 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a)     The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)     THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
: 2. Within 2 hours either:
a)     Reduce the QUADRANT POWER TILT RATIO to within its limit, or b)     Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly re<' ace the Nuclear Overpower Setpoints within the ne.; 4 hours.
: 3. Verify that the QUADRANT POWER TILT 'MIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 65% of RATED T WlMAL POWER within the next 2 hours and reduce the Power Rar n Neutron Flux-High Trip 1
Setpoint to less than or equal w 72% of RATED THERMAL POWER l                    within the next 4 hours; and f
: 4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed )rovided that the QUADRANT POWER TILT RATIO is verified wit 11a     its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.
1 l  HADDAM NECK                              3/4 2-16 1

i POWER DISTRIBUTION LIMLT.S LIMITING CONDITIONS FOR OPERATION ACTION (Continued)
: b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:
: 1. Calculate the QUADRANT POWER TILT RATIO at least once por hour until either:
a)     The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)     THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
: 2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes;
: 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 65% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Set)oint to less than or equal to 72% of RATED THERMAL POWER wit 11n the next 4 hours; and
: 4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within itr. limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.
: c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:
: 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a)     The QUADRANT POWER TILT RATIO is reduced to within its limit, or j
b)     THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
HADDAM NECK                             3/4 2-17

1
                                                                                    .1 l
POWER DISTRIBUTION LIMITS l
I LIMITING CONDIfl0NS FOR OPERATION ACTION (Continued)
: 2. Reduce THERMAL POWER to less than 20% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoint to less than or equal to 23% of RATED THERMAL POWER within the next 4 hours; a.nd
: 3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 20% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified at 95% or greater RATED THERMAL POWER.
SURVEILLANCE RE0VIREMENTS 4.2.4.1     The QUADRANT POWER TILi RATIO shall be determined within 7 days after completion of the incore power distribution measurements used to determine the initial excore/incore correlation after each fuel loading or major change in excore Power Range instrumentation.
4.2.4.2     The QUADRANT POWER TILT RATIO shall be determir.ed to be within the limit above 50% of RATED THERMAL POWER by calculating the ratio at least once per 7 days.
4.2.4.3     The QUADPANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the Movable Incore Detector System at least once per 24 hours.
l
; HADDAM NECK                             3/4 2-18 l

P0MER DISTRIBUTION LIMITS DNB PARAMETERS LJMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits shown in Table 3.2-1:
: a. Reactor Coolant System T Cold'
: b. Pressurizer Pressure, and
: c. Reactor Coolant System Flowrate.
APPLICABILITY: MODE 1.
ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5%
of RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE RE0VIREMENTS 4.2.5.1      Each of the parameters of Table 3.2-1 shall be verified to be within its limit at least once per 12 hours.
4.2.5.2     The RCS total flow rate shall be determined by heat balance within seven EFPD of achieving RATED THERMAL POWER after a refueling.
4.2.5.3 The RCS total flow rate indicators shall be subjected to a CHANNEL CAllBRAi!0N at least once per 18 months.
HADDAM NECK                             3/4 2-19

=
E
[                                                                 TABLE 3.2-1
%                                                               DNB PARAMETERS 7:
PARAMETER                                                                         LIMITS Four Loops In        Three Loops In Operation And Operation            Looo Stoo Valveis) Closed Indicated Reactor Coolant System T Cold                           s 542.0'F            1 540.6'F In/.icated Pressurizer Pressure                                   2 2000 psig*         1 2000 psig*
Indicated Reactor Coolant System Flowrate                         1 246,000 gpa        2 194,000 gpa w
2 7
Es
* Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:       (1) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the (OCA analyses are met and the ECCS Interim Acceptance Criterion limit of 2300 F peak cladding temperature is not exceeded.
3/4.2.1     AXIAL 0FFSET The AXIAL OFFSET specification provides continuous confirmation of acceptable LINEAR HEAT GENERATION RATES (LHGR) during the time interval between incore measurements.
3/4.2.2     LINEAR HEAT GENERATION RATE Limiting the peak LINEAR HEAT GENERATION RATE (LHGR) during Condition I events provides assurance that the initial condition assumed for LOCA analyses are met and the ECCS Interim Acceptance Criterion limit of 2300 0F peak cladding temperature is not exceeded.
o These limits are based on a minimum inlet temperature of 536 F at RATED THERMAL POWER.
3/4.2.3     NUCLEAR ENTHALPY pISE H0T CHANNEL FACTOR The    limit onDNBR the minimum      the NUCLEAR           ENTHALPY RISE HOT ACHANNEL FACTOR (F"H limit is not exceeded.
HADDAM NECK                                                             B3/4 2-I

PLJER DISTRIBUTION LIMITS BASES NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
N The F    is measurable, but will normally only be determined periodically as specifYedinSpecification4.2.3.1.2and4.2.3.2.2.           This periodic surveillance is sufficient to insure that the limits are maintained provided:
: a. The control rod insertion limits of Specification 3.1.3.6.1 and 2 are maintained, and
: b. The AXIAL OFFSET limits of Specification 3.2.1.1 and 2 are maintained.
The relaxation of F N as a function of THERMAL POWER allows changes in the radial power shape fur all permissible rod insertion limits. The full power limits (1.60 for four loop, and 1.64 for three loop) include a 4% incore I       measurement uncertainty.
l l
3/4.2.4 OVADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power
(       distribution satisfies the design values used in power capability analysis.
I       Radial power distribution measurements are made during STARTUP testing and periodically during power operation.
The limit of 1.02, at which corractive action is required, provides DNB and LINEAR HEAT GENERATION RATE protection with x-y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainly associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a drop)ed or misaligned control rod. In the event such action does not correct t1e tilt, the margin for uncertainty on the limiting LHGR is reinstated by reducing the maximum allowed power by 3% for each percent of tilt.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is ccnsistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with at least one set of four quadrant symmetric thimbles.
HADDAM NECK                         BS/4 2-2

E0WER DISTRIBUTION LIMITS BASES 3/4.2.5    DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits gare consistent with the accidgntassumptions.Theindicatedvaluesof540.6F(ThreeLoop)and 542.0 F (Four Loop) and 2000 psig correspond to analytical limits of 544.10F and 1960 psia, respectively, with allowances for measurement uncertainty.
The four and three loop flow rates are the values assumed in the analyses.
The measured values are reduced by the appropriate measurement uncertainty prior to comparison with the limits. The four loop measurement and uncertainty are based on a heat balance and the results are used to calibrate the RCS flow indicattrs. The three loop flow rate will be deemed acceptable if 0.76 times the measured four loop flow rate is greater than the three loop flow rate requirement. Previous flow rate measurements have shown that the three loop flow rate is 0.77 times the four loop flow rate with measurement uncertainties included.
I f
HADDAM NECK                         B3/4 2-3

SECTION 3/4.9 REFUELING OPERATIONS I
i

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1     The boron concentration of all filled unisolated portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure a K eff f 0.94 or less.
APPLICABILITY: MODE 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 greater than or equal to 30 gpm of a solution containing a minimum of 14000 ppm boron or its equivalent until K eff is reduced to less than or equal to 0.94.
SURVEILLANCE RE0VIREMENTS                                                     ,
4.9.1.1     The above reactivity condition shall be determined prior to:
: a. Removing or unbolting the reactor vessel head, and
: b. Withdrawal of any control rod in excess of 3 feet from its fully inserted position within the reactor vessel except when control rod drag testing is being performed, at which time the reactivity condition shall be determined prior to the first control rod withdrawn in excess of 3 feet and every 4 hours thereafter until completion of the testing.
4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours.
The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
HADDAM NECK                            3/4 9-1

REFUELING OPERATIONS 3/4 9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two Source Range Neutron Flux Monitors shall be OPERABLE and operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room when CORE ALTERATIONS or positive reactivity changes are taking place.
When CORE ALTERATIONS or positive reactivity changes are not taking place, at least one Source Range Neutron Flux Monitor shall be OPERABLE and operating with a visual indication in the control room and audible ir;dication in the containment.
APPLICABILITY: MODE 6.
ACTION:
: a. With less than the required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours.
EURVEILLANCE RE0VIREMENTS 4.9.2     Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:
: a. A CHANNEL CHECK at least once per 12 hours,
: b. An ANALOG CHANNEL OPERATIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and
: c. An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.
l HADDAM NECK                            3/4 9-2

REFUELING OPERATIONS 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3    The reactor shall be subcritical for at least 100 hours.
APPLICABILITY: During movement of irradiated fuel in the reactor vessel.
ACTION:
With the reactor suberitical for less than 100 hours, suspend all operations involving movement of irradiated fuel in the reactor vessel.
SURVEILLANCE REQUIREMENTS 4.9.3    The reactor shall be determined to have been suberitical for at least 100 hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor vessel.
IIADDAM NECK                         3/4 9 3

f REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
: a. The equipment hatch installed and secured in place,
: b. A minimum of one door in the airlock is OPERABLE, and
: c. Each penetration providing direct access from the containment
:             atmosphere to the outside atmosphere shall be, either:
: 1)    Closed by an isolation valve, blind flange, manual valve, or special device, or
: 2)   -Be capable of being closed by an OPERAELE containment purge,
;                   bypass or exhaust isciation valve.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.
l i
ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building, other than placing an irradiated fuel assembly into a safe storage location.
SURVEILLANCE RE0VIREMENTS 4.9.4     Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE containment purge, bypass or exhaust isolation valve or capable of being closed by an OPERABLE airlock door within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by:
: a. Verifying the penetrations referred to in Specification 3.9.4c are in their closed / isolated condition,
: b. Testing the containment purge and exhaust isolation valves in accordance with Specification 4.9.9, and
: c. Verifying that at least one of the doors in the airlock is capable of being closed.
HADDAM NECK                            3/4 9 4 1

I REFUELING OPERATIONS 3/4.9.5 COMMUNICATIONS LIMITING CCNDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling cavity manipulator crane area.
APPLICABILITY: During CORE ALTERATIONS.
ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building other than placing an irradiated fuel assembly into a sr.fe storage location.
SURVEILLANCE REQUIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling cavity manipulator crane area shall be demonstrated within I hour prior to the start of and at least once per 12 hours during CORE ALTERATIONS.
HADDAM NECK                           3/4 9 5

REFUELING OPERATION'z 3/4.9.6 MANIPULATN CRANE LIMITING COND!il0N FOR OPERATION 3.9.6 The manipulator crane and auxiliary hoist shall be used for movement of control rod drive shafts
* or fuel assemblies and shall be OPERABLE with:
: a. The manipulator crane used for movement of fuel assemblies having:
: 1)   A minimum capacity of 1650 pounds, and
: 2)   An overload cut off limit less than or equal to 175 pounds above the indicated weight for wet conditions for a stainless steel clad fuel assembly with a full length RCCA inserted.
: b. The auxiliary hoist used for latching and unlatching control rod drive shafts having a minims upacity of 1500 pounds.
APPLICABILITY: During movement of control rod drive shafts or fuel assemblies within the reactor vessel.
ACTION:
With the requiremen'.s for crane and/or hoist OPERABILITY not satisfied, place any fuel assembly or control rod in transit into a safe storage location and suspend use of the inoperable manipulator crane and/or auxiliary hoist from operations involving the movement of control rod drive shafts or fuel assemblies within the reactor vessel.
SURVEILLANCE RE0VIREMENTS 4.9.6.1     Each manipulator crane used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least 2063 pounds and demonstrating an automatic load cut off when the crane load exceeds 175 pounds above the indicated weight for wet conditions for a stainless steel clad fuel assembly with a full length RCCA inserted.           t 4.9.6.2 Each auxiliary hoist used for movement of control rod drive shafts     ,
within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations. If loads greater than the weight of a stainless steel clad fuel assembly with an RCCA inserted are to be lifted    ,
within the reactor vessel with the auxiliary hoist then a load test will t;    t performed for 125% of the weight of the load to be lifted.
* The RCCA's may be attached to the control rod drive shafts during some i
refueling evolutions.
l HADDAM NECK                            3/4 9-6 t

REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING LIMITING CONDITION FOR OPERATION 3.9.7     Loads other than fuel assemblies in excess of 1650 pounds shall be prohibited from travel over fuel assemblies in the storage pool.
APPLICABILITY: With fuel assemblies in the storage pool.
ACTION:
: a. With the requirements of the above specification not satisfied, place the crane load in a safe condition.
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.9.7 Administrative controls regarding the movement of loads in the Spent Fuel Storage Pool Building shall be adhered to by all personnel involved with lifting a load in excess of 1650 pounds.
HADDAM NECK                          3/4 9 7

REEQfLING OPERATIONS 1/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATEP. LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1               At least one RHR LOOP shall be OPERABLE and in operation.*
APPLICABILITY: MODE 6, when the water level is greater than or equal to 23 feet above the top of the reactor vessel flange.
ACTION:
With no RHR LOOP OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR LOOP to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours.
l SURVEILLANCE RE0VIREMENTS 1
I 4.9.8.1               At least once per 12 hours verify at least one RHR LOOP is in operation and circulating reactor coolant at a flow rate of greater than or equal to 2000 gpm.
* The RHR LOOP may be removed from operation for up to I hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
HADOAM NECK                                    3/4 9 8

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent RHR LOOPS shall be OPERABLE, and at least one RHR LOOP shall be in operstion.
APPLICABILITY: MODE 6 when the water level is less than 23 feet above the top of the reactor vessel flange.
ACTION:
: a. With less than the required RHR LOOPS OPERABLE, immediately initiate corrective action to return the required RHR LOOPS to OPERABLE status, or establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible,
: b. With no RH3 LOOP in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR LOOP to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours.
SVRVEILLANCE RE0VIREMENTS 4.9.8.2.1    At least once per 12 hours verify at least one RHR LOOP is      in operation and circulating reactor coolant at a flow rate of greater than or equal to 2000 gpm.
4.9.8.2.2 The RHR LOOP not in operation shall be determined OPERABLE at least once per 7 days t / verifying correct breaker alignments and indicated power availability.
HADDAM NECK                             3/4 9 9

REFUELING OPERATIONS 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9     The containment purge bypass and exhaust isolation valves shall be OPERABLE.
APPLIC9BILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.
ACTION:
: a. With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building, other than placing an irradiated fuel assembly into a safe storage location,
: b.  ' The provisions of Specifications 3.0.3 and 3.0.4 are not appitcable.
SURVEILLANCE REQUIREMENTS 4.9.9 The containment purge bypass and exhaust isolation valves shall be verified OPERABLE within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS and prior to movement of irradiated fuel within containment by verifying that containment purge, bypass or exhaust isolation valves are accessible for manual operation or verifying that the associated penetrations are blind flanged.
HADOAh NECK                               3/4 9 10

REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - REACTOR VESSEL LiliLTING CONDITION FOR OPERATION 4
3'.9.10    At least 23 feet of water shall be maintained over the top of the reactor vessel flange.  .
APPLICABILITY: During movement of fuel assemblies or control rods within the containment while in MODE 6.
ACTION:
With the requirements of the above ';pecification not satisfied, place any fuel assembly or control rod in transit into a safe storage location and suspend all further o)erations involving movement of fuel assemblies or control rods within t1e reactor vessel.
SURVEILLANCE RE0VIREMENTS 4.9.10     The water level shall be determined to be at least its minimum 4     required depth within 2 hours prior to the start of and at least once por 24 j     hours thereafter during movement of fuel assemblies or control rods.
i i
}
.I d
i i
i RADDAM NECK                           3/4 9 11

REFUELING OPERATIONS 3/4.9.12 FVEL STORAGE BUILDING AIR CLEANVP SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12     The Fuel Storage Br.ilding Air Cleanup System shall be OPERABLE and in operation.
APPLICABILITY: During operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool.
ACTION:
: a. With the Fue". Storage Building Air Cleanup System inoperable, or not operating, suspend all operations involving movements of fuel within the storage pool or crane operation with loads over the storage pool,
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SVRVEILLANCE RE0VIREMENTS 4.9.12     The Fuel Storage Building Air Cleanuo System shall be demonstrated OPERABLE and in operation:
: a. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:
: 1)    Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 4000 cfm i 10%;
: 2)    Verifying, within 31 days r.fter removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 4%; and
: 3)    Verifying a system flow rate of 4000 cfm      10% during system operation when tested in accordance with ANSI H510-1975.
HADOAH NECK                            3/4 9 13

REFUELING OPERATIONS 3/4.9.11 WATER LEVEL-STORAGE POO(
LIMITING CONDITION FOR OPERATION 3.9.11               At least 21 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABII,'IliWhenever irradiated fuel assemblies are in the storage pool.
EIIDR:
: a.            With the requirements of the above specification not satisfied, suspend all movements of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours,
: b.            The provisions of Specifications 3.0.3 r ! 3.0.4 are not i
applicable.
SURVEILLANCE RE0VIREMENTS 4.9.11               The water level in the storage pool shall be determined to be at least its minimum required depth at 12ast once per 7 days when irradiated fuel assemblies are in the fuel storage pool, t
f KADDAM NECK                                               3/4 9-12

