View source for IA-87-643, Trip Rept of 870722-23 Visit to Braintree,Ma to Discuss Proposed Safety Enhancement Program at Plant

{{Adams
| number = ML20237L175
| issue date = 08/17/1987
| title = Trip Rept of 870722-23 Visit to Braintree,Ma to Discuss Proposed Safety Enhancement Program at Plant
| author name = Su T
| author affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| addressee name = Hodges M
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000293
| license number = 
| contact person = 
| case reference number = FOIA-87-643
| document report number = NUDOCS 8708200109
| package number = ML20234D382
| document type = MEETING MINUTES &amp; NOTES--CORRESPONDENCE, TRIP REPORTS
| page count = 3
}}

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2(                     AUG 171967 l     MEMORANDUM FOR:       M. W. Hodnes, Chief Reactor Systems Branch Division of Engineering & Systems Technology                    ?
THRU:                 R. C. Jones, Section Chief                                       '            '
Thermal-Hydraulics Performance Section Reactor Systems Branch                                           g Division of Engineering & Systems Technology rROM:                 T. M. Su                                                           t Thermal-Hydraulics Performance Section                            ;!
Reactor Systems Branch Division of Engineering & Systems Technology                       3j

==SUBJECT:==
TRIP REFORT - MEETING JULY 22-23, 1987,                              $
BOSTON EDISON COMPANY g
s On July 22, 1987, a meeting was held at Boston Edison Company (BECo), Braintree, MA       j oetween the staff and the representatives from BEco. The purpose of this                 l$
meeting was to discuss the Safety Enhancement Program (SEP) proposea by the               %
BEco for the Pilgrim Nuclear Station.                                                       Y t
The SEP includes modifications and/or additions to the existing containment                    '
systems, electrical systems and reactor systems for the Pilgrim facility.                    -
Enclosure 1 is a copy of the meeting handout, which shows all issues discussed in the meeting.                                                                          v 9
The discussion of the issues related to the Reactor Systems, for which I was                A responsible for the safety review, are summarized in the following:                       1 ~
: 1. Automatic Depressurization System (ADS) Logic Modifications                                      !
lhis modification provides a timed bypass of the high drywell pressure                                 I initiation signal and a manual inhibit of existing ADS actuation logic. This modification responds to the BWROG evaluation for Item II.K.3.18 of NUREG                              i 0737.
j The licensee has submitted a proposed Technical Specification change (BECo
     !etter dated May 20, 1987). I indicated that SRXB has completed its review and finds the proposeo modifications acceptable.
: 2. Addition of Enriched Boron to Standby Liquid Control System                                      i The use of enriched sodium pentaborate in the Standby Liquid Control System                        'l (SLCS) allow Pilgrim to meet the requirement of the Anticipated Transient                              l Without Scram ( ATWS) Rule (10 CFR 50.6%) with one pump operable, thereby                              !
retaining the redundancy of the SLCS design.                                                           I i
3 8708200109 070817 b                                     (
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j M. W. Hodges                                                                                '
4e licensee f ubmitted a proposed Technical Specification change which was opproved by the staff.                                                                          l I         !
: 3.       AlWS Fu dwater pump Trip This change hiil provide an automatic trip to all feedwater pumps at 1400 psig    j reactor vessel pressure. This setpoint is selected so that feedwater pump        i trip occurs only when an ATWS event occurs following closure of Main Steam         i isolation Valves. It serves as a backup to the existing ATWS protection. The     4 current ATWS design co'isists of trips of the recirculation pumps and               !
initiation of the Automatic Rod Insertion (ARI) system on low water level or       :
high reactor pressure.                                                              -
The existing reactur feedwater pump trip logic Will be rodified to accept an oaditional trip signal from ATWS. A new trip coil (in addition tn the existing trip coil) will be installed in the breaker associated with each reactor feed pump. The coils are " energized to trip" coils.
I inoicated that since the trip pressure of 1400 psig is expected to be well obove most transient pressures, this rrodification may not contribute to the reduction of core melt f requency. Instead, this modification may increase                   ' '
the potential of feedwater pump trip f requency. BECo indicated that they have not determined how much improvement to core rnelt frequency results from this irodification. However, BECo expected that the additinnal trip logic will not have en adverse safety impact.                                                                 ;
4 Modification to Reactor Core Isolation Cooling (RCIC) System Turbine Exhaust Trip Setpoint During Station Blackout (SBO) events, the RCIC system is available to supply                   i cooling water to the reactor and maintain the reactor water level. The RCIC purp is driven by a turbine using the primary system steam. The turbine exhaust is piped to the suppression pool. Continuous discharge of the steam                     l to the suppression pool, however, will increase the suppression pool                         j temperature and the containment pressure. The existing RCIC exhaust trip pressure is 25 psig, which will be reached at about 8.5 hours into the SB0                    -
event.                To extend the use of the RCIC system, the licensee proposed to ircrease the trip pressure to ab psig. This increase of trip pressure will allow the RCIC system to operate until about 15.5 hours into the event.
I indiccted that our experience with suppression pool hydrodynamic loads show that steam condensation loads increase significantly with increasing exhaust back pressure. Assessment of the magnitude of these loads is required in                      ;    3 order to ensure that the RCIC exhaust pipe will not fail during the increased trip setpoint. The SECo staft indicated that they had assessed the loads on the batis of static pressure.                Since experittents and analytical methods             i' i
indicate that the dynamic load differs substantially from static load, BEco's                 '     '
pr esent methed based on static pressure is not sufficient to ensure the                      r
)       integrity of the PCIC exhaust pipe. I indicated that ddditi"al analysis on

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M. W. Hodges                                                                                                                                                                                                                                                                                                               .
the dir clearing loads and steam condensation loads will be required to support this modification.                                                                                                                                                                                 <:
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: 5. Additinnal ATWS Recirc.ulation Pump Trip
                                                                                                                                                                                                                         's Trip of the recirculation pumps 15 a feature for the mitigation of ATWS events. Pilgrim currently has the capability of tripping the recirculation pumps by opening the field breakers. Installation of a new trip coil within fg
'                      the breaker associated with each recirculation pump motor / generator set will                                                                                                                   e increase the pump trip reliability.                                                                                                                                                              }
T The licensee has analyzed this modification and concluded that it does not                                                                                                                          }
degrace the existing recirculation system, ATWS system or safety related power                                                                                                                     ;
supplies. I indicated that the ICSB staff will review the logic modification                                                                                                                      y and provide coment later.                                                                                                                                                                          f=
On July 23, the staff visited the P11 grim facility and met with the fiRC                                                                                                                           }
resicent inspector and Pcgion I personnel to brief the discussions we had on                                                                                                                        t July 22 meeting at BEco office.                                                                                                                                                                   f I
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T. M. Su                                                                                                                                            j Thermal Hydraulics Performance Section                                                                                                                f Reactor Systems Branch                                                                                                                                a Division of Engineering & Systems Technology                                                                                                          6 7

==Enclosure:==
f As stated atsrRIBUTION <               p , . . . .W Central fit 6 " q                                                                                                                                                                                }43 SWXB R/F
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M. W. Hodges                                                                                                                                                                                               i R. Jones T. Su T.'Su Rif SEAD Pctrbers l
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