..~         _ _     - .                  .            .                             .
F REFUELING OPERATIONS t
SURVEILLANCE RE0VIREMENTS (Continued)
: b. After every 720 hours of charcoal adsorber operation by verifying,
;               within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with              i Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2,       i March 1978, meets the laboratory testing criteria of Regulatory       '
Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 4%;
: c. At least once per 18 months by:
: 1)      Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 4000 cfm i 10%, and                                                -
: 2)      Verifying that the syst e maintaiD the spent fusi storage pool area at a negative pressure of greater than 0 inch Water -
Gauge relative to the outside atmosphere during system operation.
: d. After each complete or partial replacement of a HEPA filter bank, by verifying tht the cleanup system satisfies the in-place penetration a;'J bypass leakage testing acceptance criteria of less than 1% in a.cordance with ANSI N510-1975 for a DOP test aerosol while operating the syste:* at a flow rate of 4000 cfm i 10%; and
: e. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less    !
than 1% in accordance with ANSI N510-19',5 for a halogenated hydrocarbon refrigerant test gas while operating the system at a      ,
flow rate of 4000 cfm i 10%.                                          !
HADOAM NECK                              3/4 9-14

3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that: (1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses. A value of 0.94 or less for K          is required for this accident.      For a core configuration of all rods'f$ an additional 0.05 K        penalty is required to account for a heavy load crushingthecore'f$toamorereactiveconfiguration.
1 9.2 INSTRUMENTATION The OPERAnlllTY of the Source Range Neutron Flux Monitor: ensures that monitoring capability is available to detect changes in the reactivity condition of the core. Redundant monitoring capability is required to detect changes in the reactivity condition of the core during fuel movement.
3/4.9.3 DECAY TIME The minimum requirement for reactor subertticality prior to movement of irradiated assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products.
This decay time is consistent with the assumptions useo in the safety analysis.
3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel elemant rupture based upon the lack of containment pressurizatica potential while in the REFUELING MODE.
3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling cavity manipulator crane area personnel can be promptly informed of sigificant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.
HADDAM NECK                            B3/4 9-1

REFUELING OPERATIONS BASES 3/4.9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that:    (1) manipulator cranes will be used for movement of control rod drive shafts and fuel assemblies, (2) each crane has sufficient load capacity to lift a drive shaft or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive liftinq forces in the event they are inadvertently engaged during lifting operation,.
3/4 9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analysis.
3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one RHR LOOP be in operation ensures that: (1) sufficientcoolingcapacityisavailablegoremovedecayheatandmaintain the water in the reactor vessel below 140 F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incidet and prevent boron stratification.
The requirement to have two RHR LOOPS OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR LOOP will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and 23 feet of water above the reactor vessel flange, a large heat sink is available for core coolinn. Thus, in the event of a failure of the vperating RHR LOOP, adequate time is provided to initiate emergency procedures to cool the core.
3/4.9.9 CONTAINHENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations can be isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.
HADDAM NECK                          B3/4 9-2

REFUELING OPERATIONS BASES 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis.
3/4.9.12 FUEL STORAGE BUILDING AIR CLEANUP SYSTEM The limitations on the Fuel Storage Building Air Cleanup System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA fiters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the safety analysis.
ANSI N510-1975 will be used as a procedural guide for surveillance testing.
f RADDA.M NECK                          B3/4 9 3

SECTION 3/4.10 SPECIAL TEST EXCEPTIONS l

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1       The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 or 3.1.1.4 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated rod worth is available for trip insertion from OPERABLE control rod (s).
APPLICABILITY: MODE 2.
ACTION:
: a. With any control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 30 gpm of 14000 rpm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 or 3.1.1.4 is restored,
: b. With all control rods fully inserted and the reactor suberitical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 30 gpm of 14000 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 or 3.1.1.4 is restored.
SURVEILLANCE RE0VIREMENTS 4.10.1.1       The position of each control rod either partially or fully withdrawn shall be determined at least once per 2 hours.
4.10.1.2       Each control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1 or 3.1.1.4.
HADDAM NECK                            3/4 10-1

SPECIAL TEST EXCEPTIONS 3/4.10.2 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.2    The limitations of Specifications 3.1.1.5, 3.1.1.6, 3.1.3.1, 3.1.3.5, 3.1.3.6.1 and 3.1.3.6.2 may be suspended during the performance of PHYSICS TESTS provided:
: a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,
: b. The Reactor Trip Setpoints on the OPERABLE Power Range channels are set at less than or equal to 23% of RATED THERMAL POWER, and
  ,                                             c. The Reactor Coolant System lowest opergting loop temperature                                                                                                                                              t (T,yg) is greater than or equal to 515 F.
APPLICABILITY: MODE 2.
ACTION:
: a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immedittely open the reactor trip breakers.
: b. With a React 5 Coolant System operating loop temperature (T3V9) less than 515 F, restore T                                                                                                                                          to within its limit within 15 minutes or be in at least $69 STANDBY within the next 15 minutes.
SURVEILLANCE RE0VIREMENTS 1
l                                        4.10.2.1    The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.
I                                         4.10.2.2 Each Power Range channel shall be subjected to an ANALOG CHANNEL                                                                                                                                            t OPERATIONAL TEST within 12 hours prior to initiating PHYSICS TESTS.
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            .<                            4.10.2.3    TheReactorCoolantSysgemtemperature(T                                                                                                                                                 shall be determined ;
to be greater   than or equal to 515 F at least once pD9)0                                                                                                                                     3    minutes during   !
PHYSICS TESTS.                                                                                                                                                                                                       l E
j HADDAM NECK                                                                                                                                                              3/4 10-2

SPECIAL TEST EXCEPTIONS 3/4.10.3 POSITION INDICATION SYSTEM - SHUT 00WN LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual shutdown and control rod drop time              ,
measurements provided:
: a. Only one shutdown or control bank is withdrawn from the fully inserted position at a time, and                                  '
: b. The analog rod position indicator is OPERABLE during the withdrawal of the rods.*
APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements.
ACTION:
With the Position Indication Systems inoperable, or more than one bank of rods withdrawn, immediately open the reactor trip breakers.                   :
SURVEILLANCE RE0VIREMENTS 4.10.3     The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours prior to the start of and at least once per 24 hours thereafter during rod drop time measurements by verifying the Bank   ;
Average Analog Rod Position Indication System and the Digital Rod Position Indication System agree within 16 steps.
This requirement is not applicable during the initial calibration of    I the Analog Rod Position Indication System provided: (1) K      is       :
maintainedlessthanorequalto0.94,and(2)onlyonesRNdownor control rod bank is withdrcwn from the fully inserted position at one   l time.                                                                   l f
HADDAM NECK                          3/4 10-3

3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 SHUTD7WN MARGIN This Special Test Fxception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occuring as a result of fuel burnup or fuel cycling operations.
3/4.10.2 PHYSICS TESTS This Special Test Exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER with the RCS T         slightly lower than normally allowed so that the fundamental nuclear chaflEteristics of the core and related instrumentation can be verified. In order for various characteristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications.
For instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not                                    ,
normally be allowed by Specification 3.1.3.6 which in turn may cause the RCS                               '
T      t 38Y91.o  4. fall slightly below the minimum temperature of Specification                                   .
3/4.10.3 POSITION INDICATION SYSTEM-SHUTDOWN
* This Special Test Exception permits the Position Inoication Systems to be inoperable during rod drop tih4 measurements. The exception is required since the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Position Indication Systems remain OPERABLE.
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HADDAM NECK                             B3/4 10-1

6 6
SECTION 3/4.11 RADIDACTIVE EFFLUENTS l
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3/4.11    RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION LM] JING CONDITION FOR OPERATION 3.11.1.1     The concentration of radioactive material released from the site (see Figure 5.1-1) shall not exceed the concentrations specified in 10 CFR Part 20, Appendix B. Table II, Column 2 for radionuclides other than dissolved or entrained noble gases.       For dissoly gases, the concentration shall not exceed 2 x 10'gdmicroci/ml or entrained totalnoble activity.
APPLICABILITY: At all times.
ACTION:
With the concentration of radioactive material released from the site exceeding the above limits, restore the concentration to within the above limits within 15 minutes.
SURVEILLANCE RE0VIREMENTS 4.11.1.1.1      Radioactive liquid wastes shall be sampled and analyzed in accordance with the sampling and analysis program specified in Section I of the REM 00CM.
4.11.1.1.2 The results of the radioactive analysis shall be used in l accordance with the methods of Section 11 of the REH00CM to assure that the concentration of the point of release are maintained within the limits of Specification 3.11.1.1.
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   '4ADDAM NECK                         3/4 11-1

RADIOACTIVE EFFLUENTS DOSE. LIOUIDS
,                                                 LIMITING CONDITION FOR OPERATION                                              f 3.11.1.2       The dose or dose comitment to any MEMBER OF THE PUBLIC from    .
radioactive materials in liquid effluents released from the site (see Figure  "
a 5.1-1) shall be limited
: a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ, j                                                            and                                                                ;
: b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.      i i
APPLICABILITY: At all times.
r ACTION:
: a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and i                                                            submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be
 !                                                           taken to reduce the releases of radioactive materials in liquid
,                                                            effluents during the remainder of the current calendar quarter and j                                                           during the remainder of the calendar year so that the cumulative dose or dose commitment to any MEMBER OF THE PUBLIC from such release during the calendar year is within 3 mrem to the total i                                                            body and 10 mrem to any organ,
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not
.                                                            applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.2.1       Cumulative dose contributions from liquid effluents shall be determined in accordance with Section II of the REM 00CM.
4.11.1.2.2 Relative accuracy or conservatisms of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as octailed in the REMODCM.
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HADDAM NECK                            3/4 11-2 l

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RADIOACTIVE EFFLCENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1      The dose rate, at any time, offsite (see Figure 5.1-1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:                                                   3
: a. The dose rate limit for noble gases shall be less than or equal to 500 mre#yr to the total body and less than or equal to 3000 mrem /yr to the skin, and
: b. The dose rate limit due to inhalation for Iodine-131. Iodine-133, tritium and for all radioactive materials in particulate form with half-lives greater than 8 days shall be less than or equal to 1500    ,
mrem /yr to any organ.
APPLICABILITY: At all times.
ACTION:
With the dose rate (s) exceeding the above limits, decrease the release rate within 15 minutes to comply with the limit (s) given in Specification 3.11.2.1.
SURVEILLANCE RE0VIREMENTS 4.11.2.1.1 The release rate, at any time, of noble gases in gaseous effluents shall be controlled by the offsite dose rate as established above in Specification 3.11.2.1. The corresponding release rate shall be determined in accordance with the methodology of Section II of the REM 00CM.
4.11.2.1.2       The noble gas effluent monitors of Specification 3.3.3.8 shall be used to control release rates to limit offsite doses within the values established in Specification 3.11.2.1.
4.11.2.1.3 The release rate of radioactive materials in gaseous effluents shall be determined by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Section I of the REM 00CM (Table D 1). The corresponding dose rate shall be determined using the methodology and parameters given in the 00CM (Section II of the REM 00CM).
HADDAM NECK                                   3/4 11-3

l RADI0ACTI/E EFFLUENTS DOSE-NOBLC GASES LIMITING CONDITION FOR OPERATION 3.11.2.2    The air dose offsite (see Figure 5.1-1) due to noble gases released in gaseous effluents shall be limited to the following:
: a. During any calendar quarter, to less than or equal to 5 mrad for gama radiation and less thaa or equal to 10 mrad for beta radiation, and
: b. During any calendar year to less than or equal to 10 mrad for gama radiation and less than or equal to 20 mrad for beta radiation.
APPLICABILITY: At all times.
ACTION:
: a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeoing any of the above limits, prepare and submit to the Comission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for and defines the corrective actions to be exceeding    the limit taken to reduce        (s) leases of radioactive noble gases in qascous the re effluents during the remainder of the current calendar quarter and during the remainder of the calendar year so that the cumulative dose during the calendar year is within 10 mrad for gama radiation and 20 mrad for beta radiation,
: b. The provisions of Specifications 3.0.3 :nd 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.2.1 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with Section II of the REM 00CM once every 01 days.
4.11.2.2.2 Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I of the REMODCM.
dADDAM NECK                            3/4 11-4

RADI0 ACTIVE EFFL'JENTS DOSE. RADIOI0 DINES. RADIOACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASEE LIMITING CONDITION FOR OPERATION 3.11.2.3     The dose to any MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium and radioactive materials in particulate form with half lives greater than 8 days in gaseous effluents released offsite (see Figure 5.1-1) shall be limited to the following:
: a. During any calendar quarter to less than or equal to 7.5 mrem to any organ;
: b. During any calendar year to less than or equal to 15 mrem to any organ.
APPLICABILITY: At all times.
ACTION:
: a. With the calculated dose from the release of radionuclides, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce tte releases during the remainder of the current calendar quarter and during the remainder of the calendar year so that the cumulati've dose or dose commitment to any MEMBER OF THE PUBLIC from such releases during the calendar i                                                        year is within 15 mrem to any organ.
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: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS                                              _
l                                             4.11.2.3.1                Cumulative dose contributions for the current calendar quarter and critrent calendar year shall be deterwined in accordance with Section II of the REM 00CM onco every 31 days.
4.11.2.3.2 Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in the REMODCM.
HADOAM NECK                                                  3/4 11-5

RADI0 ACTIVE EFFLUENTS 3/4.11.3                          TOTAL DOSE LIMITING CONDITION FOR OPERATION
;                            3.11.3                         The dose or dose commitment from the site to a MEMBER OF THE PUBLIC is limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem) over a period of 12 consecutive months.
APPLICABILITY: At all times.
ACTION:
: a.                        With the calculated dose from the release of radioactive materitis in liquid or gaseous effluents exceeding twice the limits of Specification 3.11.1.2, 3.11.2.2, or 3.11.2.3, prepare and submit a Special Resort to Commission pursuant to Specification 6.9.2 and limit the suasequent releases such that the dose or dose commi'tment from the site to any MEMBER OF THE PUBLIC is limited to less than or equal to 25 mrem to the total body or any organ (except thyroid, which is limited to less than or equal to 75 mrem) over 12 consecutive months. This Special Report shall         ,
include an analysis which demonstrates that radiation exposures     l from the site to any MEMBER Of THE PUBLIC (including all effluent   [
pathways and direct radiation) are less than the 40 CFR Part 190    i Standard. If the estimated doses exceed the above limits, the       ,
special report shall include a request for a variance in            !
accordance with the provisions of 40 CFR Part 190. Submittal of     i the report is considered a timely request, and a variance is granted until staff action on the request is complete,              j
: b.                        The provisions of Specification 3.0.3 and 3.0.4 are not             i applicable.                                                         !
SURVEILLANCE RE0VIREMENTS c
4.11.3                         Cumulative dose contributions from liquid and gasoous effluents and direct radiation shall be determined in Specifications 4.11.1.2.1,                                  ,
 !                           4.11.2.2.1 and 4.11.2.3.1 and in accordance with Section !! of the REMODCH                          L once per 31 days.                                                                                   [
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i HADDAM NECK                                               3/4 11-6

3/4.11     RADIOACT'VE EFFLUENTS BA3ES 3/4.11.1     LIOUID EFFLUENTS 3/4.11.1.1     CONCENTRATION This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within: (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a HEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
3/4.11.1.2     DOSE This specification is provided to impicment the requirements of Sections
!!.A. I!!.A and IV.A of Aspendix I, 10 CFR Part 50. The Limiting Condition for Operation implements t1e guides ret forth in Section II.A of Appendix I.
The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". The dose calculation methodology and parameters in the REMODCM implement the requirements in Section III.A of Appendix ! that conformance with the guides of Appendix ! is to be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the REM 0DCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man f ca doutine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.
HADDAM NECK                         B3/4 11-1

RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2      GASEOUS EFFLUENTS 3/4.11.2.1      DOSE RATE This specification is provided to ensure that the dose rate at anytime from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for all areas offsite. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the ex)osure of an individual offsite to annual average concentrations exceeding tie limits specified in Appendix B, Table 11 of 10 CFR Part 20 (10 CFR 20.106(b)). For individuals who uay at times be within the site boundary, the occupancy of that individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to less than or equal to 500 mrem / year to the total body or to less than or equal to 3000 mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant to less than or equal to 1500 mrem / year from inhalation and milk pathways.
3/4.11.2.2      DOSE. NORLE GASES This specification is provided to implement the requirements of Sections II.B., Ill. A and IV. A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix 1.
The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix !
assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section Ill.A of Appendix I that conform with the guides of Appendix ! be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases !n gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculational of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50 Appendix 1," Revision 1, October 1977 and Regulatory Guide 1.111, 'Hethods for Estimating Atmospheric Transport and Disperston of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977.
The 00CM equations provided for determining the air doses at the site boundary are based upon utilizing successively more realistic dose calculational methodologies. More realistic dose calculational methods are used whenever simplified calculations indicate a dose approaching a substantial portion of the regulatory limits. The n,ethods used, in order are, previously determined air dose per released activity ratio, historical meteorological data and actual radionuclide mix releastd or real time meteorology and actual radionuclides released.
HADDAM NECK                          B3/4 11-2

RADIOACTIVE EFFLUENTS BASES 3/4.11.2.3     DOSE. RADIOI0 DINES. RADI0 ACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER THAN N0BLE GASES This specification is provided to implement the requirements of Sections II.C. III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix ! that conformance with the guides for Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestinated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials will to be consistent with the methodology provided in Regulatory Guide 1.109, "Calculating of Annual Dose to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix       I," Revision I, October 1977 and Regulatory Guide 1.111. "Methods for Estimating Atmospheric Transport and Dispersien of Gaseous Effluents in Routine Releases from Light Water-Cooled Reactors,"
Revision I, July 1977. The release rate specifications for radiotodines, radioactive material in pa'.*ticulate form and radionucildes other than noble gases are dependent on the existing radionuclide pathways to man. The pathwayr. which are examined in the development of these calculations are:
: 1) individual inhalation of Airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by iaan, and 4) deposition on the ground with subsequent exposure of man.
3/4.11.3     TOTAL DOSE This specification is provided to meet the reporting requirements of 40 CFR Part 190. For the purposes of the Special Report, it may be assumed that the dose commitment to any MEMBER OF THE PUBLIC f om other fuel cycle sources is negiigible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be I considered.
I HADDAM NECK                             B3/4 11-3

i 1
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:!   DESIGN FEATURES i
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5.0 DESIGN FEATURES                                                                                                                           [
T 5.1' SITE EXCLUSION AREA 5.1.1                             The Exclusion Area shall be as shown in Figure 5.1-1.
LOW POPULATION ZONE                                                                                                                            !
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;                                                                   5.1.2                            The Low Population Zone shall be as shown in Figure 5.1-2.                                                   t i
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i HADOAM NECK                                                    5-1 L

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HADDAM NECK                                                                     5-3                                 .
_ - - _ _ _ _ _ _ _ _ _ _ _ _ _                                                                                r   j

DESIGN FEATURES 5.2 CONTAINMENT CONFIGURATIM 5.2.1           The containment building is a steel-lined, reinforced concrete building of cylindrical shape, with a hemispherical dome roof and a flat base, and having the following design features:
: a.        Nominal inside diameter of cylinder - 134 feet - 11.25 inches.
: b.       Nominal inside height of cylinder = 119.5 feet (Not including dome).
: c.       Nominal inside height of containment - 188 feet - 11.5 inches (From containment floor to inside top of dome).
: d.       Minimum thickness of concrete walls = 4.5 feet,
: e.       Minimum thickness of concrete dome - 2.5 feet.
: f.       Minimum thickness of concrete bottom mat - 9 feet.
: g.        Nominal thickness of steel liner - 1/4 to 1/2 inch (bottom - 1/4 inch, spherical dome - 1/2 inch and side wall - 3/8 inch).
6
: h.       Net free volume - 2.232 x 10 cubic feet (nominal).
DESIGN PRESSURE AND TEMPERATURE 5.2.2 Thecontainmentbuildingisdesignedandshallbemaigtainedfora maximum internal pressure of 40 psig and a temperature of 260 F.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1          The core shall contain 157 fuel assemblies with each fuel assembly containing 204 fuel rods clad with Type 304 stainless steel except that up to 4 fuel assemblies may be clad with Zircaloy-4 for experimental purposes.
Each fuel assembly excluding experimental assemblies shall have a nominal active fuel length of 120.5 inches and contain a typical weight of 411.5 kilograms uranium. The core loading shall have a maximum enrichment of 4.00 weight percent U 235.
HADOAM NECK                                           5-4

i 4
l DESIGN FEATURES l
l CONTROL R00 ASSEMBLIES f
5.3.2     The core shall contain 45 full-length control rod assemblies. The                 !
full-length control rod assemblies shall contain a nominal 120.25 inches of                l absorber materini. The nominal values of absorber material shall be 80%                    :
silver, 15% indium, and 5% cadmium.                                                        !
5.4 REACTOR COOLANT SYSTEM 1                                                                                                                                l
  ;                                  DESIGN PRESSURE AND TEMPERATURE l
5.4.1     The Reactor Coolant System is designed and shall be maintained:                  !
: a. In accordance with the Code requirements specified in the FSAR,                 .
:                                                with allowance for normal degradation pursuant to the applicable 1                                                 Surveillance Requirements,                                                     i i                                                                                                                                 :
j                                            b. For a pressure of 2485 psig, and                                               j 0
i                                           c. For                                                                             l 680,a F. temperature of 650 F, except for the pressurizer which is               i j                                    VOLUME                                                                                       !
5.4.2 The total water and steam volume of the Reactor Coolant System is approximately 8780 cubic feet.
5.5 METEOROLOGICAL TOWER LOCATION                                                          [
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.                                     5.5.1     The meteorological tower shall be located as shown on Figure 5.1-1.              j i                                                                                                                                t
;                                     5.6 FUEL STORAGE                                                                           i 4
i CRITICAllTY i
l 5.6.1     The spent fuel storage racks are designed and shall be maintained                f j                                    with i
l                                            a. A nominal 10.75 inch center-to center distance including neutron               [
i                                                 absorber surrounding each assembly to ensure a K          less than or         l
;.                                                equal to 0.95 when flooded with unborated water,'Nd                            !
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: b. A maximum fuel enrichment to be stored in the fuel pool will be a              i i                                                 nominal 4.00 weight percent U-235.                                             l 1
DRAINAGE                                                                                    .
:                                                                                                                                i 1
5.6.2      The spent fuel storage pool is designed and shall be maintained to               !
l                                    prevent inadvertent draining of the pool below elevation 27.0 feet MSL.                     {
:                                                                                                                                i I
j                                    HADOAM NECK                              5-5

DESIGN FEATURES CAPACI1Y 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1168 fuel assemblies.
5.7                       REACTOR VESSEL DESIGN TRANSIENTS 5.7.1                       The reactor vessel design transients are as identified in Table 5.7-1. The transient description followed by the number of design cycles is provided.
HADOAM NECK                                             56

f b
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!  SE 5                                                                                                                                                                                       Table 5.7-1
!                                                                                                                           Reactor Vessel Desian Transients
;j         Transient Descrintion                                                                                                                                                                       Number of Desien Cycles Plant heatup at S 100'F per hour                                                                                                                                                                 150
!          Plant cooldown at i IGO'F per hour                                                                                                                                                               153 1
j          Loss of load accident, reactor scrammed                                                                                                                                                           60
}          by high pressurizer water level
}          Reactor scram, turbine tripout                                                                                                                                                                   300 Loss of flow, one reactor coolant pump                                                                                                                                                            60 Y
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                                                                                                                                                                                                                                 'l J

I i
SECTION 6.0 ADMINISTRATIVE CONTROLS l
l l
I I
l

1 1
6.0 ADMINISTRATIVE CONTROLS l
6.1   RESPONSIBILITY 6.1.1    The Station Superintendent shall be responsible for overall facility operation and shall delegate, in writing, the succession to this responsibility during his absence.
o.2 ORGANIZAT1.C3 521, ONSITE AND OFFSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and.
corporate management, respectively. The onsite ar.d offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.
: a. Lines of authority, responsibility, and communicatio 4 shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Quality Assurance Topical Report.
: b. The Station Superintendent shall be resoonsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.
: c. The Senior Vice President, Nuclear Operations, shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.
: d. The individuals who carry out quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.
6.2.2    FACIllTY STAFF
: a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1;
: b. At least one licensed Operator shall be in the control room when   >
fuel is in the reactor. In addition, while the facility is in MODE 1, 2, 3 or 4, at least one licensed Senior Operator shall be in the control room; HADDAM NECK                              61

ADMINISTRATIVE CONTROLS
: c. An individual qualified in radiation protection procedures
* shall be on site when fuel is in the reactor;
: d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;
: e. A site Fire Brigade of at least five members
* shall be maintained on site at all times. The Fire Brigade shall not include the Shift Supervisor and two other members of the minimum shift crew necessary for safe shutdown of the plant and any personnel required for other essential functions during a fire emergency; and
: f. Administrative procedures shall be developed and implemented to limit the working hours of facility 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).
* The individual qualified in radiation     )rotection procedures and Fire Brigade composition may be less than t1e minimum requirements for a period of time not to exceed 2 hours, in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.
HADDAli NECK                             6-2

TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION EQ1[Il0H              NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1. 2. 3. or i        MODE 5 or 6 S0P                              2                         1 R0P                              2                         1 A0P                              2                         2 STA                              1*                        None Abbreviations:
S0P  -
Licensed Senior Reactor Operator R0P  -
Licensed Reactor Operator A0P  -     Additional Operator STA  -
Shift Technical Advisor The shift crew composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours in ordnr to accommodate unexpected absence of on-duty shift crew members proviried immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1. This provision does not rermit s    any shift crew position to be unmanned upon shift enange due to an onr.oming shift crewman being late or absent.
During any absence of the Shift Supervisor froc,the control roor,while the unit is in MODE 1, 2, 3, or 4, an individual (other than the Shift Technical Advisor) with a valid Senior Operator license shall be designated to assume the control room command function. During any absence of the Shift Supervisor from the control room while the unit is in MODE 5 or 6, an individual with a valid Senior Operator license or Operator license shall be designated to assume the control room command function.
* The STA position can be filled by either of the two Senior Reactor Operators (a dual-role individual), if he meets the requirements of Specification 6.3.1.2.1.
HADDAM NECK                               6-3

ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF OVALIFICATIONS 6.3.1    Each member of the facility staff shall maet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions.
6.3.1.1 The position of Health Physics Supervisor shall meet the following minimum qualifications:
: a. Academic degree in an enginering or science field or equivalent as per Section 6.3.1.1.c.
: b. Minimum of five years professional technical experience in the area of radiological safety, three years of which shall be in applied radiation work in a nuclear facility dealing with problems similar to those encountered in a nuclear power reactor,
: c. Technical experience in the area of radiological safety beyond the five year mirimum may be substituted on a one-for-one basis towards the academic degree requirement (four years of technical experience being equivalent to a four year academic degree).
: d. Academic and technical experience must total a minimum of nine years.
6.3.1.2     The position of Shift Technical Advisor (STA) shall meet the requirements of Specification 6.3.1.2.1 or 6.3.1.2.2.
: 1. Dual-role individual: Must hold a senior reactor operator's license at The Haddam Neck Plant, meet the STA training criteria of NUREG-0737, Item I. A.1.1, and meet one of the following educational alternatives:
: a. Bachelor's degree in engineering from an accredited institution;
: b. Professional Engineer's license obtained by the successful completion of the PE examination;
: c. Bachelor's degree in engineering technology from an accredited institution, including course work in the physical, mathematical, or engineering sciences;
: d. Bachelor's degree in a physical science from an accredited institution, including course work in the physicai, mathematical, or engineering sciences;
: e. Successful completion of the Memphis State University (MSU)
STA program.    (Note:  This alternative is only acceptable for individuals who have completed the program prior to December 31, 1986); or HADOAM NECK                               6-4

ADMINISTRATIVE CONTROLS
: f. Successful completion of the Thames Valley State Technical College associate's degree in Nuclear Engineering Technology program, provided that the individual was enrolled in the program by October 1, 1987.
: 2. Dedicated STA: Must meet the STA training criteria of NUREG-0737, Item I.A.I.1, and have received specific training in plant design, and response and analysis of the plant for transients and accidents.
6.4 TRAINING 6.4.1     A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Coordinator assigned program responsibility and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 10CFR55.59.
6.4.2     A training program for the Fire Brigade shall be maintained under the direction of the Training Department and shall meet or exceed the intent of Section 27 of the NFPA Code-1976, except for Fire Brigade training sessions which shall be held at least quarterly.
6.5 REVIEW AND AUDIT 6.5.1    PLANT OPERATIONS REVIEW COMMITTEE (PORC)
FUNCTION 6.5.1.1     The PORC shall function to advise the Station Superintendent on all matters related to nuclear safety.
COMPOSITION 6.5.1.2     The PORC shall be composed of the:
Chairperson:    Station Superintendent Member:         Unit Superintendent Member:         Operations Supervisor Membar:         Maintenance Supervisor Member:         Instrument and Control Supervisor Member:         Reactor Engineer Member:         Engineering Supervisor Member:         Station Services Superintendent Member:         Quality Services Supervisor Member:         Chemistry Supervisor Member:         Health Physics Supervisor Member:         Security Supervisor HADOAM NECK                              65

ADMINISTRATIVE CONTROLS ALTERNATES 6.5.1.3 All alternate members shall be ap)ointed in writing by the PORC Chairperson to serve on a temporary basis; iowever, no more than two alternates shall participate as voting members in PORC activities at. any one time.
MEETING FRE0VENCY 6.5.1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairperson or his/her designated alternate.
OUORUM 6.5.1.5 The quorum of the PORC shall consist of the Chairperson or his/her designated alternate and four members including alternates.
RESPONSIBILITIES 6.5.1.6    The PORC shall be responsible for:
: a. Review of: (1) all procedures required by Specification 6.8 and changes thereto, and 2) any other proposed procedures or changes thereto as determined by the Station Superintendent to affect nuclear safety;
: b. Review of all proposed tests and experiments that affect n': dear safety;
: c. Review of all proposed changes to the Technical Specifications;
: d. Review of all proposed changes or modifict ions to plant systems or equipment that affect nuclear safety;
: e. Investigation of all violations of the Technical Specifications, including the preparation and forwarding of rep?rts covering evaluation and recommendations to pr m t recurrence, to the Vice President-Nuclear Operations and to . Chairperson of the Nuclear Review Board;
: f. Review of all REPORTABLE EVENTS; i
: g. Review of facility operations to detect potential safety hazards;
: h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairperson of the Nuclear Review Board or the Station Superintendent; HADDAM NECK                             6-6

ADMINISTRATIVE CONTR01f
: 1. Review of the Security Plan and implementing procedures and shall submit recommended changes to the Chairperson of the Nuclear Review Board;
: j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairperson of the Nuclear Review Board; AUTHORITY 6.5.1.7 The PORC shall:
: a. Report to and be advisory to the Station Superintendent on those areas of responsibility specified in Section 6.5.1.6(a) through (j);
: b. Render determinations in writing to the Station Superintendent with regards to whether or r.ot each item considered under Specification 6.5.1.6a. through e., above, constitutes an unreviewed safety question; and
: c. Provide written notification, meeting minutes may be used for this purpose, to the Vice President Nuclear Operations and the Chairperson of the Nuclear Review Board of disagreement between the PORC and the Station Superintendent; however, the Station Superintendent shall have responsibili'.y for resolution of such disagreements pursuant to Specification 6.1.1.
RECORDS 6.5.1.8 The PORC shall maintain written minutes of each meeting that, at a minimum, document the results of all PORC activities performed under the responsibility and authority provisions of these Technical Specificatior.s.
Copies shall be >rovided to the Vice President-Nuclear Operations and Chairperson of t1e Nuclear Review Board.
6.5.2 NUCLEAR REVIEW BOARD (NRB)
OVALIFICATIONS 6.5.2.1    The minimum qualifications of NRB members are as follows:
: a. The Chairman and NRB members shall have:
: 1. an academic degree in engineering or physical science field, or hold a senior management position, and
: 2. a minimum of five years technical experience in their respective field of expertise, and
: 3. a minimum of nine (9) years combined academic and technical experience.
HADDAM NECK                             6-7

1 l
ADMINISTRATIVE CONTROLS                             ,                       i I
: b. The NRB shall collectively have the experience and corapetence required to review activities in the following areas:
: 1. Nuclear power plant operations
: 2. Nuclear engineering
: 3. Chemistry and radiochemistry
: 4. Metallurgy
: 5. Nondestructive testing
: 6. Instrumentation and control
: 7. Radiological safety
: 8. Mechanical and electrical engineering
: 9. Administration
: 10. Quality assurance practices COMPOSITION 6.5.2.2 The NRB shall consist of no less than eight, nor more than eleven members including the Chairperson and the Station Superintendent. The Chairperson and members of the NRB shail be appointed in writing by the Senior Vice President - Nuclear Engineering and Operations.
CONSULTANTS 6.5.2.3 Consultants shall be utilized as determined by the NRB Chairperson to provide expert advice to the NRB.
MEETING FRE0VENCY 6.5.2.4     The NRB shall meet at least once per 6 months.
OV0 RUM 6.5.2.5 The quorum of the NRB necessary for the performance of the NRB review and audit functions of these Technical Specifications shall consist of at least enough members to constitute a majority of the assigned members including the chairman or his designated alternate. No more than a minority of the quorum shall have line responsibility for operation of the facility.
HA00AM NECK                             6-8

ADMINISTRATIVE CONTROLS REVIEW 6.5.2.6   The NRB shall function as an independent review body and shall review:
: a. The safety evaluations for: (1) changes to precedures, equipment, or systems; and (2) tests or experiments completed under the provisions of 10 CFR 50.59, to verify that such actions did not constitute ar. unreviewed safety question;
: b. Proposed changes to procedures, equipment, or systems which involve an unreviewed safety question as defined in 10 CFR 50.59;
: c. Proposed tests cr experiments which involve an unreviewed safety question as defined in 10 CFR 50.59;
: d. Proposed changes in these Technical Specifications or this Operating License;
: e. Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance;
: f. Significant operating abnormalities or deviations from normal and 3xpected performance of plant equipment that affect nuclear safety;
: g. All REPORTABLE EVENTS;
: h. Indications of a significant unanticipated deficiency, affecting nuclear safety, in some aspect of design or operation of structures, systems, or components;
: i. Reports and meeting minutes of the PORC; AUDITS 6.5.2.7 Audits of facility activities shall be performed under the cogniunce of the NRB. These audits shall encompass:
: a. The conformance of unit operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months;
: b. The performance, training, and qualifications of the facility staff at least once per 12 months;
: c. The resdts of actions taken to correct deficiencies occurring in facility equipment, structures, systems, or method of operation that affect nuclear safety, at least once per 6 months; HADOAM NECK                            6-9

ADMINISTRATIVE CONTROLS
: d. The performance of activities required by the Quality Assurance Program to meet the criteria of Appendix B,10 CFR Part 50, at least once per 24 months;
: e. Any other area of facility operation considered appropriate by the NRB or the Senior Vice President - Nuclear Engineering and Operations;
: f. The Fire Protection Program and implementing procedures at least once per 24 months;
: g. The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant. An outside independent fire protection consultant shall be used at least every third year.
: h. The performance of activities in accordance with the RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL and PROCESS CONTROL PROGRAM at least once per 24 months; and
: i. The performance of activities required by the quality controls section of Regulatory Guide 1.21, Rev. 1, June 1974 and 4.1, Rev.
1 April 1975. at least once per 12 months.
J. The Facility Emergency Plan anti implementing procedures at least once per 24 months;
: k. The Facility Security Plan and implementing procedures at least once per 24 months; AUTHORITY 6.5.2.8 The NRB shall report to and advise the Senior Vice President -
Nuclear Engineering and Operations on those areas of responsibility specified in Specifications 6.5.2.6 and 6.5.2.7. Heeting minutes may be used for this purpose.
RECORDS 6.5.2.9 Records of NRB activities shall be prepared, approved and distributed as indicated below:
: a. Minutes of each NRB meeting shall be prepared, approved, and forwarded to the Senior Vice President Nuclear Engineering and Operations within 14 days following each meeting; HADDAM NECK                            6 10

ADMINISTRATIVE CONTROLS l
: b. Reports of reviews encompassed by Specification 6.5.2.6 shall be   l prepared, approved, and forwarded to the Senior Vice President-Nuclear Engineering and Operations within 14 days following completion of the review. Meeting minutes may be used for this purpose; and
: c. Audit reports encompassed by Specification 6.5.2.7 above, shall be forwarded to the Senior Vice President-Nuclear Engineering and Operations and to the management positions responsible for the areas audited within 30 days after completion of the audit.
Meeting minutes may be used for this purpose.
6.6 REPORTABLE EVENT ACTION 6.6.1    The following actions shall be taken for REPORTABLE EVENTS:
: a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10CFR Part 50, and
: b. Each REPORTABLE EVENT shall be reviewed by the POR' and the results of this review shall be submitted to the NRB and the Vice President-Nuclear Operations.
6.7    SAFETY LIMIT VIOLATION 6.7.1    The following actions shall be taken in the event a Safety Limit is violated:
: a. The unit shall be placed in at least HOT STANDCY within one hour.
: b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within I hour. The Vice President-Nuclear Operations and the Chairman of the NRB shall be notified within 24 hours;
: c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PORC. This report shall describe:
(1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence;
: d. The Safety Limit Violation Resort shall be submitted to the Commission, the Chairman of tie NRB and the Vice President-Nuclear Operations within 14 days of the violation;
: e. Operation will not be resumed until authorized by the Commission.
j 6.8 PROCEDURES AND PROGRAMS 6.8.1    Written procedures and/or administrative policies shall be established, implemented, and maintained covering the activities referenced below:
1 HADDAM NECK                              6-l?
I 1

ADMINISTRATIVE CONTROLS
: a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978;
: b. The requirements and recommendations of Sections 5.1 and 5.3 of ANSI N 18.7-1976.
: c. Fire Protection Program implementation.
: d. Quality controls for effluent monitoring, us'.ag the guidance in Regulatory Guide 1.21 Rev.1, June 1974.
l         e. RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION j               MANUAL (REM 00CM) implementation except for Section I.E,
[               Radiological Environmental Monitoring.
: f. PROCESS CONTROL PROGrtAM implementation.
6.8.2        Each procedure of Specification 6.8.1, and changes thereto, shall be reviewed by the PORC and shall be approved by the Station Superintendent i
prior to implementation and reviewed periodically as set forth in each document or in administrative procedures.
6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made provided:
: a. The intent of the original procedure is not altered;
: b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the unit affected; and
: c. The change is documented, reviewed by the PORC and approved by the Station Superintendent within 14 days of implementation.
6.8.4        Written procedures shall be established, implemented and maintained covering Section I.E., Radiological Environmental Monitoring, of the REM 03CH.
6.8.5 All procedures and procedure changes required for the Radiological Environmental Monitoring Program of Specification 6.8.4 above shall be reviewed by an individual (other than the author) from the Radiological Assessment Branch te the Production Operation Services Laboratory (POSL) and approved by appropriate supervision.
Temporary changes may be made provided the intent of the original procedure is not altered and the change is documented and reviewed by an individual (other than the author) from the Radiologica! Assessment Branch or the POSL, within 14 days of implementation.
HADDAM NECK                               6-12

                                                                              )
ADMINISTRATIVE CONTROLS                                                        l 6.9 REPORTING RE0VIRQ' Bill                                    ,
ROUTINE REPORTS l
6.9.1    In addition to the applicable reporting requirements of Title 10,     l Code of Federal Regulations, the following reports shall be submitted to the U. S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C.
20555, one copy to the appropriate Regional Office of the NRC, and one copy to the appropriate NRC Resident Inspector, unless otherwise noted.
STARTUP REPORT 6.9.1.1    A summary report of plant startup and power escalation testing shall be submitted following: (1) amendment to the license involving a planned increase in power level, (2) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (3) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.
6.9.1.2 The Startup Report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the op2 rating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions b. sed on other commitments shall be included in this report.
6.9.1.3 Startup Reports shall be submitted within: (1) 90 days following comnietion of the Startup Test Program, (2) 90 days following resumption or cemencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation supplementary rept.ts shall be submitted at least every 3 months until all),
three events have been completed.
ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the facility as described below for the previous calendar year shall be submitted prior to March 1 of each year.
HADDAM NECK                           6-13

ADMINISTPATIVE CONTROLS 6.9.1.5    Reports required on an annual basis shall include:
: a. A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions * (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing and refueling) . The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for.
In the aggregate, at least 80% of the total whole-body dose received from external sources shall be assigned to specific major work functions;
: b. Documentation of all challenges to the pressurizer power-operated relief valves (PORVs) and safety valves; and
: c. The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit.
Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.
ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.1.6 Routine Annual Radiological Environmental Operating Reports covering the operation of the facility during the previous calendar year shall be submitted prior to May 1 of each year.
The Annual Radiological Environmental Operating Reports shall include that information delineated in the REMODCH.
* This tabulation supplements the requirements of Section 20.407 of 10 CFR Part 20.
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HADDAM NECK                            6-14 l

ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELFfSE REPORT 4
6.9.1.7 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the facility during the previous 6 months of operation shall be submitted withir. 60 days after January 1 and July 1 of each year.
A supplemental report containing dose assessments for the previous year shall be submitted annually within 90 days after January 1.
The report shall include that information delineated in the REMODCH.
Any changes to the REMODCM shall be submitted in the Semiannual Radioactiva Effluent Release Report.
MONTHLY OPERATING REPORTS 6.9.1.8 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, with one copy to the appropriate Regional Office of the NRC, and one copy to the appropriate NRC Resident Inspector, no later than the 15th of each month following the calendar month covered by the report.
SPECIAL REPORTS 6.9.2     Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C.      20555, with one copy to the appropriate Regional Office of the NRC, and one copy to the appropriate NRC Resident Inspector, within the time period specified for each report.
6.10     RECORD RETENTION 6.10.1     In addition to the appiicable record retention recuirements of Title 10, Code of Feueral Regulations, the following recorcs shall be retained for at least the minimum period indicated.
6.10.2     The following records shall be retained for at least 5 years:
: a. Records and logs of facility operation covering time interval at each power level;
: b. Records and logs of principal maintenance activities, inspections, repair, and repiscement of principal items of equipment related to nuclear safety;
: c. All REPORTABLE EVENTS;
: d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications; HADDAM NECK                            6-15 e

ADMINISTRATIVE CONTROLS
: e. Records of reactor tests and experiments; 4
: f. Records of changes made to the procedures required by Specification 6.8.1;                                                 '
: g. Records of radioactive shipments;
: h. Records of sealed source and ff ssion detector leak tests and results;
: i. Records of annual physical inventory of all sealed source material of record; and 6.10.3    The following records shall be retained for the duration of the facility Operating License:
: a. Record and drawing changes reflecting facility design modifications made to systems ara equipment described in the FSAR;
: b. Records of new anu irradiated fuel inventory, fuel transfers, and assembly burnup histories;
: c. Records of facility radiation and contamination surveys;
: d. Records of radiation exposure for all individuals entering radiation control areas;
: e. Records of gaseous and liquid radioactive material released to the environs;
: f. Records of transient or operational cycles for those facility components identified in Table 5.7-1;
: g. Records of training and qualification for current members of the facility staff;
: h. Records of inservice inspections performed pursuant to these Technical Specifications;
: i. Records of quality assurance activities required by the Quality 1             Assurance Manual not listed in Specification 6.10.2;
: j. Records of reviews parformed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR Part 50.59;
: k. Records of meetings of the PORC and the NRB; 1   Records for Environmental Qualification.
l HADDAM NE M                           6-16

ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM                                             t 6.11.1    Procedures for personnel radiation protection shall be prepared consistent with the requirement of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation    4 exposure.                                                                      1 5J 2 HIGH RADIATION AREA 6.12.1    Pursuant to paragraph 20.203(c)(5) of 10 CFR Part 20, in lieu of the "control device" or "alarm signal" required by paragra)h 20.203(c), each high radiation area, as defined in 10 CFR Part 20, in whic1 the intensity of radiation is greater than 100 mR/h but less than 1000 mR/h at 45 cm (18 in.)
from the radiation source or from any surface which the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied hv one or more of the following:
: a. A radiation monitor 99 device which continuously indicates the radiation dose rate in the area; or
: b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or
: c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive centrol over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Health Physics Supervisor in the RWP.
6.12.2    In addition to the requirements of Specification 6.12.1, areas accest,ible to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control ef the shift supervisor on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of +he stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) cnntinuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activitits being performed within the area.
HADOAM NECK                            6 17

1 1
I ADMINISTRATIVE CONTROLS For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and where no enclosure can be raasonably constructed around the individual area, that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device or continually guarded.
6.13 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REM 0DCM)
Section I, Radiological Effluents Manitoring Manual, shall outline the sampling and analysis programs to determine the concentration of radioactive materials released offsite as well as dose commitments to individuals in those exposure pathways and for those radionuclides released as a result of facility operation.      It shall also specify operating guidelines for RADI0 ACTIVE WASTE TREATMENT SYSTEMS and report content.      Changes to Section I shall be submitted to the Commission for approval prior to implementation.
Section II, the Offsite Dose Calculation Manual, shall describe the methodology and parameters to be us2d in the calculation of offsite doses due to radioactive gaseous and liquid affluents and in the calculations of gaseous and liquid effluent monitoring instrumentation Alarm / Trip Setpoints consistent with the applicable LCO's contained in these Technical Specifications. Changes to Section 11 need not be submitted to the Commission for approval prior to implementation, but shall be included in the next Semiannual Radioactive Effluent Release Report.
6.14 RADI0 ACTIVE WASTE TREATHEt(I Prncedures for liquid and gaseous radioactive effluent discharges from the facility shall be prepared, approved, maintained anu adhered to for all operations involving offsite relea:es of radioactive effluents.      These procedures shall specify the use of appropriate RADI0 ACTIVE WASTE TREATMENT SYSTEMS utilizing the guidance provided in the REM 00CH.
The Solid RADI0 ACTIVE WASTE TREATMENT SYSTEM shall % operated in accordance with the PROCESS CONTROL PROGRAM to process wet raoloactive wastes to meet shipping and burial ground requirements.
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HADDAM NECK                              6-18

ADMINISTRATIVE CONTROLS                    __
6.15 SYSTEMS INTEGRITY The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:
: a. Provisions establishing preventive maintenance and periodic visual inspection requirements, and
: b. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle fqtervals.
6.16 PASS /Samolina and Analysis of Plant Effluents The licensee shall implement and maintain a program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. This program shall include the following:
: a. Training of personnel
: b. Procedures for sampling and analysis, and
: c. Provisions for maintenance of sampling and analysis equipment.
HADOAM NECK                           6-19

Docket No. 50-213 B13056 Attachment 2 Haddam deck Plant Description of Individual Froposed Changes to Technical Specifications and Discussion on
.           Significant Hazards Consideration l
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l October 1988

Attachment 2 Page 1 l
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l Haddam Neck Plant Technical Specifications Section 1.0, Definitions l
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Attachment 2 Page 2 Technical Soecification Section 1.0. Definitions Descriotion of the Proposed Revised Technical Soecifications The Definitions section in the proposed revised Technical Specifications (to be referred to as the "RTS") for the Haddam Neck Plant consists of only definitions applied throughout the Technical Specification. No limiting conditions for operation exist in the revised Technical Specifications, therefore, no specific significant hazards consideration is included below.
Rather, discussion of any additional terms or existing defined terms with significantly different definition:: will be contained in the evaluations of the section they appear in, if warranted.
The following defined terms have been added to the revised Technical Specifications:
o    Average Disintegration Energy o    End of Core Life o    Linear Heat Generation Rate o    Nuclear Enthalpy Rise Hot Channel Factor o    RHR Loop o    Staggered Test Basis o    Trip Actuating Device Operational Test The definitions of the following existing terms have been modified as described below:
o     Identified Leakage - This term has been changed to include leakage through the steam generator to secondary coolant system and is consistent with the Westinghouse Standard Technical Specification, l                    Draft Revision 5 (to be referred to as H STS).
i               o    Containment     Integrity  -
The revised Technical Specifications contain conditions on containment leakage rates and penetration sealing mechanism not addressed in the Definitions section of the existing Technical Specifications. Also, the existing Technical Specifications provide for operability exceptions to allow for surveillance testing on certain containment isolation valves when containment integrity is required. This is now provided for in the    revised Technical Specification section       on  containment integrity rather than the "Definitions" section, o    Channel      Calibration,    Channel  Check,  Core   Alteration and Unidentified Leakage - These definition terms have been further clarified in the RTS to be consistent with the W STS.

1 Attachment 2 Page 3 Haddam Neck Plant Technical Specifications Section 3/4.2, Power Distribution Limits i
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1 Page 4 Technical Specification Section 3/4.2. Power Distribution Limiti Description of the Procosed Revised Technical Soecifications The Power Distribution Limits section (3/4.2) in the proposed revised Technical Specifications (to be referred to as the RTS) for the Haddam Neck Plant is the same as the existing Technical Specification Section 3.17. The existing  Technical Specification Section 3.17 has pregipusly been reviewed and approved by the Nuclear Regulatory Commission.             The power range neutron flux-high trip setpoints provided in the ACTION statements of the proposed RTS Section     3.2.3.1, 3.2.3. 2 and 3.2.4 do not include the 2%
drift. The setpoints in the existig Technical Specifications and the allowable value in the proposed RTS Section 2.2 (which includ2s the drift allowance) are equivalent. Therefore, this distinction in the proposed RTS is not considered a change relative to the existing Technical Specifications. In addition, the RTS are an enhancement to the existing Technical Specifications. They provide clear applicability, action, and surveillance requirements modeled after the W STS.
Sianificant Hazards Consideratiort in accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration.
The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because the changes would not:
: 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The content of RTS is the same as the previously approved version of the existing Technical Specification Section 3.17. No changes have occurred to this section other than renumbering to achieve consistency with the W STS. Therefore, there is no   increase  in the probability or consequences of an           accident previously analyzed.
: 2. Create the possibility of a new or different kind of accident from any previously evaluated. The proposed changes do not impact the operation of any component or system. The proposed changes do not introduce any now failures. Therefore, the proposed changes do not create the pos::ibility of a new or different kind of accident from those previously analyzed.
: 3. Involve a significant reduction in a margin of safety.           Since the proposed changes do not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety.
(1)   F. H. Akstulewicz letter to E. J. Hroczka, Technical Specifications for Cycle 15 Operation, dated November 12, 1987.

Attachment 2 Page 5 Haddam Neck Plant Technical Specifications Section 3/4.9, Refueling Operations 7   .

Page 6 Techr:ical Soecification Section 3/4.9. Refuelina Operations The proposed revised Technical Specification (RTS) Section 3/4.9 has been prepared by converting the existing Technical Specification Section 3.13, Refueling using the M STS. The proposed RTS Section 3/4.9 deals with refueling operations. The purpose of this specification is to assure that:
o   The reactor core remains in a safe, subcritical condition.
o   Adequate core cooling is maintained.
o   The radiological impact of postulated accidents is acceptable, o   The refueling equipment can be operated with adequate safety factors.
Descriotion of the Prooosed Revised Technical Soecifications (RTS) 2/4.9.1 Boron Concentration 3.9.1      Limitina Condition for Ooeration (LCO)
This LCO is comparable to item D           ? the existing Technical Specification Section 3.13. It requires the boron concentration of all filled unisolated portions of the Reacto. Coolant System (RCS) and the refueling canal to be maintained uniform and sufficient to ensure a K,ff of 0.94 or less.
3.9.1      Action The proposed RTS requires that no core alterations or reactivity additions be made and that the RCS be borated until the LC0 is met. The existing Technical Specification Section 3.13 does not contain any action statement.
4.9.1.1    Surveillance Reautrements                                              i and 4.9.1.2    The proposed surveillance requirement 12 more restrictive than the     l current weekly surveillance. The proposed surveillance requires verification of the LCO prior to removing or unbolting the head or withdrawal of any control rod in excess of three feet. Subsequent verification is required at least once per 72 hours.
3/4.9.2    Instrumentation 3.9.2      Limitina Condition for Ooeration (LCO)
The proposed RTS requires the instrumentation to be operable whenever core elterations or positive reactivity changes are made.
The existing Technical Specification requires this only when core alterations are made, not necessarily during other positive reactivity charges. This represents an additional restriction in the propo',cd RTS. The existing Technical Specification requires

Attachment 2 Page 7 an audible alarm from both neucru, ,'itors but neither are required to be operating.      The proposeu RTS requires one to be operating evithout a second necessarily being available.
3.9.2     Action The proposed RTS imposes an action statement as compared to the existing Technical Specifications (which have none). This action statement requires that the boron concentration of the RCS be determined every 12 hours if the LC0 is not met.                   ;
4.9.2     Surveillance Reauirements The proposed RTS includes surveillance requirements for the Source Range Neutron Flux Monitor whereas the existing Technical Specification does not have any surveillance requirement.
3/4.9.3   Decay Time i
3.9.3     Limitina Condition for Ooeration i
The proposed RTS includes a new requirement compared to the existing Technical Specifications (which have none). The LCO requires the reactor to be subcritical for at least 100 hours      ,
i                                                                  prior to movement of irradiated fuel in the reactor vessel. These restrictions are presently included in current station procedures. -
 ;                                                        3.9.3     Action                                                             i This is a new requirement that requires suspension of irradiated fuel movement if the LC0 is not met.
4.9.3     Eurveillance Reouirements This is a new requirement that nquires verification of the LCO.
i                                                         3/4.9.4   Containment Buildina Penetradons 3.9.4     Upitino Condition for Ooeratio-The proposed RTS imposes a a% ve tirement for containment integrity durit g core alterations or movement of irradiated fuel.
Requirements are impcsed n- the equipment hatch, tM airlock, and each penetration providing direct accen from the containment atmosphere to the uutsidt atmosphere.
3.9.4    Action The  proposed RTS imposes a new requiremcnt to suspend operation involving core alteration or movement of irradiated fuel if the LCO is not met.

Attachment 2 Page 8 4.9.4           Surveillance Reauirements This is a new requirement that requires verification of the LCO.
3/4.9.5         Communications 3.9.5           Limitina Ccndition for Ooeration i
The proposed RTS LCO is equivalent in intent to the existing Technical Specification Section 3.13G.
3.9.5           Action The action statement in the proposed RTS requires suspending all operations involving core alterations or movement of irradiated fuel whenever the LC0 is not met.
4.9.5           Surveillance Reauirements Hie      surveillance included in the proposed RTS requires demonstration of direct communication between these two areas at                                         ,
prescribed intervals prior to and during core alterations.
3/4.9.6         Manioulator Crane 3.9.6           Limitina Condition for Ooeration                                                                         ,
The proposed RTS imposes a new requirement as compared to the                                            ;
e.tisting Technical Specifications (which have none). Operability requirements are stated for the manipulator crane and auxiliary hoist. Minimum capacity as well as overload cut-off limit for the manipulator crane is specified.
3.9.6           Action With the LCO not met, suspension of the use of the crane or hoist                                        !
for movement of control rod drive shafts or fuel assemblies within the vessel is required.
Surveillance Recuirements                                                                                E 4.9.6.1 and                                                                                                                 i 4.9.6.2         It is required io demonstrate operability within 100 hours prior                                         !
to the start of movement of fuel assemblies or control rod drive                                         '
shafts. This testing adds a margin of safety and is consistent (45 the W STS.                                                                                           ;
t M7              a ine 1,* v p nt Fuel Storaae Pool Buildina A.9.-             ,.
                         +
t on for Operation                                                                 [
:                              $         n's prevents loads, other than the fuel assemblies,                                 !
a>           1650 pounds from travel over the storage pool. The pr ,             -
also removes the ability to move spent fuel casks                                   '
ov .r tt .         A- sn a limited basis as is allowed by the existing Technicai Spec.           ations.

r Attachment 2 Page 9 3.9.7                         Acli.gn With the LCO not met, placing the crane load in a safe condition is required.
4.9.7                         Surveillance Reautrements Administrative control regarding the movement of loads in the        i spent fuel storage pool building are required to be adhered to by all personnel involved with liftir.g a load in excess of 1650 pounds.
3/4.9,8 Residual Heat Removal and Coolant Circulation 3.9.8.1                       Hioh Water level                                                     ,
3.9.8.1                       Limitina Condition for Ooeration 1                                                                The proposed RTS LCO is equivalent to the existing Technical         ;
Specification Section 3.13.C(2).                                     ;
3.9.8.1                       Action The    proposed RTS action staterent requires initiation          of t corrective action to return the RHR system to operable status.
4.8.8.1                       Surveillance Reautrements                                            !
The proposed RTS im, nose a new surveillance requirement as compared ;
to the existing Technical Specifications (which have none).      The operability of the RHR loop and adequacy of flow are verified.
3.9.8.2                       Low Water level 3.9.8.2                       Limitina Candition for Ooeration                                     ,
An additional requirement to have one of the two operable RHR loops in operation is included in the proposed RTS.
3.9.8.2                       Action An additional requirement for the situation of when no RHR loops are in cperation has been added to the proposed RTS.
4.9.8.2                       Surveillance reouirements The proposed RTS imposes new surveillance requirements as compared to the existing Technical S)ecifications. The operability of the RHR loop in operation and acequacy of flow is verified. Also, the operability of the RHR loop not in operation is verified.
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L Attachment 2 Page 10 3/4.9.9   Containment Purae and Exhaust Isolation System The addition of th.s new requiremer.t into the proposed RTS ensures that the assumption made in the design basis calculation for the radiological consequences of the fuel handling accidents will be                                            i met. The operability of this system ensures that the containment vent and purge penetrations can be isolated upon detection of high radiation levels within the containment.
3/4.9.10 Water level - Reactor Vessel                                                                                 ,
r 3.9.10    Limitina Condition for Ooeration l
The proposed RTS LCO requirement is esse. Wily equivalent to the                                            ,
existing Technical Specification Section 3../.C(l).
3.9.10    Action The proposed RTS specifies that any fuel assembly or control rod                                            -
,             in transit should be placed in a safe location.                   Overall, the                              l intent of this action is the same as that within the existing                                               l Technical Specifications, 4.9.10    Surveillance Reautrements                                                                                   f The proposed RTS imposes a new requirement as compared to the existing Technical Specification.               It requires water level to be verified periodically.
i l   3/4.9.11 Water level - Storace Pool The proposed RTS imposes a new requirement as compared to the
,             existing Technical Specifications. The restrictions on minimum water level (21 feet) ensures that sufficient water depth is
,             available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The
,             minimum water depth is consistent with the assumption of the                                                :
safety analysis.                                                                                            !
il                                                                                                                        t 3/4.9.12 fuel Storaae Buildina Air Cleanuo System
                                                                                                                         \
!             The proposed RTS imposes a new requirement as crapared to the                                               '
;             existing Technical Specifications. The fuel storage building air                                            ,
4             cleanup system is required to be operable during movement of fuel                                           '
within the storage pool or crane operation with loads over the                                              !
storage pools.         This ensures that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and chLrcoal adsorber prior to discharge to the atmosphere.                                                                                                 ;
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Attachment 2 Page 11 Sionificant Hazards Consideration In accordance with 10CFR50.92, CYAPCO has reviewed the proposed RTS and has concluded that they do not involve a significant hazards consideratior.             The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compremised. The proposed RTS do not involve a significant hazards consideration because the changes would not:
: 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The determination of whether or not a proposed change is equivalent, more restrictive (or a new requirement),
or less restrictive is based on the Limiting Condition for Operation (LCO) and Applicability Requirements since it is these requirements which will impact the design basis accidents,             in general, the conversion to W STS yields more extensive and/or restrictive Action and Surveillance Requirements. The comparison of the proposed RTS, based on LCO     and    Applicability, relative to the existing Technical Specification is provided below:
o    Equivalent 3.9.1, Boron Concentration
                                                 -     3.9.5, Communications 3.9.8.1, Residual Heat Removal and Coolant Circulation (High WaterLevel) 3.9.10 Water Level - Reactor Vessel o    More Restrictive (or new)
                                                 -     3.9.2, Instrumentation 3.9.3, Decay time 3.9.4, Containment Building Penetration, 3.9.6, Manipulator Crane 3.9.7, Crane Travel - Spent Fuel Storage Pool Building 3.9.8.2, Residual Heat Removal and Coolant Circulation (Low WaterLevel) 3.9.9, Containment Purge and Exhaust Isolation System 3.9.11, Water level - Storage Pool 3.9.12, Fuel Storage Building Air Cleanup System o    less Restri.tive
                                                 -     Deletion of the existing Technical Specification Section 3.13(A), Radiation Monitoring Deletion of the existing Technical Specification Section 3.13(F), Verification of Subcriticality Cases where the proposed RTS are equivalent or more restrictive than the existing Technical Specifications are considered acceptable whereas less restrictive requirements need justification. As described above, most of the changes are more restrictive in that there are no comparable requirements in the existing Technical Specifications. For the few changes that are less restrictive, justification is provided                                          l

Attachment 2 Page 12 for the change.      The specifications listed as "More Restrictive (or New)" above, have no negative impact on the design basis accidents.
These proposed changes assure that the consequences of a fuel candling accident stay within analyzed bounds. The proposed RTS SectiON 3.9.3, 3.9.4, 3.9.6, 3.9.9, 3.9.11, and 3.9.12 impose restrictions on fuel decay time, containment buildinq penetrations, the manipulator crane, the containment purge and exhaust isolation system, spent fuel pool water level,      and the fuel    storage building air cleanup system, respectively, assuring the assumptions made in performing radiological i                                       consequence analyses are not violated. Section 3.9.8.2 is more restrictive because it requires one of the two operable RHR loops to be in operation as compared to the existing Technical          Specification requirement to have two RHR loops operable. Section 3.9.7 is more restrictive since it removes the ability to move spent fuel casks over 1                                        the pool on a limited basis as is currently allowed by the existing Technical Specifications. Section 3.9.2 is more restrictive since an addie onal restriction is proposed by requiring instrumentation to be operable not only when core alterations are made but also when positive j                                        reactivity changes are made, l                                       The proposed deletion of existing Technical Spr,cification Section l                                       3.13(F) could potentially affect the detection of an inadvertent criticality. However, adequate maintenance and monitoring of the core reactivity condition is assured by the refueling boron concentration requirements (the proposed RTS 3.9.1) and via the audible indication of l                                       neutron flux (the proposed RTS 3.9.2).          The deletion of Section 3.13(A), which requires continuous radiation monitoring during fuel movement, does not mitigate the consequences of any accident and is more appropriately located in plant operating procedures rather than Technical Specifications.
I                                        Based upon the above discussion, the proposed RTS will not increase the probability or consequences of any accident previously analyzed.
I                                  2. Create the possibility of a new or different kind of accident from any l                                        previously evaluated. Because there are no hardware modifications associated with the proposed changes, the performance of safety related i                                        systems    remains unaffected during operations.        The   operability requir m nt.s are increased over the current requirements thus enhancing
)
the performance of safety systems. Therefore, the proposed RTS will not modify the plant response to the point where it can be considered a j!                                       new accident nor are any credible failure modes created.
;                                  3. Involve a significant reduction in a margin of safety.       Because the i
changes proposed herein provide acceptable results for the design basis accidents, no additional burden will be placed on the protective a                                        boundaries for postelated accidents. In addition, there are no plant hardware modifications associated with this change and hence, there is j                                        no direct impact on the protective boundaries. The proposed RTS do not affect the safety limits of the protective boundaries and the bases of j                                        the proposed RTS have been modified to reflect the proposed changes.
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Page 13 Haddtm Neck Plant Technical Specifications Section 3/4.10 Special Test Exceptions

Attachment 2 Page 14 1                          Technical Soecification Sectiot 3/4.10. Soecial Test Exceptions
[                           Descriotion of the Procosed Revised Technical Soecifications The Special Test Exceptions Section (3/4.10) in the proposed revised                                                                               ;
;                Technical Specifications (to be referred to as RTS) for the Haddam Neck                                                                           .
4 Plant is the same as the existing Technical Specification Section 3.24. The j                existing Technical S                                            3.24 has 1.reviously been reviewed and                                            l approved by the NRC.gification                           The power range neutron flux-high trip setpoint                                          j
.                provided in the ACTION statement of the proposed RTS Section 3.10.2 does not                                                                       !
,                include the 2% drift.                                      The set)oints in the existing Technical 4
Specifications and the allowable va' ue in the proposed RTS Section 2.2 (which includes the drift allowance) are equivalent.                                           Therefore, this                                    .
distinction in the proposed RTS is not considered a change relative to the                                                                        [
i               existing Technical Specifications. In addition, the proposed RTS are an                                                                           i j                enhancement to the existing Technical Specifications. They provide clear applicability, action and surveillance requirements modeled after the H STS.
j                Sionificant Hazards Consideration l
i J
in accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and                                                                      :
has concluded that they do not involve a significant hazards consideration.                                                                      -
!                The basis for this conclusion is that the three criteria of 10CFR50.92(c)                                                                        :
are not compromised. The pro)osed changes do not involve a significant                                                                           .
hazards consideration because t1e changes would not:                                                                                            '
: l. Involve a significant increase in the proDability or consequences of an                                                                    t accident previously evaluated. The content of RTS is the same as the previously approved version of the existing Technical Specification                                                                        :
Secti M 3.24. No changes have occurred to this section other than the                                                                      ;
renumbering to achieve consistency with the W STS. Therefore, there is no increase in the probability or consequences of an accident previously analyzed.
: 2. Create the possibility of a new or different hind of accident from any previously evaluated. The proposed changes do not impact the operation                                                                     i of any component or system. The proposed changes do not introduce any                                                                      l new failures. Therefore, the proposed char.ges do not create the possibility of a new or different kind of accident from those                                                                              ;
previously analyzed.
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: 3. Involve a significant reduction in a margin of safety.                                           Since the                                 I proposed changes do not affect the consequences of an accident                                                                             l previously analyzed, there is no reduction in the margin of safety.                                                                        l i
t (1)   F. M. Anstulewicz letter to E. J. Hroczka, "Technical Specifications                                                                       I for Cycle 15 Operation," dated November 12, 1987.                                                                                          l l
t

Attachme:.t 2 Page 15 Haddam Neck Plant Technical Specifications Section 3/4,11 Radioactive Effluents

Attachment 2                                                                         '
Page 16 Technical Soecification Section 3/4.11. Radioactive Effluents i
Descriotion of the Prooosed Revised Technical Soecifications:
I i
in the proposed revised Technical
  ;                             The   Radioactive Specificatit,ns  (to beEffluents referred toSection as RTS) (3/4.11) for the Haddam Neck Plant is th same as the existing Technical Specification Section 7/8.1. The existing Technica1g' )ification   7/8.1 has previously been reviewed and approved by 3                               the NRC.            No changes have occurred to this section other than             !
renumbering to achieve consistency with the W STS.      The proposed RTS are an     !
:                               enhancement to the existing Technical Specifications. They provide clear            !
applicability, action and surveillance requirements modeled after the W STS.        i Stanificant Hazards Consideration                                                   l t
l                               !n accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration.         ;
;                               The basis for this conclusion is that the three criteria of 10CFR50.92(c)           '
i                               are not compromised. The prososed changes do not involve a significant              (
i                              hazards consideration because t1e changes would not.
:                                                                                                                   i 1
: 1. Involve a significant increase in the probability or consequences of an       '
,                                    accident previously evaluated. The content of RTS is the same as the i                                    previously approved version of the existing Technical Specification            i Section 7/8.1. No changes have occurred to this section other than the         '
j                                    renumbering to achieve consistency with the .W. STS. Therefore, there is no increase in probability or consequences of an accident previously           l 1                                    analyzed.                                                                      ,
l                                                                                                                   !
!                               2. Create the possibility of a new or different kind of accident from any
!                                    previously evaluated. The proposed changes do not impact the operation I                                    of any component or system. The proposed changes do not introduce any          !
new failures.      Therefore, the proposed changes do not create the           j i                                    possibility of a new or different kind of accident from those                  i i                                   previously analyzed.                                                           !
i
{                               3. Involve a significant reduction in a margin of safety.          Since the i                                    proposed changes do not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety, i
l I
t i
(1)  J. A. Zwolinski letter to J. F. Opeka, "Radiological Effluent Technical        !
Specifications (RETs)," dated September 5, 1985.                               l (2)  F. M. Akstulewicz letter to J. F. Opeka, "Administrative and Reporting Requirement Technical Specification Changes," dated August 6, 1986.
I Page 17 Haddam Neck Plant Technical Specifications Section 5.0, Design Features i
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                                          -)

o                                                                                                                                                                :
Attachrent 2 Page 18                                                                                                                                   ;
Technical Soecification Section 5.0. Desian Features Descriotion of the Proposed Revised Technical Soecifications                                                                    j
.                                                                                                                                                                  i
!                        The pro)osed revised Technical Specification (RTS) Section 5.0 has been preparet by converting the existing Technical Specification 5.0,                                                                  Design ;
Features, using the i STS.         The proposed RTS Section 5.0 provides                                                                 i information concerning the plant's design featurer.                                                                                      !
Section 5.1. Sjlg                                                                                                                        l The proposed RTS replaces the existing Technical Specification Section 5.2.                                                              [
The purpose of this section is to describe the site as used in radiological                                                              -
consequences calculations as the distance to Exclusion Area Boundary (EAB)                                                               l and low Population Zone have a major impact on calculated doses. The proposed RTS provides more detail (e.g., distance to EAB in each section)                                                                l than the existing Technical Specifications. The proposed RTS deletes the                                                                 >
requirement that no part of the site shall be sold or leased and all the
;                        structures are to be owned by Connecticut Yankee.                                                                  This deletion has no  i l                        effect on the safety of the plant.
i
;                        Section 5.2/ Containment                                                                                                                 ,
;                        The proposed RTS Section 5.2 replaces the existing Technical Specification                                                               ;
i                        Section 5.4. The proposed RTS provides additional dimensions for the                                                                     j l                        containment building and the maximum temperature for which the containment                                                               ,
building is designed. The information included in the proposed RTS is                                                                ,
censistent with the design basis accident analyses and the recently updated                                                              i j                        FSAR. The proposed RTS deletes reference to materials used in construction,                                                            [
stress levels and penetrations.        Moreover, this information is already                                                             ;
included in the updated FSAR. 1herefore, this deletion has no effect on the                                                              l 4                        consequences of the design basis accident analyses.                                                                                      ,
i Section 5.3. Reactor Core                                                                                                                t J                                                                                                                                                                 l The proposed RTS provides less information concerning the overall core                                                                   !
dimensions, core material content, and excess reactivity control. Also, the                                                              !
proposed RTS ocletes the existing Technical Specifications condition that up                                                             l to two target rods of Np02 - Zr02 pellets clad with stainless steel may be                                                               !
inserted. Since changes in these design features are always evaluated for                                                                !
impact on the design basis analysis, removal from the design features                                                                    t section of the Technical Specification is acceptable.                                                                                    :
The proposed RTS however, provides more detail concerning the dimensions and                                                             !
the material content of the control rods. Parameters introduced are W STS                                                                i rcquirements and the format is consistent with the M STS.                                                                                l The change does not affect any reactor core parameters or systems and does not impact the consequences of any design basis accident.                                                                                i I

Attachment 2 Page ;9 Section 5.4. Reactor Coolant Systtm The requirements in the proposed RTS are equ ' talent to the existing Technical   Specifications and the W STS.        The proposed RTS adds the pressurizer    design   temperature   of 680'F.       The   existing Technical Specification specifies the RCS volume as approximately 8800 cubic feet.
Instead, the proposed RTS specifies the RCS volume as approximately 8780 cubic feet. The new number is equivalent to the nominal value of 8782 cubic feet. As such, the cheage does not have an impact on any design basis accident.
Section 5.5. Meteoroloaical Tower location t
The change is an additional requirement and is aquivalent to the W STS.      It provides the location of the meteorological tower.
Section 5.6. Fuel Storaae The proposed RTS Section 5.6 is an additional requirement to the existing Technical Specification and is equivalent to the W STS.             The design parameters specified for the spent fuel storage pool are consistent with the    :
updated FSAR. The minimum level required to be maintained is 27 feet MSL which is consistent with the plant design.                                      ,
Section 5.7. Reactor Vessel _p.gsion Transients This is an additionel requirement to make the RTS consistent with W STS and the proposed RTS provides design transient limits for the reactor vessel.
The design transients which have been considered in the reactor vessel analytical reports are conservative for the actual transients which were addre: sed.
;                                                                                                        r Based upon the discussion above, it is concluded that there is no adverse impact on the design basis accidents due to the changes.
;                        S_lanificant Hazards Consideration in accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration.
The basis for this conclusion is that the ttree criteris of 10CFR50.92(c) are not compi nised. The proposed changes do not involve a significant hazards consideration because the changes would not:
: 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. As stated earlier, the pro)osed RTS merely converts the existing Technical Specification for tie design        l features specified in Section 5.0 into W STS format. The proposed changes can be considered as an administrative change and do not           !
include any hardware changes. Therefore, these proposed changes do not have any impact on the design basis accidents.
: 2. Create the possibility of a new or different kind of accident from any previously analyzed. Since there are no changes in the way the plant is operated, the potential for an unanalyzed accident is not created.
No new failure modes are introduced.

Attachment 2 Page 20
: 3. Involve a significant reduction in a margin of safety. The changes do not have any impact on the consequences of an accident previously analyzed, therefore there is no reduction in a margin of safety.
l l
l

Attachment 2 Page 21 l
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l l
Haddam Neck Plant                                          l Technical Specifications Section 6.0, Administrative Controls l
l

i                                                                                                                         ?
Attachment 2 Page 22           .
{
4 Itshairal Soecification Section 6.0. Administrative Controls                           '
Descriotion of the Proposed Revised Technical Soecificationi                           -
2 i                           The pro)osed revised Technical Specification               Section 6.0 has been preparec by converting the existing Technical (RTS)   Specification Section 6.0, i                           Administrative Controls, using the W STS. The proposed changes generally consist of editorial or format type changes.         These changes do not affect                 l intent. Major changes from the existing Technical Specifications are described below.                                                                                !
Section 6.2. Oraanization l                          The organization charts (Figures 6.2-1 and 6.2 2 of the existing Technical                      l are   not  included    in    the  proposed    RTS. Instead,               t 4
Specifications)has Section 6.2.1          been changed to add requirements that capture the
  ,                          essential aspects of the organizational structuro that are defined by the                       '
4 existing onsite and offsite organization charts. Appendix B to 10CFR50 and                      t 10CFR50.54(a)(3) govern changes to organization Cescribed in the Quality                        !-
;                            Assurance Program. The QA Topical Report contains the organization charts                       ;
i                            for all of Northeast Utilities nuclear facilities.         Since the QA Topical                 !
1                            Report is updated on an annual basis, it is unnecessary to amend its                            ;
Technical Specifications for each organizational change. The proposed                           i
  ,                          change is also consistent with the recommendations of Generic Letter (GL)                       l l                            88-06.                                                                                          ;
Section 6.3. Facility Staff Oualifications                                                      i The   proposed   RTS    Section 6.3.1.2. replaces the existing Technical                        l
: j.                           Specification    Section 6.3.1.2.      The    proposed RTS provides explicit                    r
 ;                           qualification requirements for the Shift Technical Advisor (STA) while the
* l existing Technical Specification is more generil.                                               >
i
;                            Section 6.5.2. Nuclear Review Board i
'                            In letters dated April 29 and July 21, 1988          CYAPCO submitted a proposed                  .
,                            revision to the existing Technical Specifications which clarifies the j                            qualifications, function, and quorum requirements for the Nuclear Review                        !
!                            Board. The proposed RTS section is consistent with the changes propost:d in                  .
L                            thcse letters. The significant hazards consideration discussion included in                     [
j                            the above referenced letters is also applicable to the proposed RTS.                            !
Sionificant Hazards Consideration                                                               :
!                                                                                                                            I In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and                     ,
has concluded that they do not involve a significant hazards consideration.                     !
The basis for this conclusion is that the three criteria of 10CFR50.92(c)                       !
i                           are not compromised. The proposed changes do not involve a significant                          i j                            hazards consideration because the changes would nott
[
j                                                                                                                            ,
i                            1. Involve a significant increase in the probability or consequences of an                   i
!                                 accident previously evaluated. Deletion of the organization charts                         !
!                                  from the Technical Specifications does not affect plant operation. As                     [
in the past, the NRC will continue to be informed of organizational I

j                                                   Attachment 2 Page 23 changes through other required regulatory controls. These changes are also consistent with the recomendations of GL 88 06. Therefore, the pro >osed changes will not involve a significant increase in the pro) ability or consequences of an accident previously evaluated. The remaining proposed RTS sections are either consistent with the existing Technical Specification or involve minor administrative changes that do not have any impact on the design basis accidents. Therefore, the proposed RTS will not involve a significant increase in the probability or consequences of an accilent previously evaluated.
: 2.           Create the possibility of a new or different kind of accident from any previ m ly evaluated. The proposed changes are administrative in nature and no physical alterations of plant configuration or changes to setpoints or opercting          parameters are proposed. Therefore,   the proposed changes do not create the possibility of a new or different kir.d of accident from those previously analyzed.
: 3.          Involve a significant reduction in a margin of safety. CYAPCO, through its quality assurance programs, its commitment to maintain only qualified personnel in positions of responsibility, and other require <J control, assures that safety functions will be performed at a high level of competence. Therefore, removal of the organization charta from the Technical Specifications will not affect the margin of safety.
Also, the remaining proposed changes do not affect the consequences of an accident previously analyzed, therefore there is no reduction in the margin of safety.

Docket No. 50-213 B13056 Attachment 3 Technical Specification Ccmparison Matrix 1
l 1
l                         October 1988 l
l i
L _

<                                                                                                         i Attachment 3 Page 1 TECHNICAL $PECIFICATION COMPARIS0N NATRIX                                             ,
and EXISTING T. 5. CROS5-REFERENCE Introduction The Technical    Specification Comparison Matrix              was prepared to facilitate the revision of the existing Haddam NeckTechnical (TSCM)Specifications (T.S.). The TSCM is set up denoting the proposed revised Technical Specification (to be referred to as RTS) section numbers in the left hand                               ,
column followed by a short description.          The next column lists the                              '
corresponding existing T.S. section number. The final two columns compare the requirements contained in the proposed section with the existing T.S.                               -
and the Westinghouse STS, respectively. The key at the bottom of each page provides an explanation for the symbols located in *he two comparison columns. The equivalent notation "E" may either denote that exact wording                              '
has been transposed from the existing T.S. or different verbage conveying equivalent requirements has been used. In many cases, there was not a                                   ,
one for-one relationship, but rather multi-sectinn relationships whereas the requirements in a given T.S. section may be divided between several different sections in the RTS. The additional requirement notation "++"
denotes that the RTS is more restrictive because it is an entirely new requirement as compared to the existing T.S. or it is more restrictive in                               '
the sense that the existing T.S. requirements have been changed such that                               j they are more restrictive.                                                                              ;
l e

Attachment 3 Page 2 TECHNICAL SPECIFICATION COMPAr.ISON MATRIX

l                                                                      '
Page 3                                                                           j l
Haddam Neck Plant Proposed Revised Technical Specifications Section 1.0 Comparison

Attachme;t 3 Page 4 TECHNICAL SPECIFICATION COMPARISON MATRIX Comparison  Comparison Existing   With Existing   Withif T.S.#         Descriotion                   T.S. #        T.S.        STS 1.0           Definitions 1.1           Action                       1.26           E            E 1.2           Analog Chan Operational Test 1.29           E            E 1.3           Axial Offset                 1.24           E            -
1.4           Channel Calibration          1.9            E            E 1.5           Channel Check                1.10           E            E 1.6           Containment Integrity        1.8            ++(1)        E 1.7           Controlled Leakage           1.17           E            E 1.8           Core Alteration              1.12           E            E 1.9           Dose Equivalent I-131        1.30           E            E 1.10          Avg. Disintegration Energy    -             ++           D 1.11          End of-core Life              -
                                                          ++           ++
1.12          Frequency Notation           1.21           E            E 1.13          Identified Leakage           1.14           ++(2)        E 1.14          LHGR                          -
                                                          ++           ++
1.15          Members of the Pubile        1.31           E            E 1.16          Nuclear Enthalpy Rise Hot     -             ++           i; Channel Factor 1.17          Operable                     1.32           E            E 1.18          Operatir.ial Mode            1.4            E            ++(3) 1.19          Physics Tests                1.25           E            E 1.20          Pressure Boundary Leakage    1.16           E            E 1.21          Purge                        1.33           E            E 1.22          Quadrant Power Tilt Ratio    1.18           E            D 1.23          Radioactive Waste Treatment 1.34            E            ++
System 1.24          REH0DCH                      1.35           E            ++
1.25          Rated Thermal Power          1.3            E            E 1.26          Reportable Event             1.7            E            E 1.27          RHR Loop                      -             ++           ++
1.28          thutdown Margin              1.13           E            E 1,29            1te Boundary               1.36           E            E 1.30          Source Check                 1.37           E            E 1.31          Staggered Test Basis          -             ++           E 1,32          Thermal Power                1.2            E            E 1.33          Trip Actuating Device         -             ++           E Operational Test 1.34          Unidentified Leakage         1.15           E            E 1.35          Venting                      1,39           E            E
++ = Additional requirements E - Equivalent Requirements D - Different Requirements

Attachment 3 Page 5 l
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l 1
Haddam Neck Plant Proposed Revised Technit.al Specifications Section 3/4.2 Comparison

Attachment 3 l      Page 6 l
TECHNICAL SPECIFICATION l                                       COMPARISON MATRIX 1
l Comparison  Comparison Existing   With Existing   With M T.S.#         Descriotion                     T.S. #        T.S.         STS 3.2.1.1       Axial Offset - 4 loop          3.17.1.1        E            -(4)
Applicability                  3.17.1.1        E            *(5)
Action                         3.17.1.1        E            -
4.2.1.1.1     Determination of Axial         3.17.1.1        E            -
Offset 4.2.1.1.2     Verification of excore/        3.17.1.1        E            -
incore correlation 4.2.1.1.3     Determination of excore/       3.17.1.1        E            -
incore 4.2.1.1.4     Calibration of Power Range 3.17.1.1            E            -
detectors 3.2.1.2       Axial Offset - 3 loop          3.17.1.2        E            -(4)
Applicability                  3.17.1.2        E            *(5)
Action                         3.17.1.2        E            -
4.2.1.2.1     Determination of Axial         3.17.1.2        E            -
Offset 4.2.1.2.2     Verification of excore/        3.17.1.2        E            -
incore correlation 4.2.1.2.3     Determination of excore/       3.17.1.2        E            -
incore correlation 4.2.1.2.4     Calibration of Power Range 3.17.1.2            E            -
detectors 3.2.2.1        LHGR-4 loop                   3.17.2.1        E             -(6)
Applicability                  3.17.2.1        E             -
Action                         3.17.2.1        E             -
4.2.2.1.1      Power Distribution            3.17.2.1        E             -
Measurement 4.2.2.1.2      LHGR Allowances               3.17.2.1        E             -
3.2.2.2        LHGR - 3 loop                 3.17.2.2        E             -(6)
Applicability                  3.17.2.2        E             -
Action                        3.17.2.2        E             -
4.2.2.2.1      Power Distribution            3.17.2.2        E             -
Heasurement 4.2.2.2.2      LHGR allowances               3.17.2.2        E             -
   ++ = Additional requirements E = Equivalent requirements
   * = Less restrictive requirements

Attachment 3 Page 7 TECHNICAL SPECIFICATIQH COMPARISDN MATRIX Comparison   Comparison Existing  With Existing   With M lifil                                 Descrintion                           T.S. #       T.S.         STS 3.2.3.1                               Nuc. Hot Chan. Factor                3.17.3.1      E             (8)
Applicability                        3.17.3.1      E Action                               3.17.3.1      E 4.2.3.1.1                             Sgec.4.04notapplicable               3.17.3.1      E             E 4.2.3.1.2                             F AH Determination                   3.17.3.1      E             E 3.2.3.2                               Nuc. Hot Chan. Factor -              3.17.3.2      E             (8) 3 loop Applicability                        3.17.3.2      E Action                               3.17.3.2      E
;                              4.2.3.2.1                            Sgec. 4.04 not applicable                 -        ++            E je                             4.2.3.2.2                            F AH Determination                   3.17.3.2      E             E i                             3.2.4                                 Quad Power Tilt                      ?, .17. 4     E             E 4                                                                   Applicability                        3.17.4        E             E Action                               3.17.4        E             E ll                             4.2.4.1                              Excore/Incore Correlation            3.17.4        E             +
4.2.4.2                              Calculating Ratio                    3.17.4        E             E 4.2.4.3                              Use of Moveable Incore               3.17.4        E             *(9)
Getector system i'                             3.2.5                               lDN8 Pairameters                      3.17.5        E Applis: ability                      3.17.5        E             E Action                               3.17.5        E             E 4.2.5.1                              Vert'/ication of Parameters 3.17.5                 E             E 4.2.5.2                              RCS total flow                       3.17.5        E             E(7) 4.2.5.3                              Char.nel Calibration of              3.17.5        E             ++
flowrote indicator 4
i t
                               ++ = Additional requirements
:                              E - Equivalent requirements 4                             * = Less restrictive requirements
   < -h y- - - --        --.
                                  -ew             ,,--wr-,9mmr_._c           ,-,,-.----,w e rw ,

f Attachme@t 3 Page 8 Haddam Neck Plant Proposed Revised Technical Specifications Section 3/4.9 Comparison

Attachment 3                                                                 l Page 9                                                                       I TECHNICAL SPECIFICATION COMPARISON MATRIX Co:nparison Comparison Existing   With Existing   With W   '
Idd          Descriotion                   T.S. #         T.S.        STS 3.9.1        Boron Cor.centratien         3.13.D          E            D Applicability                3.13            E            E Ae',lon                        -             ++           E 4.9.1.1      'Atermined prior to            -             ++           E 4.9.1.2      Once per 72 hours              -             ++           E 3.9.2         Instrumentation             3.13.B          E(15)        E Applicability                3.13            E            E Action                         -             ++           E 4.9.2        Demonstrated Operable          -             ++           E 3.9.3        Decay Time                     -
                                                             ++(12)       E Applicability                  -
                                                             ++           E Action                         -             ++           E 4.9.3        Verification                   -             ++           E 3.9.4        Cont. Bldg. Penetration        -             ++           D Applicability                  -             ++           E Action                         -
                                                             ++           E 4.9.4        Determination                  -             ++           D
'  3.9.5        Communications               3.13.G          E            E Applicability                3.13.G          E            E Action                         -             ++           E 4.9.5       Determination                   -             ++           E 3.9.6        Manipulator Crane              -
H(14)        D(10)
Applicability                  -             ++           E Action                         -             ++           E 4.9.6.1      Crane                          -
                                                             ++           E 4.9.6.2      Auriliary hol                  -             ++           E 3.9.7        Crane Travel                 3.13.H&I        ++(11)       E Applicability                  -             ++           E Action                         -             ++           E 4.9.7        Adm. Comrd s                   -             ++           0 3.9.8.1      RHR & Ccolant Cire.          3.13.C(2)       E            E Applicability                 3.13            E            E Action                       J13.C(2)        E            E 4.9.8.1      Verification                   -             ++           E E = Equivalent requirements D = Different requirements
   ++ = Additional requirements
   * = less restrictive requirements

Attachment 3 Page 10 4
TECHNICAL SPECIFICATION COMPARISON MATRIX Comparison                  Comparison Existing   With Existing                   With W T.S.#        Descriotion                    T.S. #       T.S.                         STS
  , 3.9.8.2      Low Water Level              3.13.C(l)      ++(16)                        E Applicability                3.13            E                            E Action                       3.13.C(l)       E                            E 4.9.8.2.1    Verify RHR Loop                 -           ++                            E 4.9.8.2.2    Verify Breaker Alignment        -           ++                            ++
3.9.9        Cont. Purge Iso. Sys.           -
                                                             ++(12)                        E Applicability                   -           ++                            E Action                          -            ++                           0 4.9.9        Verified Operable               -           ++                            E 4
3.9.10       Water level - RX Vessel      3.13.C(l)      E                             E Applicability                3.13           E                             E Action                       3.13.C(l)       E                            E 4.9.10       Determination                   -
                                                             ++(12)                        E 3.9.11      Water Level - Storage Pool    -
                                                             ++(12)                        D Applicability                    -           ++                            E Action                           -           ++                            E i   4.9.11       Determination                   -
                                                             ++                            E 3.9.12       Fuel Stor. Air Cleanup          -
                                                             ++(12)                        D
:              Applicability                    -           ++                            0 Action                           -           ++                            E 4.9.12       Demonstrated Operable           -
                                                             ++(13)                        D 1
E = Equivalent requirements i   D = Different requirements i   ++ = Additional requirements

Attachment 3 Page 11 Haddam Neck Plant Proposed Revised Technical Specifications Section 3/4.10 Comparison f
a

Attachment 3 Page 12 ILQiNICAL SPECIFICQLQN
                                     .COMPARICON MATRIX Comparison   Comparison Existing  With Existing   With M LL8          Descriotion                     T.S. 8       T.S.         STS 3.10,.1       Shutdrwn Margin               3.24.1        E             E Applicability                  3.24.1        E             E Action                         3.24.1        E             E 4.10.1.1     Control Rod Position           3.24.1        E             E 4.10.1.2     Demonstrated Capability        3.24.1        E             C 3.10.2        Physics Test                  3.24.2        E             D Applicability                  3.24.2        E             E Action                         3.24.2        E             E i 4.10.2 1     Thermal Power                  3.24.2        E             E 4.10.2.2      Power Range Channel           3.24.2        E             E 4.10.2.3      RCS System Temperature        3.24.2        E             E I 3.10.3        Position Indication System 3.24.3           E             E
                - Shutdown Applicability                  3.24.3        E             E Action                         3.24.3        E             E 3.10.3        Determination                 3.24.3        E             D I
l 1
J E    Equivalent Requirements D - Different Requirements

Attachme@t 3 Page 13 Haddam Neck Plant Proposed Revised Technical Specifications Section 3/4.11 Comparison l
i I'

Attachment 3 Page 14                                                                                                  '
TECHNICAL SPECIFICATION
 .                                    COMPARISON MATRIX                                                         ,
Comparison        Comparison             [
Existing             With Existing          With W W            Descrintiog                  T.S. f                          T.S.        STS l
3.11.1.1     Liquid Effluent Conc.       7.1.1.1                           E           E Applicability         ,     7.1.1.1                           E           E                   l Action                      7.1.1.1                           E           D                  ,
4.11.1.1.1 Sampled and Analyzed          8.1.1.1.1                         E           E                  !
 . 4.11.1.1.2 Assure Limits                 8.1.1.1.2                         E           E                   !
s 3.11.1.2     Dose Liquid                 7.1.1.2                           E           E                  l Applicability               7.1.1.2                           E           E Action                      7.1.1.2                           E           D                   i n    4.11.1.2.1 Determination                 8.1.1.2.1                         E           E
:    4.11.1.2.2 Confirmation                  8.1.1.2.2                         E           ++                  ;
3.11.2.1     Dose Rate - Gas             7.1.2.1                           E           E                  i Applicability               7.1.2.1                           E           E                  f Action                      7.1.2.1                           E           D 4.11.2.1.1 Determination                 8.1.2.1.1                         E           E
;    4.11.2.1.2 Control of Release Rates      8.1.2.1.2                         E           D 4.11.2.1.3 Release Rate of I-131,        0.1.2.1.3                         E           D 2
etc.
3 3.11.2.2     Dose Noble Gas              7.1.2.2                           E           E
!                 Applicability               7.1.2.2                           E           E                  !
Action                      7.1.2.2                           E           E                  .
4.11.2.2.1 Cum. Dose Contributions       8.1.2.2.1                         E           E                   -
4.11.2.2.2 Confirmation                  8.1.2.2.2                         E           ++                -!
3.11.2.3     Dose - Iodine               7.1.2.3                           E           E                   !
Applicability               7.1.2.3                           E           E                  i Action                      7.1.2.3                           E           E                  i 4.11.2.3.1 Cum. Dose Contributions       8.1.2.3.1                         E           E                   i 4.11.2.3.2 Confirmation                  8.1.2.3.2                         E           ++
3.11.3       Total Dose                  7.1.3                             E           E Applicability               7.1.3                             E           E Action                      7.1.3                             E           E                  <
4.11.3       Determination               8.1.3                             E           E                  i I
_                                                                                     s E - Equivalent requirements                                                                                '
D - Different requirements                                                                                !
     ++ = Additional requirements l
i l
i

Attachment 3 Pag] 15 Haddam Neck Plant Proposed Revised Technical Specifications Section 5.0 Comparison

Attachment 3 Page 16 TECHNICAL SPECIFICATION COMPARISON MATRIX Comparison            Comparison Existing- With Existing              With W W                         Description                       T.S. f          T.S.                 STS 5.1                       Site 5.1.1                     Exclusion Area                   5.2        *(17)++(18)                 E 5.1.2                     Low Population                       -      ++                          E 5.2                       Containment 5.2.1                     Configuration                    5.4        *(19)++(20)                 E 5.2.2                     Design Press &
Temp                           5.4 Basis ++(20)                       E 5.3                       Reactor Core 5.3.1                     Fuel Assemblies                  5.3.A      *(21)++(22)                 E-5.3.2                     Control Rod Ass.                 5.3.A.5    ++(23)                      E 5.4                       Reactor Coolant Sys.
5.4.1                     Design Press &                   5.3.Bl&2   E                           E Temp 5.4.2                     Volume                           5.3.B.3    E                           E 5.5                       Meteorological Tower Location                       -
                                                                                            ++                          E 5.6                       fuel Storage 5.6.1                     Criticality                          -      ++                          E 5.6.2                     Drainage                             -      ++                          E 5.6.3                     Capacity                             -      ++                          E 5.7                       Component Cycle limits                             -      ++                          *(24)
If9121 E = Equivalent Requirements
                      +1    - Additional Requirements
                      * = less Restrictive Requirements

Page 17 Haddam Neck Plant Proposed Revised Technical Specifications Section 6.0 Comparison

r     Attachment 3 Page 18 TECHNICAL SPECIFICATION COMPARISON MATRIX Comparison   Comparison Existing    With Existing   With W T.S.8       Descrioliqo                   T.S. #_        T.S.         STS 6.1         Responsibility               6.1.1           E 6.2         Organization 6.2.1       Onsite and Offsite           6.2.1 and       0(25)         0(25)
Organization              6.2.2 6.2.2       Facility Staff               6.2.2           E             E 6.3         Facility Staff               6.3             E(26)         D Qualifications 6.4         Training                     6.4             E(27)         E 6.5         Review and Audit 6.5.1       PORC 6.5.1.1     Function                     6.5.1.1         E             E 6.5.1.2     Composition                  6.5.1.2         E             E 6.5.1.3     Alternates                   6.5.1.3         E             E 6.5.1.4     Heeting Frequency            6.5.1.4         E             E 6.5.1.5     Quorum                       6.5.1.5         E             E 6.5.1.6     Responsibilities             6.5.1 6 &       E             D 6.6.1.b 6.5.1.7     Authority                    6.5.1.7         E             E 6.5.1.8     Records                      6.5.1.8         E             E 6.5.2       NRB 6.5.2.1     Function                     6.5.2.1         E(28)         1:
6.5.2.2     r.omposition                 6.5.2.2         E(28)         -"
6.5.2.3     Consultants                  6.5.2.3         E             E 6.5.2.4     Meeting Frequency            6.5.2.4         E             E 6.5.2.5     Quorum                       6.5.2.5         E(28)         E 6.5.2.6     Review                       6.5.2.6         E             E 6.5.2.7     Audits                       6.5.2.7 Adm & E               +(29) 6.5.2.7 6.5.2.8     Authority                    6.5.2.8         E             E                       i 6.5.2.9     Records                      6.5.2.9         E             E 6.6         Reportable Event Action      6.6             E             E 6.7         Safety Limit Violation       6.7             E(30)         0 6.8         Procedures & Programs        6.8 & 6.8 Adm E               *(31)                    ,
6.9         Reporting Requirements 6.9.1       Routine Reports 6.9.1.1     Startup Reports              6.9.1.a         E             E 6.9.1.4     Annual Reports               6.9.1.c         ++(32)        E 6.9.1.6     Radiological Envir. Oper. 6.9.1.f Adm     E             E 6.9.1.7     Radioactive Eff. Release     6.9.1.g Adm     E             E                       .
Monthly Operating Reports    6.9.1.d 6.9.1.8                                                  E             E 6.9.2       Special Reports              6.9.2           E             ++
i

m Attachment 3 Page 19 TECHNICAL SPECIFICATION COMPARISON MATRIX Comparison Comparison Existing   With Existing       With W luluf        Descriotion                    T.S. #    _
T.S.                     STS 6.10          Record Retention             6.10        ++(33)                                  D 6.11          Radiation Protection Prog. 6.11           E                                       E 6.12         High Radiation Area           6.13        ++(34)                                   E 6.13          REM 00CM                     6.17         E                                      D 6.14          Radioactive Waste Treatst. 6.18         E                                      D 6.15          Systems Integrity            6.15         E                                       E 6.16          PASS / Sampling and Analysis 6.16        +(35)                                    E of Plant Effluents 1
+
l j
   ++ = Additional requirements E = Equivalent requirements j  D    Different requirements
;  * = less Restrictive requirements i
l

Attachment 3 Page 20 Notes to Technical Specification Comoarison Matrjx

Attachment 3 Page 21
                                                   . Notes to Technical Specifications Comparison Matrix (1)                 The RTS contains conditions on containment leakage rates and penetration sealing mechanism not addressed in the Definition Section of existing Technical Specifications. Also, the existing Technical Specifications provides for operability exceptions to allow for surveillance testing on certain containment isolation valves when containment integrity is required. This is now provided for in the RTS section on containment integrity rather than the "Definitions" section.
(2)                 Identified leakage in RTS includes leakage through a steam generator to Secondary Coolant System while existing Technical Specifications do not include this provision.
(3)                 Mode 6 in the RTS has a lower reactivit.< requirement than Mode 6 in the W STS.
(4)                 The W STS employs Axial Flux difference., while the RTS employs Axial Offset.
(5)                 The W STS is applicable for 15% Power and above, while the RTS uses 40% Power.
(6)                 The W STS for heat flux hot channel factor monitors Fq rather than kw/ft and is not comparable to the RTS.
(7)                 The flow requirement in the y STS is included with the Nuclear Enthalpy Rise Hot Channel Factor Specification.
(8)                 The M STS is combined with the RCS flowrate specification, and is not comparable with the RTS.
(9)                 The determination of the ratio under given conditions is proposed for once per 24 hours vs. once per 12 hours in the W STS.
(10)                The W STS calls for the use of a load indicator to prevent lifting excessive drive rod loads while the RTS does not require this.
(11)                The RTS removes the ability to move spent fuel casks over the pool on a limited basis as is currently allowed in existing Technical Specifications.
(12)                The addition of these requirements into the RTS ensures that the assumptions made in the design basis calculation for the radiological consequences of the fuel handling accidents will be met.
(13)                Current station procedures impose less specti'c controls for the testing of fuel building filters.

1 l,    Attachment 3 Page 22 (14)         These requirements are attainable with existing equipment. The proposed surveillance is not currently being done, but will provide an additional margin of safety by detecting problems prior to fuel handling.
(15)         The existing Technical Specifications requires an audible alarm from both neutron monitors but neither are required to be operating. The                                                                                s RTS require one to be operating without a second necessarily being available. The intent of the RTS remains equivalent to the existing Technical Specifications (16)         The RTS require at least one RHR LOOP be in operation while er.isting Technical Specifications only requires RHR LOOPS be operable.                                                                                      s (17)         The RTS deletes th< requirement that no part of the site shall be sold or leased and all structures are to be owned by cc necticut Yankee.
(18)         The RTS provides more detail (e.g., distance to Exclusion Area Boundary in each section) than current Technical Specification.
(19)         The RTS dJletes reference to materials used in construction, stress levels and penetrations.
(20)         The proposed RTS provides additional dimensions for the containment building and the maximum temperature for which the containment building is designed.
(21)          Less detail concerning the overall core dimensions, core material content, and excess reactivity control are provided in the proposed RTS.
(22)         The proposed RTS deletes the current Technical Specifications condition that up to 2 target rods of NP2                                                                             0 Zr02 pellets clad with stainless steel may be inserted.
(23)         The proposed RTS provides more detail concerning the dimensions and material content of the control rods.
(24)         The proposed RTS Section 5.7 provides design transient limits for the reactor vessel only.
(25)         The organization charts (Figures 6.21 and 6.2-2 of the existing Technical Specifications) have been deleted. This is consistent with Generic letter 88 06.
(26)         The proposed RTS provides explicit qualification requirements for the STA while the existing Technical Specification is more general.
(27)         The proposed RTS refers to the more recent regulation                                                                                    (i.e.,
10CFR55.59 instead of Appendix A to 10CFR55).

Attachment 3 Page 23 (28)        The proposed Ri$ sections are consistent with the proposed itcense amendment request submitted to the NRC via letters dated April 29, 1988 and July 21, 1988.
(29)        The proposed RTS requires NRB cognizance for audit activities on the Facility Emergency Plan and Secitrity Plan while the W STS does not impose this requirement.
(30)        The proposed RTS Section 6.7.1.e has been added to be consistent with the NRC's regulations.
(31)        The   W   STS  requires    written             procedures                             covering additional activities.
(32)        Additional reports (i.e., documentation of all challenges to PORVs ar.d safety valves and results of specific activity analysis in which primary coolant exceeds specified limits) are specified in the proposed RTS.
(33)        Tne proposed RTS adds the requirement to maintain records of fission detector leak tests and results.
(34)        The proposed RTS provides additional clarification on the control of a"eas where dose rates exceeds 1000 mr/hr.
(35)        The proposed RTS includes an additional requirement to implement and maintain a Post Accident Sampling System.

Docket No. 50 213 B13056 Attachment 4 Commitments to NRC October 1988

Page k of Il Lo==1Leen1 Lie h*G Date of                                                                                             STS 1*11er                                          Letter  Svidett                            Co m itment                                            Section 8      Comet W. G. Counsti to D. M. Crutchfield &            5/13/81 Rad 6olog6(al Consequences of      To include provisions for esse and testing of          3/4.9.12       Note I D. M. Crutchfield to W. G. Counsel              6/9/81  fuel Damaging A((tdents ($1P       charcoal f 6 f ters during f uel movement lepic IV-20)                       operations in accordance with $15.
W. G. Counssi to D. M. Crutchiteld              5/26/81 (ontrol Rod Position Indica.       Def ers sutaittal of 1.5. (kanges (uncern-             3/4.1.3.2 & mote 2 l                                                           tion                               tag rod position until 5IS package is sub-             3/4.1.3.3 sitted.
l l
W. G. Counsel to D. M. Crutchf 6 eld            9/28/81 Radiolog6(al Consequences of       Converston to 515 (ontains provtston for               3/4.4.8        Note 3 l                                                           failure of Small 16nes Carry-l todene *oncentration. (T may incorporate l                                                           ing Primary Coolav.t Outside       the 1.0 microCerie/qm iodine limit, but (ontainment (SEP fopic IV-         shossid not te mandated. The total specific 16;                                a( t i e t t y l imi t o f 68/f m ic roc ur n e/ge may l
be adequate.
l l   W. G. Counsel ta D. M. Crutthfield              2/1/82  Test eng of Rea(tor Irap Sys-      Antic 6 pates resolution between survetl-              3/4.3.1 &      Motes 4 & 5 tee ased logineered Safety         lante f requency shom. In current le(h.                3/4.3.2 f eatures, including Response      Spec. & 515. There is also some necessary Time festing (SIP Topic VI-        surveillan(e be6ag performed wheth is not 10.A)                              part of (urrent ieth. Spets. This is to be resolved in 515.
W. G. Counsli to D. M. Crutchiteld              2/8/82  Containment Purge and Vent-        The 6ssue of locked closed purge valve well            3/4.6.1.7      hate 6 6ag                                be addressed in $15.
W. G. Counsil to D. M. Crvtthfield              8/19/82 Temporary Butidings On-56te        Plans to ellainate Tech. Spet. Sc(160s 5.2             [listnoted when $15 is sutaitted.
D. M. Crut(hf teld to W. G. Counstl             8/27/82 SIP Iopic IV-19 loss of            Ihe Safety lealuation (51) for this topic is           3/4.6.1.2 Coolant At<6 dent. Resultsng       dependent on the assu=pt 6ons Insted in Table 1        3/4.6.2 from Spe(trum of Postulated        of the 51 f or (ontainment lealtage recircula-         3/4.1.11 P6peng Breaks with the if(P8       tion fan toolers felter eff4(sency and primary (Radiologic al (onsequent es)      availiary building f alters ef f t(sency.
W. G. Couns el to D. M. Crutthfield &           9/8/87  %diological Consequences of        515 limits f or iodine contentrat ton will be          3/4.4.8 &      hate 3 D. M.         truts hf ield to tf. G. Couns e l F/15/82 Main Steam tone fasture out-       inc luded in t he 515.                                 3/4.7.I.4 s ette Cont ainment (SIP fops (
IW 2)

Pa9e 2 of Il (1Lausil_ernts to umG Date of Letter                                                                                                                                                                                                 SIS Letter         hatuett                                         Casussteret                                                                       Section #    (eentets D. G. Counsti to D. G. t tsenhut                   12/30/s2       Aesponse to Generic letter                      Settten 3/4.6.1.8 of if 515 addresses                                             3/4.6,1.7    hete 6 82 86. InfRIG 073F Tech.                        (entainment purge valve integrity and Spets.
serwelliance requirements. These changes are scheduled as part of 515. he other (eemsteents.
D. G. Counsel to D. M. Crutchaseld                 3/15/83        Systematic Evaluat ten Pro-                     I.      SIP Topic V IO.8 *aMR System Sella-                                       3/4,4.9.3    mete 2 gram Integrated Assessment                              b686ty*
I.5. well se modefied in $15 to requ6re the overgeressure protect 6en system to be in serette preer to tattiat6en of AMR
                                                                                                                                                <eeltag and to remain in serv 6te at all ttars AMel (eeltag is in progress and
                                                                                                                                                 ***ernate wenting means are not prev 6ded.
: 2.      SIP Top.- VI-I "Organic m terials 4                                       3/4.5.4     mete 8 Post Actie..-t Cheesstry*
Surve611ance requiremsets for Trise-dtum Phosphate baskets will be encluded en STS.
: 3.      SIP Top 6c VI-lo. A
* testing of Reacter                                  3/4.3.1 &   hate 4 Trip System & tagineered Safety f eatures.                                3/4.3.2 Including Sesponse Itse festing*
thannel (hecks of the steam flew than.
nels will be required on a ence per shift basis. Seguirements for channel t hetss, f unctlenal tests and calsbratten w611 be included in 115 for all RP5 parameters.
: 4.      51P lepic VIII 3.A ~5tation Battery Test                                   3/4.8.2     hote 9 peguireersts*
Convert 6ag to 515 includes requirement for battery testeng.

Cennitaants to IK Date of                                                                              STS Letter                             Latter  inhiect                         Ca mitammt                                 $ection f Casamats C. C. Ceumsll to D. H. Crutchfleid 6/2/83  Systematic Evaluatten           I. 5(P Topic V-5 (Section 84.16.2-        3/4.4.6.1 mete le Program Integrated                  "Operability Aeguirements*)
Assessment Supplemental             Provide specified servelliance Infomet ten                         requirements for leakage detectlen systems.
: 2. SEP Topic V-IO.A (Sectlee 84.18 -       3/4.4.7   Note 11
* Residual Heat Asmoval Systee Heat tachanger Tube falleres*)
Adept lletts en pr6 eery coelaat cheelstry.
: 3. SIP Topic V-10.8 (Sectlen 4.19.I -     3/4.4.9.3 mete 7   '
                                                                                   *Petential for Sestdoal heat Somov-                       l al System Overpressertration*; aise                       j g                                                                                  Sectica 4.70.3
* Residual Meat                        l Reesval Systen")
Place overpressere protecties system in operatten estecever BR cooling is                      l la progress.
4 5(P Topic V-le-B (Sectlen 4.19.3 -        3/4.5.1 & mete 12
                                                                                   *0utage tielts for leergency Core      3/4.5.2 Coeling System C: ;_     -
ts")
Amend I.5. to govern the allemed outage time of an ECC5 trata.
: 5. SIP Topic VI I (Section 4.21 -         3/4.5.4   note S
                                                                                   *0rgaalt Meterials and Post accident Cher'stry")
P; Je means to adjest s g eater Ph.
                                                                                                                        /

Page 4 of Il Ge oi;eents te_]ms Date of                                                                                                 $15 Letter                        Letter   Sad m t                         (set teret                                                   Socitan #      Caesents
: 6. 54? Iepet vi-IS. A (Sect 6ee 4.25 -                      3/4.3.1        sietes 4 & 5
                                                                               *Iesting of Seacter Irsp Systes and                      3/4.3.2 f agineered i.fety f eatures, includ-                    3/4.6.3 4 tag Response-Iter Testeng*)                             3/4.8.1 RPs servesitance and selected res.
poase t eme test ing (e.g., diesel gen-                                                j erater stertup and segment 6ng and
{
tentalmeent tselatten salves).
F. SIP Topic Wil-3. A (Section 4.30 -                       3/4.8.2        Ilote 9
                                                                               *5tatten 84ttery fest Sequirescats*)
Adopt 515 statten battery test requtreernt.
S. SIP Topic RW-2 (Section 4.35
* Spec-                     3/4.F.I.4 & metes 3 & 13 true of Stese System P6peng failure                      3/4.4.8 lastde and Outside Centalement*;
aise Settson 4.39.2)
Adept STS 16etts for ledsne acttwity ta the presery and secondary coolants.
: 9. SIP Topic EW-16 (Section 4.38 -                          3/4.4.8        Note 3 "Rad 6elegical Ceaseguences of Isliere of 5mell tone (arrytag Primery Coelant Outside Contatement")
Adept SIS primary coolant todtne i
Itest.
I U. G. Covasti to M. R. Denten F/6/83  Seacter vessel Serveillance      the regu6 resent in 1.5. Section 6.9.2 to                    Elseinated (apsule Report                  supply this report to Segsee stil be else6+
noted.

Page 5 of Il Cammitments to IK Date of                                                                                                                                  STS Letter                             1. attar LMatt                          Camitatet                                                                                      Sectlan f     M C. C. Ceensil to D. N. Crutchfield 18/21/83 Addittenal lafemettee          le tscorporate the reelsed Millstone Unit 2                                                    3/4.7.4       Note 14 segerdlag peelsed Saatter      snutter T.5..s. appropriately endsfied for Technical Spectitcattens       maddas neck, tate the ST5.
O. C. Counsel to D. M. Crotchfield 3/11/94  Status of Evaluatlee           The fellesta9 latograted Plant Safety As-Completten Schedules from      sessment poport (IP5AA) (Imm(C-8826)
The fotograted Assessment      Sections will he resolved by conversten to STS.
Section 4.16.2 ........................................ 3/4.4.6.1                                                    mote le 4.18 .......................................... 3/4.4.7                                                      mote II 4 .19.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 . 4 . 9. 3      mete F 4.19.3 ........................................                                               3/4.5.1 &      mete 12
                                                                                         .......................... .............                                            3/4.5.2 4.20.3 ........................................                                               3/4.4.9.3      hete F 4.25 ..........................................                                               3/4.3.1
                                                                                      ..........................................                                             3/4.3.2        metes 4 & 5
                                                                                      ..........................................                                             3/4.6.3 &
                                                                                      ..........................................                                             3/4.s.1 4.30 ..........................................                                               3/4.8.2        mete 9 4.35 ..........................................                                               3/4.4.8 &      hetes 3 & 13
                                                                                       .........................................                                             3/4.7.3.4 4.38 ..........................................                                               3/4.4.8        mete 3 4.39.2 ........................................                                               3/4.4.8 &      hotes 3 & 13 l
                                                                                         ........................................                                            3/4.7.3.4                   l l
l

Cammitampts te M Date of                                                                               515 Letter                              Letter   SiddeC1                       Ca mitatat                                   Sectlas f                                   Cammals C. C. Covasil to D. C. fisenhet     2/24/M   NURIG-013F T.5.s. Generic      Plans to address the InsafG-8F37 Items                                                  mete 15 tetter me. 83 3F              d l(b were scheduled for leptementallen by 12/3I/81 as part of comoersten to STS.
These stees are:
I. semble Gas Efsivent steelter (II.F.I.I) 3/4.3.3.5
: 2. Sampitag and halysis of Plant           6.16 (fflueets (II.?.I.2)
: 3. Contatamrat Nige Range Radiatten        3/4.3.3.5 neelter (II.r.l.3)
: 4. Centalamrat Pressure stretter           3/4.3.3.5 (ll.f.I.4)
: 5. Centalampet Water level stentter        3/4.3.3.5 (II F.I.5)
: 6. Contanament Hydrogen leset a                --
(II.r.l.6; U. C. Ceumsil to D. 80. Crutchfield 2/28/M  Saubber f onctlen.; 'esttag    Intends to propese alth the STS an idents-   3/4.7.4                                   mete le tal provisten as contataed la letter dated I/IS/M dich proposed a Ilcease amend-apet for snubber tcstlag for stillstone 2.
O. G. sounstl to J. R. sellier/     3/1/M   Invironmental Technical and    STS package will tacorporate the appre-      3/4.82                                   Note 16    l D. M. CrutchfIeld                           Specification Reporttag        priate radislegical LCOs.                                                                        l Regeiroments l
C. G. Covesil to D. M. Crutchfield  3/22/M  SEP Topic VIII 3.0. DC Pener   Cammits to provide an laspectica r w as      4.8.2.8                                 IIcte 9 t                                               System Sus asttage             dich is coaststent with the surwelllance feentterlag and AnnuncIatlea   prowIsIens centa9med 1a 'he ST5.
u

MU              M                 l Page 7 of II
((**M"en1L19 N8C flate of                                                                                    STS Lilit!                            Ltiltf   LWh.ltil                        (Z"_11*In"                                      Section_f      WA W. G. Covasti t o D. G. l esenhet 1/25/84  9BJ2fG 0731 I 5.                Will address tires (?) thru (ll) en SI$
(f.ener n letter 83 31)         convers6en.
Itees; (2) Post A4 ( edett 'ampl6=3 (11.8.3)           6.16           hote II (3) long Tern Aas al tary f ee3=ater            3/4.1.1.2      Note IF Syst em f eatuat ion (ll .f .l .8)
(4) Able Gas E f fluent ma sters (II.f . l. 8)  3/4.3.3.8      mote IF (5) Sampling and Aaalyses of Plant              3/4.3.3.8 & Note IF Iffluents (11.f.l.7)                     6.16 (6) (omtainment High Range Radtatico            3/ 4. 3. 3. 5  hate 11 Natter (ll.f.l.3)
(1) (antainsect Pressure Mon ster               3/4.3.3.5      hote !1 (11.5.I.4)
(8) (oatainment Water level Monitor             3/4.3.3.5      hote 17 (ll.f.l.5)
(9) (omtainment Hydrogen Monitor                    --
note 18 (ll.f.I 6)                                                            ,
(10) lastrumentattoo for Detettien of           3/4.3.3.5      mote 19 Inadequate Core tooling (II.F.2)
(11) (ont rol Room Habstab6]ity                     --
Note 18 Requ6rement s (Ill.D.3.4)
W. G. (ouasti to W A. Paulson     9/5/84   Gener6t letter 84-13            Intends to proposed restsed souther I.5.s       3/4.7.4        Note le with convers6on to $15. It will t.e sies tar t o t hese a ssued on Mill st one 2 ( Amendment 96).                                                                       .
J. f. Opeba to J   A. Iwol tate t 11/1/85  hew Design of Seat ter          T.S. well regetre plant shutdown if p($         3.4.6.2        hote 20 (avit y Pool Seal               pressure teundary lea 6 age teneet te determined.
t

EUM E Page 8 of Il Cgsmilinn11_10 NBC Date of                                                                                                   STS Letter                              letter   Suhdett                           CommitacSt                                                   Sectirt f     C m is
: f. N. Akstolewics to J. Opeka       4/10/86   Integrated Plant Safety          Include a containment pressere llett of 2.8                  3.6.1.4       mote 23 Assessment Report (NUREG-         psig and a ( ptainment butt tempe.ature                      3.6.1.5 OSM) Section 4.32.4, Service      Itatt of 140 f in the ST5.
Water System (Systee flew Requtrements)
J. f. Ope 6a to C. I. Grimes        9/18/86   IMI Action Plan                 Appropriate changes to T.S. concerning                        3/4.3.3.5     hote 19 Itee II.f.2 . Inadequate Core    ICC system will be included with SIS                                                   l feeling Instrumentation          coneersten.
System Implementatten Report J. f. Opeta to C. I. Grimes         9/19/86  ISAP DeterstnisttC Reutew         Incorporate the requirements for the vent                    3/4.4.11      hote 21 1.42 RCS Vents T5                 system into the $15 conversten.
: f. M. A6stulewicz to J. Opeta       31/13/86 Outstanding Technical             IP5AR Ceemsteents for Te(balcal Spectilca-Spectf6cattens issues
* tiens to be f acloded as part of the STS cearersten.
Sett1r>
4.16 . ... ....... .... ..... ................                3/4.4.6       mete 10 4.IP ......... ................................               3/4.4.7       hete 11 a.g r          ......         .. .            .    . . . . . . . 3/4.5.I &     Note 12 3/4.5.2 4, 69-                                                        3/4.4.9.3     hote 7 4.20 .................          .......................       3/4 4.9.3     mote 7 4.fi ................ . ...................... 3/4.5.4                      note 8 4.25 ....... .................................. 3/4.3.1                     hetes 4 & 5 3/4.3.2 3/4.6.3 3/4.8.1 4.30 .......................................... 3/4.8.2                     Note 9 4.32 .. .. ...... . ..... ...................                   --
4.35 .. . ..................................... 3/4.4.8                     Note 3
: 1. J. nrouka to Document Cent. Desk 2/3/87   Response to open Itees           Table 3.6-1 etll be restsed.                                  3/4.6.3       hote 22 3dentifled Dertag the my S.1986 neet tag on

Cammi ^ - ^s to M fcon't.1 Cammmets
    *f aclosere to this letter conttlood Technical Specif 6catten item to be sehmitted setside the STS conversten effort. The STS will aise laceperate the taclosure I re strements.
not2 I             Several 515 Serveillance requiremsets are met included. These emisstems reflect the actual servelliance capattittles of the CV Feel Storage tulldtag ititer Saan.
sesta 2            med posttise sad 6catten systems are included in these sections.
80012 3            5T5 Itatts for ledine concentrattee are included.
Itet2 4            The servettlance fregusacy for chamael checks are consisteet alth SIS. The serves tlance frequency for channel calibratten is also coaststeet with one escapiten. There are addittenal enceptions to the servet11ance fregusecy *er fonctiemal tests from the ST1. All servettlance frogmencies more deteve6eed to be adognete and not to lepeci the reliability of the system.
Itote 5           Sospense t$ee testtag is incorporated only for (DG startup and seguenctag (Sectlen 3/4.8.I) and Contatammet Iseletten Valves (Sectten 3/4.6.3) he<mese the staf f concluded in ansetG-esM that further actise was met eerranted.
thet 6             This tssee is addressed.
seet2 7            The comtteset to have the LTOP System in service prior to initiattee of aper coeltag is tecluded.
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Page 10 of 11 (peilgestis te MC fcon't.)
CSE ents f(on't.1 seets 8  Surveillance reestrespets for ISP bas 6ets is enciasded. This provides the erans to Mjest same water pes.
I6ots g   the 515 reestreerots for battery testing ts taclanded.
seats 10 81WittG-stM states that tv es to prov6de act 6 ens to be tden (e.g., plant thetde==) to a(comuudete any reselting fa61eres of ledage detection egulperet due to settetc esset. The twised T.S. 6actisees a sierveellance regelremset (not to SIS) to declare tte systems operable fellestag a setse6c event of a geven me9n6tude, state II The STS leetts en presery coolant chee6stry are adopted.
smote 12 The allemed estage tsee for an Etti subsystem is incorporated and is coaststent with the 515 for Taeg 2 354*f.
mete 13  the SIS leetts for .edene actietty on secondary $6de has been 6acorpe.ated.
mete le  Souther test tag is tacorporated. 9tillstone 2 had a provisten that requiremrats to perform the funct6enal test 6ag de not become effective untti test tag eguiparat is procered and appropriate testing procedures and acceptance crateria developed. CT will have a s6e6ter preetsten.
o                                _ . _ _ _ . _ . - _ _ _ _ . _ _ _ _ . - - - . - , _ - . . - _ . - - - . - - - - - - -                            -- -        - -     -- -

Page il of Il W inamis te E fcon't.1 Cammets itma't.1 mete 15  All 12/31/88 empafC-GF3F tiens are addressed with the esception of the centlasses tedicattee of hydrogse concentratten le costelauset. This stem is heleg addressed as part ef the 15AP program.
mote 16  Radlelegical Lt0s are 4ecorporated.
mete IF  This itse has been addressed.
siete IS This stem ts betag addressed as part of the 15AP program.
mete 19  K5 Suhteeltag storgte sesetters and meacter Coelant Teasprature Dif fereettal testrumsetetten are esed for the detecties of feedegnete core cooltag and are addressed to the referenced sectism.
mete 29  This roomtrumpet is lacorporated.
mote 21  The K 5 weets are locarporated.
mete 22  Table 3.6-8 contates the cometterets.
teste 23 These lletts are *acorporated into the ST5.
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